Pub Date : 2026-01-26DOI: 10.1016/j.anucene.2026.112152
Owen Mylotte, Benoit Forget
The Random Ray Method (TRRM) neutron transport solver is a robust variation of the Method of Characteristics (MOC) for which ray sampling is uniform in space and angle, as opposed to the typical fixed quadrature cyclical ray tracking. However, there are several classes of problems for which the implicitly assumed uniform sampling distribution of TRRM may not be optimal. This work introduces the Ray Adaptive Stochastic Transport (RASTr) method, which computes statistical weights for an arbitrary spatially and angularly biased ray sampling distribution to provide variance reduction for problems where non-uniform sampling is desirable. The RASTr algorithm is implemented and tested on 1D and 2D test cases with demonstrated improvement in statistical relative uncertainty.
{"title":"Ray Adaptive Stochastic Transport (RASTr): Importance sampling based variance reduction for characteristics method transport solvers","authors":"Owen Mylotte, Benoit Forget","doi":"10.1016/j.anucene.2026.112152","DOIUrl":"10.1016/j.anucene.2026.112152","url":null,"abstract":"<div><div>The Random Ray Method (TRRM) neutron transport solver is a robust variation of the Method of Characteristics (MOC) for which ray sampling is uniform in space and angle, as opposed to the typical fixed quadrature cyclical ray tracking. However, there are several classes of problems for which the implicitly assumed uniform sampling distribution of TRRM may not be optimal. This work introduces the Ray Adaptive Stochastic Transport (RASTr) method, which computes statistical weights for an arbitrary spatially and angularly biased ray sampling distribution to provide variance reduction for problems where non-uniform sampling is desirable. The RASTr algorithm is implemented and tested on 1D and 2D test cases with demonstrated improvement in statistical relative uncertainty.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112152"},"PeriodicalIF":2.3,"publicationDate":"2026-01-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074814","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}
Pub Date : 2026-01-24DOI: 10.1016/j.anucene.2026.112149
Alvin J.H. Lee, Tomasz Kozlowski
In Molten Salt Reactors (MSRs), the easy accessibility of the fuel salt is a potential avenue for plutonium diversion and safeguard inspectors must have the means to detect such diversion to limit nuclear proliferation. In this work, we developed a mathematical basis to translate the radioactivities of select gamma emitting fission products into the fissile isotope ratios of the reactor, and described a methodology to detect plutonium diversion. This work identified 138mCs/134mI as the species pair that can predict the fissile isotope ratios with good accuracy while under chemical effects such as reprocessing. Simulated diversion cases demonstrated good accuracy of around 0.96% discrepancy from the Monte Carlo estimate for 138mCs/134mI, and around 0.034% for other potential pairs when less limiting conditions were considered (e.g., finite but slow precursor removal vs decay). The detection methodology contributes to the existing MSR safeguards efforts, which is necessary for the successful deployment of MSRs.
{"title":"Plutonium diversion detection in molten salt reactors via gamma emitter signature","authors":"Alvin J.H. Lee, Tomasz Kozlowski","doi":"10.1016/j.anucene.2026.112149","DOIUrl":"10.1016/j.anucene.2026.112149","url":null,"abstract":"<div><div>In Molten Salt Reactors (MSRs), the easy accessibility of the fuel salt is a potential avenue for plutonium diversion and safeguard inspectors must have the means to detect such diversion to limit nuclear proliferation. In this work, we developed a mathematical basis to translate the radioactivities of select gamma emitting fission products into the fissile isotope ratios of the reactor, and described a methodology to detect plutonium diversion. This work identified <sup>138m</sup>Cs/<sup>134m</sup>I as the species pair that can predict the fissile isotope ratios with good accuracy while under chemical effects such as reprocessing. Simulated diversion cases demonstrated good accuracy of around 0.96% discrepancy from the Monte Carlo estimate for <sup>138m</sup>Cs/<sup>134m</sup>I, and around 0.034% for other potential pairs when less limiting conditions were considered (e.g., finite but slow precursor removal vs decay). The detection methodology contributes to the existing MSR safeguards efforts, which is necessary for the successful deployment of MSRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112149"},"PeriodicalIF":2.3,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035494","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}
Pub Date : 2026-01-24DOI: 10.1016/j.anucene.2026.112168
Yifan Zhou, Houjian Zhao, Yang Liu
Shell and tube heat exchangers are widely used in nuclear engineering and the petrochemical industries. In the current investigation, cross flow over inline tube bundles with various pitch-to-diameter ratios is simulated by SST k-ω-γ. The mesh near the shear layer region is refined due to the large velocity gradient. The effects of the bounding wall, end wall, and pitch ratio on time-averaged and transient flow fields are systematically analyzed. The increased streamwise pitch results in the impinging point shifting to near θ = 0°. The increased transverse pitch results in a larger influence on side passages. The recirculation region near the end wall is attenuated, resulting in reduced drag and large velocity magnitude. The separated vortices after the middle tubes sway into the main flow. There is a separation vortex near the bounding wall due to the entrainment of separate vortices.
{"title":"Numerical investigation of wall effects on cross flow over inline tube bundles with various pitch-to-diameter ratios","authors":"Yifan Zhou, Houjian Zhao, Yang Liu","doi":"10.1016/j.anucene.2026.112168","DOIUrl":"10.1016/j.anucene.2026.112168","url":null,"abstract":"<div><div>Shell and tube heat exchangers are widely used in nuclear engineering and the petrochemical industries. In the current investigation, cross flow over inline tube bundles with various pitch-to-diameter ratios is simulated by SST <em>k-ω-γ.</em> The mesh near the shear layer region is refined due to the large velocity gradient. The effects of the bounding wall, end wall, and pitch ratio on time-averaged and transient flow fields are systematically analyzed. The increased streamwise pitch results in the impinging point shifting to near <em>θ</em> = 0°. The increased transverse pitch results in a larger influence on side passages. The recirculation region near the end wall is attenuated, resulting in reduced drag and large velocity magnitude. The separated vortices after the middle tubes sway into the main flow. There is a separation vortex near the bounding wall due to the entrainment of separate vortices.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112168"},"PeriodicalIF":2.3,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035468","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}
Pub Date : 2026-01-24DOI: 10.1016/j.anucene.2026.112171
Hoang Hai Nguyen
This study examined the effects of the moderator-to-fuel volume ratio, fuel debris shape, and the number of damaged fuel assemblies on the neutronic features of a partially damaged reactor, where the fuel assemblies in the center region collapse into debris and the fuel assemblies in the outer region are kept unchanged. The partially damaged reactor model was based on the Watts Bar Nuclear 1 reactor. The investigations were performed by the Serpent code. The findings show that in cases where the debris is surrounded by intact fuel assemblies, the change of keff depends on the geometry of the debris. Conversely, in scenarios where the debris is not fully encircled by intact fuel assemblies, the geometry of the debris has a negligible impact on the keff. Additionally, the number of neutrons entering and leaving the debris determines how its shape affects the keff.
{"title":"Neutronic characteristics of a partially damaged reactor model with varying numbers of damaged fuel assemblies","authors":"Hoang Hai Nguyen","doi":"10.1016/j.anucene.2026.112171","DOIUrl":"10.1016/j.anucene.2026.112171","url":null,"abstract":"<div><div>This study examined the effects of the moderator-to-fuel volume ratio, fuel debris shape, and the number of damaged fuel assemblies on the neutronic features of a partially damaged reactor, where the fuel assemblies in the center region collapse into debris and the fuel assemblies in the outer region are kept unchanged. The partially damaged reactor model was based on the Watts Bar Nuclear 1 reactor. The investigations were performed by the Serpent code. The findings show that in cases where the debris is surrounded by intact fuel assemblies, the change of k<sub>eff</sub> depends on the geometry of the debris. Conversely, in scenarios where the debris is not fully encircled by intact fuel assemblies, the geometry of the debris has a negligible impact on the k<sub>eff</sub>. Additionally, the number of neutrons entering and leaving the debris determines how its shape affects the k<sub>eff</sub>.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112171"},"PeriodicalIF":2.3,"publicationDate":"2026-01-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035493","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}
The performance of various oxygen sensors in liquid lead–bismuth eutectic (LBE) alloy varies significantly across different temperature ranges. Therefore, it is necessary to establish a comprehensive temperature-related calibration database to achieve real-time dynamic calibration and compensation for sensor measurements. In this paper, multiple types of oxygen sensors have been developed based on 8YSZ ceramic tubes. The air reference (LSCF/Air, LSM/Air and Ag/Air) oxygen sensors and metal/metal oxide (Cu/Cu2O, Fe/Fe3O4, Ni/NiO, Bi/Bi2O3, and In/In2O3) reference oxygen sensors were tested under different temperature variations to obtain their operating characteristics in different temperature ranges. The air reference oxygen sensors have been demonstrated to exhibit excellent response speed, accuracy and stability within the range of 205 ∼ 550℃. The metal/metal oxide reference oxygen sensor is more suitable for applications in the medium to high temperature range (≥350 ℃). It provides reference data for the operation of non-isothermal lead–bismuth system oxygen sensors.
{"title":"Performance of multiple-type reference electrode oxygen sensors in LBE","authors":"Ruixian Liang, Hui Li, Huiping Zhu, Hao Wu, Haicai Lyu, Zulong Hao, Yang Liu, Fang Liu, Fenglei Niu","doi":"10.1016/j.anucene.2026.112129","DOIUrl":"10.1016/j.anucene.2026.112129","url":null,"abstract":"<div><div>The performance of various oxygen sensors in liquid lead–bismuth eutectic (LBE) alloy varies significantly across different temperature ranges. Therefore, it is necessary to establish a comprehensive temperature-related calibration database to achieve real-time dynamic calibration and compensation for sensor measurements. In this paper, multiple types of oxygen sensors have been developed based on 8YSZ ceramic tubes. The air reference (LSCF/Air, LSM/Air and Ag/Air) oxygen sensors and metal/metal oxide (Cu/Cu<sub>2</sub>O, Fe/Fe<sub>3</sub>O<sub>4</sub>, Ni/NiO, Bi/Bi<sub>2</sub>O<sub>3</sub>, and In/In<sub>2</sub>O<sub>3</sub>) reference oxygen sensors were tested under different temperature variations to obtain their operating characteristics in different temperature ranges. The air reference oxygen sensors have been demonstrated to exhibit excellent response speed, accuracy and stability within the range of 205 ∼ 550℃. The metal/metal oxide reference oxygen sensor is more suitable for applications in the medium to high temperature range (≥350<!--> <!-->℃). It provides reference data for the operation of non-isothermal lead–bismuth system oxygen sensors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112129"},"PeriodicalIF":2.3,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035495","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}
Pub Date : 2026-01-23DOI: 10.1016/j.anucene.2026.112151
Samuel Gant Dotson , Madicken Munk , Kathryn Dorsey Huff
Energy system optimization models are a class of tools designed to optimize energy planning and are used by energy planners and decision-makers to generate insights that inform energy policy. However, existing tools are challenged by real-world scenarios which require optimization across multiple objectives. In this paper, the multi-objective energy system optimization framework, Osier, is demonstrated. Osier leverages genetic algorithms to calculate a set of co-optimal solutions called a Pareto front. Osier also introduces a novel algorithm to identify a subset of maximally different solutions within the sub-optimal space to address structural uncertainty related to unmodeled objectives. By producing multiple solutions, Osier gives modelers and decision-makers the tools to meaningfully engage with public stakeholders and learn their preferences, thereby attending to issues of procedural and recognition justice. This work verifies Osier’s suitability for energy modeling problems with two in silico experiments. The first set of experiments compare Osier to a more mature energy system optimization model, Temoa, to verify that Osier produces results consistent with known methods. The results for a least-cost optimization with Osier and Temoa show strong agreement, within 0.5% of each other. In the second, Osier reanalyzes a set of nuclear fuel cycle options from the SET tool through the lens of Pareto optimality.
{"title":"Demonstrating the Osier framework for energy system and nuclear fuel cycle optimization","authors":"Samuel Gant Dotson , Madicken Munk , Kathryn Dorsey Huff","doi":"10.1016/j.anucene.2026.112151","DOIUrl":"10.1016/j.anucene.2026.112151","url":null,"abstract":"<div><div>Energy system optimization models are a class of tools designed to optimize energy planning and are used by energy planners and decision-makers to generate insights that inform energy policy. However, existing tools are challenged by real-world scenarios which require optimization across multiple objectives. In this paper, the multi-objective energy system optimization framework, <span>Osier</span>, is demonstrated. <span>Osier</span> leverages genetic algorithms to calculate a set of co-optimal solutions called a <em>Pareto front</em>. <span>Osier</span> also introduces a novel algorithm to identify a subset of maximally different solutions within the sub-optimal space to address structural uncertainty related to unmodeled objectives. By producing multiple solutions, <span>Osier</span> gives modelers and decision-makers the tools to meaningfully engage with public stakeholders and learn their preferences, thereby attending to issues of procedural and recognition justice. This work verifies <span>Osier</span>’s suitability for energy modeling problems with two <em>in silico</em> experiments. The first set of experiments compare <span>Osier</span> to a more mature energy system optimization model, <span>Temoa</span>, to verify that <span>Osier</span> produces results consistent with known methods. The results for a least-cost optimization with <span>Osier</span> and <span>Temoa</span> show strong agreement, within 0.5% of each other. In the second, <span>Osier</span> reanalyzes a set of nuclear fuel cycle options from the <span>SET</span> tool through the lens of Pareto optimality.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112151"},"PeriodicalIF":2.3,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035492","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}
Pub Date : 2026-01-23DOI: 10.1016/j.anucene.2026.112164
Feifan Liu , Simian Qin , Quanyao Ren , Haidong Liu , Lele Zheng , Shanshan Bu , Penghui Zhang , Deqi Chen
This paper conducts a numerical investigation of flow characteristics and vibrations in a 7-pin wire-wrapped fuel rods bundle subjected to lead–bismuth eutectic (LBE) axial flow. In this paper, the flow characteristics of 7-pin rods bundle with two different wire-wrapped diameters is analyzed firstly. Then, the effect of the constraint of the adjacent rods to the center rods on the vibration response of the bundle is analyzed. Meanwhile, the effect of constraint conditions at the upper end of the bundle on the vibration behavior is discussed. It is indicated that increasing the wire-wrapped diameter changing from 1.9 mm to 2.0 mm impacts flow characteristics insignificantly. The fluid force on the central rod surface exceeds that on adjacent rods in the bundle configuration. Notably, when introducing normal contact between the wire of the center rod and the adjacent rods, the vibration displacement at the central rod’s midpoint decreases by 81 %, while the dominant frequency increases from 15 Hz to 40 Hz. As the upper-end constraint condition changes from simply support, Y-direction support to no support, the vibration displacement of the center rod changes significantly along the axial direction. When the lower end is fixed support and the upper end is no support, the vibration displacement of the upper end increases along the axial direction, and the root mean square at the upper end (Z = 1500 mm) reaches 68.70 μm. These findings provide a theoretical foundation for the design optimization of wire-wrapped rod bundles in lead–bismuth-cooled fast reactors.
{"title":"Numerical study on the vibration characteristics of 7-pin wire-wrapped fuel rods bundle in lead-bismuth eutectic flow with different constraint conditions","authors":"Feifan Liu , Simian Qin , Quanyao Ren , Haidong Liu , Lele Zheng , Shanshan Bu , Penghui Zhang , Deqi Chen","doi":"10.1016/j.anucene.2026.112164","DOIUrl":"10.1016/j.anucene.2026.112164","url":null,"abstract":"<div><div>This paper conducts a numerical investigation of flow characteristics and vibrations in a 7-pin wire-wrapped fuel rods bundle subjected to lead–bismuth eutectic (LBE) axial flow. In this paper, the flow characteristics of 7-pin rods bundle with two different wire-wrapped diameters is analyzed firstly. Then, the effect of the constraint of the adjacent rods to the center rods on the vibration response of the bundle is analyzed. Meanwhile, the effect of constraint conditions at the upper end of the bundle on the vibration behavior is discussed. It is indicated that increasing the wire-wrapped diameter changing from 1.9 mm to 2.0 mm impacts flow characteristics insignificantly. The fluid force on the central rod surface exceeds that on adjacent rods in the bundle configuration. Notably, when introducing normal contact between the wire of the center rod and the adjacent rods, the vibration displacement at the central rod’s midpoint decreases by 81 %, while the dominant frequency increases from 15 Hz to 40 Hz. As the upper-end constraint condition changes from simply support, Y-direction support to no support, the vibration displacement of the center rod changes significantly along the axial direction. When the lower end is fixed support and the upper end is no support, the vibration displacement of the upper end increases along the axial direction, and the root mean square at the upper end (Z = 1500 mm) reaches 68.70 μm. These findings provide a theoretical foundation for the design optimization of wire-wrapped rod bundles in lead–bismuth-cooled fast reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112164"},"PeriodicalIF":2.3,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035540","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}
We investigated several specific activities produced at a location close to the reactor core of the research reactor of KUR, by comparing experimental results with the results of simulations using MCNP5 and ORIGEN codes. In the simulation using MCNP5, we applied a detailed geometry that precisely describes the interior of the nuclear reactor and simplified geometries that homogenized materials in the nuclear fuel assembly by maintaining the bulk density and bulk chemical composition. The calculation time increased with increasing number of cells in the geometry input. The specific activities were not affected by the homogenization of the geometry. We confirmed that nuclides produced from trace elements in concrete can be simulated with an accuracy of approximately 20% at locations relatively close to the nuclear core. In the case of KUR, the homogenization of the geometries can contribute to shortening the calculation time without degrading the precision.
{"title":"Calculation speed and precision of activation simulation using detailed and homogenized geometries","authors":"Norikazu Kinoshita , Takuma Noto , Kazuaki Kosako , Yuki Sasaki , Kazuyuki Torii , Makoto Inagaki","doi":"10.1016/j.anucene.2026.112159","DOIUrl":"10.1016/j.anucene.2026.112159","url":null,"abstract":"<div><div>We investigated several specific activities produced at a location close to the reactor core of the research reactor of KUR, by comparing experimental results with the results of simulations using MCNP5 and ORIGEN codes. In the simulation using MCNP5, we applied a detailed geometry that precisely describes the interior of the nuclear reactor and simplified geometries that homogenized materials in the nuclear fuel assembly by maintaining the bulk density and bulk chemical composition. The calculation time increased with increasing number of cells in the geometry input. The specific activities were not affected by the homogenization of the geometry. We confirmed that nuclides produced from trace elements in concrete can be simulated with an accuracy of approximately 20% at locations relatively close to the nuclear core. In the case of KUR, the homogenization of the geometries can contribute to shortening the calculation time without degrading the precision.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112159"},"PeriodicalIF":2.3,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035496","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}
Pub Date : 2026-01-22DOI: 10.1016/j.anucene.2026.112143
Qifan Wang , Minyun Liu , Yanping Huang , Shanfang Huang , Ruohan Zheng , Houjun Gong
This study employs the lattice Boltzmann method to investigate the dynamics and heat transfer of carbon dioxide droplets on wall surfaces. Droplet spreading is shown to follow three distinct stages—rupture-to-rest, inertia–viscosity transition, and viscosity-dominated—each governed by power-law scaling of contact radius with time, with the rupture-to-rest stage unique to static droplets. Simulations across varying impact heights and wall inclinations demonstrate the universality of this scaling, extending classical spreading laws beyond flat surfaces. Droplet thickness evolution proceeds through accelerated shrinkage, decelerated shrinkage, and eventual stabilization. For inclined walls, the contact-line center emerges as a key dynamical reference, enabling systematic characterization of asymmetric spreading and sliding. Heat transfer analysis further identifies the cryogenic ring phenomenon as a dominant mechanism of wall conduction, while wall superheating and wettability are shown to strongly modulate thermal performance. These findings establish a unified framework for droplet morphology and heat transfer under realistic conditions.
{"title":"Heat transfer and dynamic behavior of wall-adhered carbon dioxide droplets: power-law correlation and the cryogenic ring effect","authors":"Qifan Wang , Minyun Liu , Yanping Huang , Shanfang Huang , Ruohan Zheng , Houjun Gong","doi":"10.1016/j.anucene.2026.112143","DOIUrl":"10.1016/j.anucene.2026.112143","url":null,"abstract":"<div><div>This study employs the lattice Boltzmann method to investigate the dynamics and heat transfer of carbon dioxide droplets on wall surfaces. Droplet spreading is shown to follow three distinct stages—rupture-to-rest, inertia–viscosity transition, and viscosity-dominated—each governed by power-law scaling of contact radius with time, with the rupture-to-rest stage unique to static droplets. Simulations across varying impact heights and wall inclinations demonstrate the universality of this scaling, extending classical spreading laws beyond flat surfaces. Droplet thickness evolution proceeds through accelerated shrinkage, decelerated shrinkage, and eventual stabilization. For inclined walls, the contact-line center emerges as a key dynamical reference, enabling systematic characterization of asymmetric spreading and sliding. Heat transfer analysis further identifies the cryogenic ring phenomenon as a dominant mechanism of wall conduction, while wall superheating and wettability are shown to strongly modulate thermal performance. These findings establish a unified framework for droplet morphology and heat transfer under realistic conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112143"},"PeriodicalIF":2.3,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035534","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}
Pub Date : 2026-01-22DOI: 10.1016/j.anucene.2026.112146
Juan Xu , Junhua Shen , Hong Wang , Ying Meng , Yiming Gao
In order to accurately measure the melting point of uranium dioxide, a novel visual method was developed. Uranium dioxide was heated using a fiber laser, with temperature monitored by a coaxially aligned 2-color pyrometer-Ⅱ. The temperature and melting process of uranium dioxide were observed using a 2-color pyrometer-Ⅰ equipped with a CCD camera. The melting point was determined based on the abrupt changes in the temperature curve of uranium dioxide, as well as the visual observations of the melting process captured by the CCD camera. The melting point of uranium dioxide was measured at 2855.6 ℃, while the melting point of molybdenum (Mo) was recorded at 2609.2 ℃. These values closely align with those reported in the literature.
{"title":"A new visual method for measuring the melting point of uranium dioxide","authors":"Juan Xu , Junhua Shen , Hong Wang , Ying Meng , Yiming Gao","doi":"10.1016/j.anucene.2026.112146","DOIUrl":"10.1016/j.anucene.2026.112146","url":null,"abstract":"<div><div>In order to accurately measure the melting point of uranium dioxide, a novel visual method was developed. Uranium dioxide was heated using a fiber laser, with temperature monitored by a coaxially aligned 2-color pyrometer-Ⅱ. The temperature and melting process of uranium dioxide were observed using a 2-color pyrometer-Ⅰ equipped with a CCD camera. The melting point was determined based on the abrupt changes in the temperature curve of uranium dioxide, as well as the visual observations of the melting process captured by the CCD camera. The melting point of uranium dioxide was measured at 2855.6 ℃, while the melting point of molybdenum (Mo) was recorded at 2609.2 ℃. These values closely align with those reported in the literature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112146"},"PeriodicalIF":2.3,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035535","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}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号