Progress in Nuclear Energy最新文献

{"title":"Assessment of interfacial area concentration models in RELAP5 and TRACE in application to bubbly and cap-bubbly flows in a large square channel","authors":"Haomin Sun ,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106272","DOIUrl":"10.1016/j.pnucene.2026.106272","url":null,"abstract":"<div><div>A thermal-hydraulic system analysis code is essential for evaluating the performance and safety of nuclear reactors. RELAP5 and TRACE are two widely used system codes that employ the two-fluid model to simulate gas-liquid two-phase flows. Interfacial area concentration (IAC) is one of the key flow parameters required to close the two-fluid model, and its modeling performance directly affects the simulation accuracy. Given its importance, several IAC models have been developed using existing experimental databases for various flow channels, including pipes, rod bundles, and rectangular channels. However, no IAC model development has been conducted on large square channels, despite their significance in the design of advanced light-water nuclear reactors, such as the ESBWR. To address the need for reliably analyzing two-phase flows in large square channels, the two-group (2G) IAC models in RELAP5 and TRACE were evaluated using a large square channel experiment that included 2G flow measurements for bubbly and cap-bubbly flows. Here, the 2G approach refers to a method that classifies bubbles into two bubble groups based on the drag coefficient acting on bubbles. Significant prediction errors were identified in both sub-models that comprise the RELAP5 and TRACE IAC models: the 2G void fraction (VF) model and the VF-to-IAC model (calculating IAC from VF). A recently developed 2G drift-flux correlation was recommended to improve the prediction accuracy of the 2G VF. The VF-to-IAC models in RELAP5 and TRACE were modified based on the experimental data. The mean relative deviations of IAC prediction for large square channels were −4 % and −1 % using the modified RELAP5 and TRACE IAC models, respectively, and improved from 73 % and −26 % using their respective default models.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106272"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039147","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental and numerical study of particle resuspension on turbine surface with film cooling","authors":"Xiaozhong Wang ,&nbsp;Qi Sun ,&nbsp;Wei Peng ,&nbsp;Yinhai Zhu ,&nbsp;Peixue Jiang ,&nbsp;Suyuan Yu","doi":"10.1016/j.pnucene.2026.106285","DOIUrl":"10.1016/j.pnucene.2026.106285","url":null,"abstract":"<div><div>Resuspension significantly influences the transport and deposition of graphite particles within the turbine of high-temperature gas-cooled reactor (HTGR). This work involved setting up a visualized experimental setup to observe particle resuspension under film cooling conditions. High-speed cameras and hot-wire anemometers were employed to measure the dynamic process of particle resuspension and the characteristics of the near-wall flow field. Based on the torque balance model, a particle resuspension model suitable for the HTGR turbine conditions was developed, and the accuracy of the model was verified with experimental results. Combining experimental measurements and numerical simulations, the mechanism by which film cooling affects particle resuspension was investigated, and the dominant role of counter-rotating vortices was revealed. Furthermore, the influence of key factors such as flow velocity, blowing ratio, and particle size on resuspension characteristics was systematically analyzed. Finally, the particle deposition process and resuspension process were coupled for analysis, and the distribution characteristics of particle deposition on the turbine surface under film cooling conditions were studied. The findings from this study not only offer experimental insights into particle resuspension behavior but also contribute valuable guidance for the design and safety assessment of HTGR helium turbines.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106285"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189704","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Modeling terminal velocity and aspect ratio of a single bubble in distorted-particle and cap bubble regimes for pool scrubbing simulation","authors":"Yuria Okagaki ,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106267","DOIUrl":"10.1016/j.pnucene.2026.106267","url":null,"abstract":"<div><div>Pool scrubbing is an essential filtration process that prevents the release of radioactive aerosols into the environment during severe accidents at nuclear reactors. During such events, a mixture of steam and/or various noncondensable gases containing radioactive aerosols is injected into the pool, where it forms bubbles. During aerosol particle transport, bubble dynamics play a notable role in determining the efficiency of particle removal within the bubbles. Clarifying these mechanisms is crucial for accurately modeling pool scrubbing processes. In practice, pool scrubbing codes based on the lumped parameter (LP) approach incorporate particle removal models; however, key hydrodynamic parameters—such as bubble rise (terminal) velocity and bubble aspect ratio—are modeled using simplified assumptions. The present study aimed to verify the validity of existing correlations for terminal velocity and bubble aspect ratio, with the long-term objective of proposing improved models. As part of this fundamental research, the behavior of a single bubble rising in a stagnant liquid, simulating a pool environment, was visualized through computational fluid dynamics (CFD) simulations employing an interface tracking/capturing method. The effective aspect ratio was calculated based on the bubble interfacial area and volume. Furthermore, correlations for single-bubble terminal velocity and aspect ratio were developed, taking into account the transition in bubble shape as the volume-equivalent diameter increased. The findings from the present study are expected to provide a foundation for the development and evaluation of future standalone pool scrubbing codes, including improved correlations for terminal velocity for bubble swarm, bubble aspect ratio, and other relevant hydrodynamic parameters.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106267"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189697","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on geyser boiling parameter characteristics in vertical sodium heat pipes","authors":"Youlan Yuan ,&nbsp;Zaiyong Ma ,&nbsp;Yugao Ma ,&nbsp;Zheng Zhou ,&nbsp;Luteng Zhang ,&nbsp;Simiao Tang ,&nbsp;Liangming Pan","doi":"10.1016/j.pnucene.2026.106270","DOIUrl":"10.1016/j.pnucene.2026.106270","url":null,"abstract":"<div><div>State of the liquid pool is vital to the stable operation of vertical sodium heat pipes. Geyser boiling in the pool region of vertical sodium heat pipes would induce significant temperature and pressure oscillations, which could negatively impact heat transfer and structural safety. In this paper, experiment was conducted to investigate the effects of input power, filling ratio, and evaporator length on the characteristics of geyser boiling in a vertical sodium heat pipe, including average oscillation period, amplitude, bubble size, and vapor generation amount. The results indicated that both the average oscillation period and amplitude decreased with increasing heating power, as the system at lower heating powers tended to exhibit non-periodic geyser boiling, leading to larger oscillation periods and amplitudes. A higher filling ratio could significantly shorten the oscillation period, indicating that more oscillation cycles occur within the same time interval. Theoretical analysis combined with experimental observation revealed that the bubble diameter during geyser boiling mainly ranged from 0.9 to 2.3 D (inner diameter of the heat pipe). Increasing filling ratio and evaporator length inhibited the formation of large bubbles, resulting in significantly reduced system fluctuations. Likewise, when the filling ratio was increased from 41.67 % to 83.33 %, the variation in bubble size over time became smaller. A dimensionless vapor generation factor <em>f</em>, based on vapor production inferred from temperature oscillations, was proposed to characterize vapor generation amount of each bubble release. The factor <em>f</em> remained generally low and exhibited no simple monotonic relationship with heating power. It showed enhanced instability at lower filling ratios and a non-monotonic dependence on evaporator length, reaching a peak value near 350 mm.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106270"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039086","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Review of self-powered neutron detectors for reactor instrumentation","authors":"Jack A. Lanza ,&nbsp;Lei R. Cao","doi":"10.1016/j.pnucene.2025.106230","DOIUrl":"10.1016/j.pnucene.2025.106230","url":null,"abstract":"<div><div>Self-powered neutron detectors (SPNDs) have been essential instruments for in-core flux monitoring in light-water reactors (LWRs) for over six decades. This paper provides a comprehensive review of SPND principles, materials, and configurations, along with their roles in the third-generation (Gen-III and Gen-III+) reactor designs. The review compares performance characteristics of common emitter materials such as Rh, V, Co, and Ag, and examines emerging candidates, including Cd, Er, and Hf, for prompt response applications. Developments in modeling, compensation algorithms, and signal processing, including analytical, analog, and digital methods such as Kalman filtering, are summarized to highlight improvements in accuracy and response time. Recent advancements have extended to fast-neutron and gamma-sensitive SPNDs, along with opportunities for integration in digital twin and AI-based reactor monitoring frameworks. The paper also identifies research gaps in detector-level integration, materials optimization, and adaptability for small modular and microreactor environments. These findings underscore the continued importance of SPNDs as reliable, radiation-tolerant sensors within evolving nuclear instrumentation and safety systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106230"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927224","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Transformation of montmorillonite under thermal-saline conditions in HLW repositories: A systematic review","authors":"Cong Liu ,&nbsp;Yong-Gui Chen ,&nbsp;Rao-Ping Liao ,&nbsp;Zhao Sun ,&nbsp;Yu-Cheng Li ,&nbsp;Sheng-Fei Cao ,&nbsp;Wei-Min Ye","doi":"10.1016/j.pnucene.2025.106201","DOIUrl":"10.1016/j.pnucene.2025.106201","url":null,"abstract":"<div><div>The long-term stability of montmorillonite under near-field thermal and saline conditions in a high-level radioactive waste (HLW) repository is a critical concern as potential transformation of montmorillonite may occur under such circumstances. This review aims to summarize current knowledge on the transformation of montmorillonite under thermal and saline conditions relevant to HLW repositories, highlighting key mechanisms and patterns reported in the literature. Laboratory and in-situ tests indicate that montmorillonite can undergo thermo-saline transformations into illite, kaolinite, chlorite, or other minerals, and may also exhibit changes in negative layer charge. The effects of changes in negative layer charge and interlayer cations on the swelling properties of montmorillonite were analyzed. In addition to temperature and salinity, factors such as montmorillonite type, accessory minerals, pressure, reaction time, and solid/liquid ratio influence these transformations of montmorillonite.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106201"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841097","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"How much residual water remains in a canister after vacuum drying for spent nuclear fuel storage?","authors":"Ji Hwan Lim ,&nbsp;Seung-Hwan Yu ,&nbsp;Gyung-sun Chae ,&nbsp;Kyung-Wook Shin ,&nbsp;Nam-Hee Lee","doi":"10.1016/j.pnucene.2025.106231","DOIUrl":"10.1016/j.pnucene.2025.106231","url":null,"abstract":"<div><div>This study presents a rigorous experimental investigation of the direct quantification of residual water remaining inside spent nuclear fuel (SNF) canisters following vacuum drying—a pivotal operation for limiting long-term corrosion, radiolytic gas generation, and structural degradation during dry storage. Using a purpose-built laboratory facility that replicates realistic canister geometries, we examined initial moisture inventories of 10–77.9 g. Drying employed a carefully modulated pressure-reduction protocol, achieving standard-compliant conditions (&lt;3 torr for 30 min), after which the canisters were backfilled to 2 bar with inert gas. Direct moisture measurements used an AquaVolt 2 electrolytic trace analyzer under controlled helium pressurization, returning highly resolved residual-moisture concentrations in ppmv. Across all scenarios, the final internal moisture converged reproducibly to ∼500–600 ppmv (≈0.15 g), indicating a predictable endpoint under standard-compliant drying, with instrument-traceable repeatability across trials. A bounding comparison placed these values 3–4 × below a conservative worst-case assumption (treating the entire residual head as H2O vapor; ∼0.569 g) yet 15–20 × above estimates based on ambient humidity entrainment (∼0.0089 g). Beyond filling a longstanding empirical gap, these measurements have direct engineering and regulatory significance: they supply defendable inputs for hydrogen source-term, corrosion, and pressure-evolution models; enable risk-informed procedure design and QA/QC traceability; and complement current compliance frameworks (e.g., NUREG-2215; ASTM C1553-21) that rely on pressure-based endpoints. Finally, the integrated methodology—controlled vacuum drying, real-time dew-point tracking, and ultra-trace quantitative moisture analysis—provides a practical blueprint for selective, confirmatory direct-moisture verification in SNF dry-storage operations and is transferable to other high-stakes vacuum-dependent industries.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106231"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884612","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on the influence of gas-ingested flow on internal flow characteristics in reactor coolant pumps","authors":"Dan Ni,&nbsp;Jinyu Yang,&nbsp;Xiang Yu,&nbsp;Hongchi Zhong,&nbsp;Sheng Lu,&nbsp;Yuquan Zhang,&nbsp;Bo Gao,&nbsp;Zhong Li,&nbsp;Ning Zhang","doi":"10.1016/j.pnucene.2025.106212","DOIUrl":"10.1016/j.pnucene.2025.106212","url":null,"abstract":"<div><div>The reactor coolant pump is essential to nuclear power plant safety. During accidents like LOCA, gas entrainment leads to complex two-phase flow, degrading performance and challenging operational integrity. Current understanding of the associated flow mechanisms remains limited. To analyze the gas-liquid interaction mechanisms in an axial-flow reactor coolant pump under gas-ingested conditions, this study conducted theoretical investigations on the gas-liquid two-phase flow in nuclear reactor coolant pump. Flow characteristics under various air volume fractions (AVF), bubble diameters, and flow conditions were obtained using large-eddy simulation (LES) method with the Mixture multiphase model. Through experimental verification, it was found that the calculation method has an error within 2 % compared to actual results, confirming the accuracy of the model and numerical methods. The research systematically summarizes the performance variation patterns of the reactor coolant pump under two-phase flow conditions. The gas–liquid interaction mechanisms were elucidated through Q-criterion vortex identification, while the modified behavior of the spherical casing vortex under two-phase operation was analyzed. These findings contribute to nuclear engineering advancements by providing fundamental data for preventing severe two-phase flow conditions in reactor coolant pumps.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106212"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884777","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimentation and simulation research on foreign object wear of cladding-rods interface in high-temperature pressurized water","authors":"Yanke Wen ,&nbsp;Hongyang Chen ,&nbsp;Chuangming Ning ,&nbsp;Yuhan Zheng ,&nbsp;Zhengyang Li ,&nbsp;Zhenbing Cai","doi":"10.1016/j.pnucene.2025.106188","DOIUrl":"10.1016/j.pnucene.2025.106188","url":null,"abstract":"<div><div>The high-speed flow of coolant in a nuclear reactor flushes the fuel rods, causing them to undergo flow-induced vibration (FIV). Additionally, tiny metal impurity particles are captured by the grid as the coolant flows, resulting in fretting wear. These foreign objects are primarily swarf generated during the machining of components in the nuclear reactor. Prolonged fretting can lead to localized wear of the fuel rod cladding, potentially resulting in fuel breakage and the release of fission products, thereby compromising the safety of the nuclear power plant. Foreign object wear is currently an important cause of fuel rod failure, yet little research has been done in this area. The small size of foreign particles renders the experimental studies of foreign object wear challenging. The finite element method (FEM) was employed to simulate the positions and quantities of variously shaped foreign object particles captured in the high-temperature pressurized water environment. In this paper, a new experimental method for foreign object wear was proposed, and the behavior of fretting wear between different foreign objects and cladding tubes is simulated by combined experimental and simulation studies. The study focused on analyzing the effects of particle shape and normal force on abrasion performance. The simulation results indicate that some foreign objects become lodged between the rigid support protrusion and the fuel rod cladding tube, remaining in place despite coolant flushing. Experimental data closely match simulation results at low loads.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106188"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788991","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"foamForNuclear: A unified OpenFOAM multi-physics platform for nuclear applications","authors":"Giovanni Nervi ,&nbsp;Thomas Guilbaud ,&nbsp;Carlo Fiorina ,&nbsp;Mathieu Hursin ,&nbsp;Alessandro Scolaro","doi":"10.1016/j.pnucene.2026.106250","DOIUrl":"10.1016/j.pnucene.2026.106250","url":null,"abstract":"<div><div>OpenFOAM has been extensively used to develop dedicated physics solvers for nuclear applications, ranging from thermal-hydraulics to structural mechanics and neutronics. However, existing frameworks still lack a general-purpose structure for coupling arbitrary physics across multiple regions and meshes. This work presents <em>foamForNuclear</em>, a new OpenFOAM-based open-source multiphysics architecture that addresses this limitation. Building on the established GeN-Foam and OFFBEAT codes, the platform introduces a flexible, modular structure that enables users to configure simulations with different combinations of physics, solvers, and coupling strategies without source code modifications. The platform architecture, physics libraries, and verification and validation processes are detailed. Its capabilities and applicability to both fission and fusion systems are demonstrated through selected test cases, including the simulation of the ORNL 19-pin sodium-cooled fuel assembly, the sloshing of liquid metals in pool-type reactors under seismic loading, and the thermal performance of a dual-coolant lithium-lead fusion blanket.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106250"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977696","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}