To study the mass transfer characteristics of convection-enhanced oxygen transport in electrochemical oxygen pumps, a theoretical model of oxygen transport was developed based on reaction kinetics. Two-dimensional and three-dimensional mass transfer models were subsequently established using CFD software. The experimental validation showed that the model error was within 10 %. The oxygen transport process in the electrochemical oxygen pump can be broadly divided into two stages: the ion-transport-dominated mass transfer within the solid electrolyte and the diffusion-dominated mass transfer in the lead-bismuth. The oxygen enhancement rate and the minimum oxygen concentration during the oxygen reduction process were fitted to the Arrhenius equation, further confirming the accuracy of the theoretical model. Based on this validated model, flow fields were incorporated, and impeller stirring was employed to achieve convection-enhanced mass transfer, aligning closely with experimental conditions. The effects of impeller speed and temperature on oxygen transport in the lead-bismuth tank were evaluated using the multiphysics field model. The results show that convection enhancement can effectively enhance the mass transfer of electrochemical oxygen pumping in lead-bismuth without considering the limitations of solid electrolyte. By strengthening the convection effect in the lead-bismuth and the electrochemical oxygen pump, one can achieve a limited improvement in overall mass transfer efficiency.
{"title":"Study of oxygen control model based on electrochemical oxygen pumping in liquid lead-bismuth alloys","authors":"Zhenhua Sheng, Weihao Wu, Tengjun Geng, Shengfei Wang, Fang Liu, Huiping Zhu, Haicai Lyu, Wentao Guo, Zhangpeng Guo, Ruixian Liang, Fenglei Niu","doi":"10.1016/j.pnucene.2025.106206","DOIUrl":"10.1016/j.pnucene.2025.106206","url":null,"abstract":"<div><div>To study the mass transfer characteristics of convection-enhanced oxygen transport in electrochemical oxygen pumps, a theoretical model of oxygen transport was developed based on reaction kinetics. Two-dimensional and three-dimensional mass transfer models were subsequently established using CFD software. The experimental validation showed that the model error was within 10 %. The oxygen transport process in the electrochemical oxygen pump can be broadly divided into two stages: the ion-transport-dominated mass transfer within the solid electrolyte and the diffusion-dominated mass transfer in the lead-bismuth. The oxygen enhancement rate and the minimum oxygen concentration during the oxygen reduction process were fitted to the Arrhenius equation, further confirming the accuracy of the theoretical model. Based on this validated model, flow fields were incorporated, and impeller stirring was employed to achieve convection-enhanced mass transfer, aligning closely with experimental conditions. The effects of impeller speed and temperature on oxygen transport in the lead-bismuth tank were evaluated using the multiphysics field model. The results show that convection enhancement can effectively enhance the mass transfer of electrochemical oxygen pumping in lead-bismuth without considering the limitations of solid electrolyte. By strengthening the convection effect in the lead-bismuth and the electrochemical oxygen pump, one can achieve a limited improvement in overall mass transfer efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106206"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841092","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 steam generator, an essential component of a pressurized water reactor nuclear power plant, has level parameters that directly influence the safety and stability of the nuclear power unit. The nonlinearity of the steam generator, coupled with the ‘false level’ issue, complicates its level control. Therefore, research on steam generator level control is considered to be of great significance. Traditional level control in steam generators employs a fixed-parameter Proportional-Integral-Derivative (PID) scheme, a widely used control strategy, which can struggle to adapt to changes in operating conditions and system characteristics. By integrating the adaptive learning capability of the Backpropagation (BP) neural network into the PID framework, a BP-PID control scheme is developed, where the BP network dynamically adjusts the PID parameters in real-time based on system feedback, enabling adaptive adjustment to varying operating conditions and enhancing control effectiveness. The proposed control algorithm was tested on a hardware-in-the-loop simulation platform. Results indicate that the BP-PID control significantly outperforms the original fixed-parameter PID control, demonstrating its potential for engineering implementation.
{"title":"Adaptive BP-PID control for steam generator level using hardware-in-the-loop simulation","authors":"YuLong Wang, Zhejun Sun, Qi Zhang, Peiwei Sun, Xinyu Wei","doi":"10.1016/j.pnucene.2025.106203","DOIUrl":"10.1016/j.pnucene.2025.106203","url":null,"abstract":"<div><div>The steam generator, an essential component of a pressurized water reactor nuclear power plant, has level parameters that directly influence the safety and stability of the nuclear power unit. The nonlinearity of the steam generator, coupled with the ‘false level’ issue, complicates its level control. Therefore, research on steam generator level control is considered to be of great significance. Traditional level control in steam generators employs a fixed-parameter Proportional-Integral-Derivative (PID) scheme, a widely used control strategy, which can struggle to adapt to changes in operating conditions and system characteristics. By integrating the adaptive learning capability of the Backpropagation (BP) neural network into the PID framework, a BP-PID control scheme is developed, where the BP network dynamically adjusts the PID parameters in real-time based on system feedback, enabling adaptive adjustment to varying operating conditions and enhancing control effectiveness. The proposed control algorithm was tested on a hardware-in-the-loop simulation platform. Results indicate that the BP-PID control significantly outperforms the original fixed-parameter PID control, demonstrating its potential for engineering implementation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106203"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841098","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 : 2026-03-01Epub Date: 2026-01-13DOI: 10.1016/j.pnucene.2025.106220
Mustafa J. Bolukbasi , Marat Margulis
Nuclear power is a cornerstone for achieving global sustainable energy goals, offering a low-carbon alternative to fossil fuels and promising significant contributions to energy security and climate change mitigation. However, the dual-use nature of nuclear technology presents inherent risks of nuclear proliferation, where materials and technologies intended for civilian energy production could be diverted to develop nuclear weapons. In this study, the impact of adding 241Am to the fuel composition on reactor operations for BWRs was analysed using CASMO-4/SIMULATE-3, focusing on achieving a proliferation-resistant fuel cycle. An examination was conducted on the breeding behaviours of plutonium atoms, as well as on power distribution and thermal behaviour. Analyses of neutronics and the fuel cycle have demonstrated that the addition of 241Am can facilitate the attainment of the necessary 238Pu ratio for a proliferation-resistant fuel cycle in BWR operations. Despite the requirement to moderately increase uranium enrichment to sustain cycle duration, it was observed that this adjustment does not significantly alter the power or thermal profiles.
{"title":"Impact of Am-241 on proliferation-resistant fuel cycle in boiling water reactors","authors":"Mustafa J. Bolukbasi , Marat Margulis","doi":"10.1016/j.pnucene.2025.106220","DOIUrl":"10.1016/j.pnucene.2025.106220","url":null,"abstract":"<div><div>Nuclear power is a cornerstone for achieving global sustainable energy goals, offering a low-carbon alternative to fossil fuels and promising significant contributions to energy security and climate change mitigation. However, the dual-use nature of nuclear technology presents inherent risks of nuclear proliferation, where materials and technologies intended for civilian energy production could be diverted to develop nuclear weapons. In this study, the impact of adding <sup>241</sup>Am to the fuel composition on reactor operations for BWRs was analysed using CASMO-4/SIMULATE-3, focusing on achieving a proliferation-resistant fuel cycle. An examination was conducted on the breeding behaviours of plutonium atoms, as well as on power distribution and thermal behaviour. Analyses of neutronics and the fuel cycle have demonstrated that the addition of <sup>241</sup>Am can facilitate the attainment of the necessary <sup>238</sup>Pu ratio for a proliferation-resistant fuel cycle in BWR operations. Despite the requirement to moderately increase uranium enrichment to sustain cycle duration, it was observed that this adjustment does not significantly alter the power or thermal profiles.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106220"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977697","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 : 2026-03-01Epub Date: 2026-01-07DOI: 10.1016/j.pnucene.2025.106234
D. Timpano , A. Vasiliev , D. Rochman , M. Hursin
The European project Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT (EVEREST) was launched in 2024 with the goal to employ advanced multiphysics tools to contribute to Long Term Operation (LTO) of nuclear power plants (NPP). The improvement of the reactor pressure vessel (RPV) fluence calculations is one of the overarching aims of this research endeavor, alongside a quantification of the modeling biases and uncertainties related to the standard computational methods. In the framework of the EVEREST project, EPFL has set up a new methodology for vessel fluence calculation based on Polaris/PARCS/Serpent and applied it to the Turkey Point 3 Pressurized Water Reactor (PWR), leveraging on public core design and operating data. This paper aims at presenting the results of the sensitivity analysis and uncertainty quantification performed for this application case. A novel integrated workflow to propagate uncertainties in the whole computational chain of vessel fluence calculations is described. The analysis covers core follow modeling, source preparation and shielding calculations, tackling the uncertainty due to material densities, geometry and nuclear data. Both standard Sandwich Formula (SF) and Total Monte-Carlo (TMC) techniques have been used for this scope. The UQ performed in this study revealed an analytical uncertainty associated with the fast flux at the thermal shield and the reactor pressure vessel location between 8 and 10%. The main contributors are the uncertainties on nuclear data and manufacturing tolerances in shielding calculations. This work contributes to the implementation of the Best Estimate Plus Uncertainty (BEPU) approach for assessing radiation damage in nuclear reactor structural materials.
{"title":"Uncertainty quantification for a vessel fluence calculation: A PWR case study","authors":"D. Timpano , A. Vasiliev , D. Rochman , M. Hursin","doi":"10.1016/j.pnucene.2025.106234","DOIUrl":"10.1016/j.pnucene.2025.106234","url":null,"abstract":"<div><div>The European project Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT (EVEREST) was launched in 2024 with the goal to employ advanced multiphysics tools to contribute to Long Term Operation (LTO) of nuclear power plants (NPP). The improvement of the reactor pressure vessel (RPV) fluence calculations is one of the overarching aims of this research endeavor, alongside a quantification of the modeling biases and uncertainties related to the standard computational methods. In the framework of the EVEREST project, EPFL has set up a new methodology for vessel fluence calculation based on Polaris/PARCS/Serpent and applied it to the Turkey Point 3 Pressurized Water Reactor (PWR), leveraging on public core design and operating data. This paper aims at presenting the results of the sensitivity analysis and uncertainty quantification performed for this application case. A novel integrated workflow to propagate uncertainties in the whole computational chain of vessel fluence calculations is described. The analysis covers core follow modeling, source preparation and shielding calculations, tackling the uncertainty due to material densities, geometry and nuclear data. Both standard Sandwich Formula (SF) and Total Monte-Carlo (TMC) techniques have been used for this scope. The UQ performed in this study revealed an analytical uncertainty associated with the fast flux at the thermal shield and the reactor pressure vessel location between 8 and 10%. The main contributors are the uncertainties on nuclear data and manufacturing tolerances in shielding calculations. This work contributes to the implementation of the Best Estimate Plus Uncertainty (BEPU) approach for assessing radiation damage in nuclear reactor structural materials.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106234"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927143","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 : 2026-03-01Epub Date: 2026-01-02DOI: 10.1016/j.pnucene.2025.106221
Hsingtzu Wu , Lihui Cai , Leyao Huang , Dawei Wang
Public acceptance of nuclear power is a major challenge for its development in China. This report concludes a study series analyzing factors that influence young Chinese people's attitude toward nuclear energy. In this report, trends in public attitudes are analyzed using data from three nationwide online surveys conducted via the Tencent platform in 2019, 2021, and 2023 (total N = 2234). These surveys employed quota sampling by respondents' current place of residence and targeted young Chinese adults aged 18–39. The results suggest that young Chinese people became more negative about nuclear power, interpreted as a side effect of fierce and widespread opposition to the release of treated radioactive water from Fukushima, which drew public attention to radioactive waste. Respondents also showed different interest in visiting Fukushima and Chernobyl, influenced not only by the severity of the accidents but also by media coverage. The surveys suggest that more than half of respondents expressed willingness to hang out with someone from a region affected by a severe nuclear accident. Furthermore, the COVID-19 pandemic has heightened health awareness, which may contribute to increased concern about the health impact of nuclear power plants and radiation. These changes are considered to exacerbate NIMBY syndrome. Those with a negative impression of nuclear power tend to have more serious NIMBY syndrome, while the growth of the NIMBY factor among the group of non-negative impression is remarkable. Finally, limitations and practical implications are discussed.
{"title":"Analysis of changing in young Chinese people's attitude toward nuclear energy using surveys conducted in 2019, 2021 and 2023","authors":"Hsingtzu Wu , Lihui Cai , Leyao Huang , Dawei Wang","doi":"10.1016/j.pnucene.2025.106221","DOIUrl":"10.1016/j.pnucene.2025.106221","url":null,"abstract":"<div><div>Public acceptance of nuclear power is a major challenge for its development in China. This report concludes a study series analyzing factors that influence young Chinese people's attitude toward nuclear energy. In this report, trends in public attitudes are analyzed using data from three nationwide online surveys conducted via the Tencent platform in 2019, 2021, and 2023 (total N = 2234). These surveys employed quota sampling by respondents' current place of residence and targeted young Chinese adults aged 18–39. The results suggest that young Chinese people became more negative about nuclear power, interpreted as a side effect of fierce and widespread opposition to the release of treated radioactive water from Fukushima, which drew public attention to radioactive waste. Respondents also showed different interest in visiting Fukushima and Chernobyl, influenced not only by the severity of the accidents but also by media coverage. The surveys suggest that more than half of respondents expressed willingness to hang out with someone from a region affected by a severe nuclear accident. Furthermore, the COVID-19 pandemic has heightened health awareness, which may contribute to increased concern about the health impact of nuclear power plants and radiation. These changes are considered to exacerbate NIMBY syndrome. Those with a negative impression of nuclear power tend to have more serious NIMBY syndrome, while the growth of the NIMBY factor among the group of non-negative impression is remarkable. Finally, limitations and practical implications are discussed.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106221"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884614","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 : 2026-03-01Epub Date: 2026-01-02DOI: 10.1016/j.pnucene.2025.106225
Junlin Chen , Wenhai Du , Xi Wang , Keyong Cheng , Xunfeng Li , Dongjiang Han , Xiulan Huai , Pengfei Lv , Hongsheng Dong
Printed circuit heat exchanger has great potential to replace traditional steam generators in molten salt-cooled high temperature reactor. In this study, an improved asymmetric-layout printed circuit steam generator (PCSG) is presented, where solar salt serves as the hot-side working fluid and water as the cold-side one. To accurately simulate phase-change heat transfer process of water (encompassing subcooled liquid, saturated liquid-steam, and superheated steam stages), a segmented Log-Mean Temperature Difference method integrated with Kandlikar's boiling heat transfer correlations is adopted. Mechanical integrity of the PCSG is verified in accordance with the American Society of Mechanical Engineers standards to ensure operational safety. Parametric analysis reveals that increasing the solar salt inlet velocity or expanding the water-side channel diameter enhances the PCSG's compactness but concurrently leads to higher pressure drops on both sides. Multi-objective optimization is performed using the genetic algorithm, followed by decision-making method, considering both two-objective (hot-side pressure drop and volume) and three-objective (hot-side pressure drop, cold-side pressure drop, and volume) scenarios. Compared to conventional shell-and-tube steam generators, optimized PCSG reduce height to 44.8–49.4 % and cold-side surface heat flux to 16.3–26.8 %, thus avoiding dryout in water or steam channels. This suggests that PCSG represent a promising alternative to conventional shell-and-tube steam generators. This study provides an efficient design approach for compact and safe PCSG in molten salt-cooled high-temperature reactors systems.
{"title":"Thermal design of asymmetrical layout in printed circuit steam generators for molten salt-cooled high temperature reactor","authors":"Junlin Chen , Wenhai Du , Xi Wang , Keyong Cheng , Xunfeng Li , Dongjiang Han , Xiulan Huai , Pengfei Lv , Hongsheng Dong","doi":"10.1016/j.pnucene.2025.106225","DOIUrl":"10.1016/j.pnucene.2025.106225","url":null,"abstract":"<div><div>Printed circuit heat exchanger has great potential to replace traditional steam generators in molten salt-cooled high temperature reactor. In this study, an improved asymmetric-layout printed circuit steam generator (PCSG) is presented, where solar salt serves as the hot-side working fluid and water as the cold-side one. To accurately simulate phase-change heat transfer process of water (encompassing subcooled liquid, saturated liquid-steam, and superheated steam stages), a segmented Log-Mean Temperature Difference method integrated with Kandlikar's boiling heat transfer correlations is adopted. Mechanical integrity of the PCSG is verified in accordance with the American Society of Mechanical Engineers standards to ensure operational safety. Parametric analysis reveals that increasing the solar salt inlet velocity or expanding the water-side channel diameter enhances the PCSG's compactness but concurrently leads to higher pressure drops on both sides. Multi-objective optimization is performed using the genetic algorithm, followed by decision-making method, considering both two-objective (hot-side pressure drop and volume) and three-objective (hot-side pressure drop, cold-side pressure drop, and volume) scenarios. Compared to conventional shell-and-tube steam generators, optimized PCSG reduce height to 44.8–49.4 % and cold-side surface heat flux to 16.3–26.8 %, thus avoiding dryout in water or steam channels. This suggests that PCSG represent a promising alternative to conventional shell-and-tube steam generators. This study provides an efficient design approach for compact and safe PCSG in molten salt-cooled high-temperature reactors systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106225"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884670","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 : 2026-03-01Epub Date: 2026-01-02DOI: 10.1016/j.pnucene.2025.106216
Wang-tao Xu , Jun-rui Liu , Yu-hao Zhou , Long-xiang Zhu , Lu-teng Zhang , Wen-xiong Zhou , Gang-yang Liu , Xu-bin Tan , Qing-che He , Liang-ming Pan
As flow pattern and its transition model are indispensable closure terms for two-phase flow modeling, accurate flow pattern definition and transition criteria are significant in thermal-hydraulic calculations. In terms of these motivation, the air-water two-phase co-current upward flow experiment has been conducted in 10, 20 and 30 mm ID vertical circular channels with annular impedance meters for void fraction measurement. The superficial velocities of water and air range from 0.25 to 4.0 m/s and 0.008 to 25.5 m/s covering bubbly to annular flow, respectively. The random forest algorithm is employed to identify the flow patterns and flow regime map are obtained. Compared with flow pattern data in published researches, the difference of flow regime map between different ID pipes is obvious, and the effect of axial development and pipe size on flow characteristics is evaluated. The result indicates that, the pressure drop and flow structure development caused by bubble expansion result in the local void fraction value and distribution, thus causing the flow pattern transition along the flow direction. Pipe size exerts a pronounced effect on the transitions from bubbly-to-slug and slug-to-churn flow. When jf remains constant, as the pipe diameter increases, a larger jg is required for flow pattern transition. However, its effect on the churn-to-annular transition is minimal and exhibits no obvious variation. Established flow pattern transition models show notable limitations. Published bubbly-to-slug models are unable to explain the obvious difference between 10 mm ID pipe and other size pipes, while the criterion of the constant jg at high flow rates for annular flow appears to be unreasonable and poor agreement with the data. The bubbly-to-slug flow transition in the 10 mm ID pipe occurs at a lower superficial gas velocity and void fraction than in the 20 mm and 30 mm ID pipes, which emphasizes the effect of pipe size limitation with bubble interaction mechanism. The new flow pattern transition model for small-diameter circular pipes is proposed with pipe size as the primary factor.
{"title":"Experimental research on flow characteristics and flow pattern transition model of co-current upward air-water flow in small-diameter circular pipes","authors":"Wang-tao Xu , Jun-rui Liu , Yu-hao Zhou , Long-xiang Zhu , Lu-teng Zhang , Wen-xiong Zhou , Gang-yang Liu , Xu-bin Tan , Qing-che He , Liang-ming Pan","doi":"10.1016/j.pnucene.2025.106216","DOIUrl":"10.1016/j.pnucene.2025.106216","url":null,"abstract":"<div><div>As flow pattern and its transition model are indispensable closure terms for two-phase flow modeling, accurate flow pattern definition and transition criteria are significant in thermal-hydraulic calculations. In terms of these motivation, the air-water two-phase co-current upward flow experiment has been conducted in 10, 20 and 30 mm ID vertical circular channels with annular impedance meters for void fraction measurement. The superficial velocities of water and air range from 0.25 to 4.0 m/s and 0.008 to 25.5 m/s covering bubbly to annular flow, respectively. The random forest algorithm is employed to identify the flow patterns and flow regime map are obtained. Compared with flow pattern data in published researches, the difference of flow regime map between different ID pipes is obvious, and the effect of axial development and pipe size on flow characteristics is evaluated. The result indicates that, the pressure drop and flow structure development caused by bubble expansion result in the local void fraction value and distribution, thus causing the flow pattern transition along the flow direction. Pipe size exerts a pronounced effect on the transitions from bubbly-to-slug and slug-to-churn flow. When <em>j</em><sub>f</sub> remains constant, as the pipe diameter increases, a larger <em>j</em><sub>g</sub> is required for flow pattern transition. However, its effect on the churn-to-annular transition is minimal and exhibits no obvious variation. Established flow pattern transition models show notable limitations. Published bubbly-to-slug models are unable to explain the obvious difference between 10 mm ID pipe and other size pipes, while the criterion of the constant <em>j</em><sub>g</sub> at high flow rates for annular flow appears to be unreasonable and poor agreement with the data. The bubbly-to-slug flow transition in the 10 mm ID pipe occurs at a lower superficial gas velocity and void fraction than in the 20 mm and 30 mm ID pipes, which emphasizes the effect of pipe size limitation with bubble interaction mechanism. The new flow pattern transition model for small-diameter circular pipes is proposed with pipe size as the primary factor.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106216"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884779","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 : 2026-03-01Epub Date: 2026-01-05DOI: 10.1016/j.pnucene.2025.106226
You Wang , Zhangjian Zhou , Junqiang Lu , Lingzhi Chen , Haodong Jia , Zhenfeng Tong , Carsten Schroer , Yang Liu , Kai Chen , Zhao Shen , Xiaoqin Zeng
The corrosion behavior of Fe-12Cr-1Al oxide dispersion strengthened (ODS) alloys, with and without 2 wt% Si addition, was examined in oxygen-controlled (10−6 wt%) liquid lead–bismuth eutectic (LBE) at 650 °C for 2000 h. Both alloys developed protective surface scales that suppressed dissolution attack, covering nearly the entire exposed surface. For the Si-containing alloy, the oxide scale was thinner and more compact, accompanied by the formation of a discontinuous SiO2 layer at the interface between the outer and inner oxides. This interfacial SiO2 impeded the outward diffusion of Fe and Cr while facilitating Al outward transport, leading to denser (Fe,Al)3O4 spinels in the outermost layer and a reduced defect density within the Al2O3-rich inner layer. A dissolution–oxidation–redeposition mechanism was proposed to rationalize the evolution of the multi-layered scales. These findings demonstrate that minor Si additions enhance the compactness and protectiveness of oxide scales on ODS alloys, offering guidance for the design of corrosion-resistant structural materials for LBE-cooled systems.
{"title":"Effect of Si addition on the oxidation of a 12Cr1Al ODS alloy in liquid LBE","authors":"You Wang , Zhangjian Zhou , Junqiang Lu , Lingzhi Chen , Haodong Jia , Zhenfeng Tong , Carsten Schroer , Yang Liu , Kai Chen , Zhao Shen , Xiaoqin Zeng","doi":"10.1016/j.pnucene.2025.106226","DOIUrl":"10.1016/j.pnucene.2025.106226","url":null,"abstract":"<div><div>The corrosion behavior of Fe-12Cr-1Al oxide dispersion strengthened (ODS) alloys, with and without 2 wt% Si addition, was examined in oxygen-controlled (10<sup>−6</sup> wt%) liquid lead–bismuth eutectic (LBE) at 650 °C for 2000 h. Both alloys developed protective surface scales that suppressed dissolution attack, covering nearly the entire exposed surface. For the Si-containing alloy, the oxide scale was thinner and more compact, accompanied by the formation of a discontinuous SiO<sub>2</sub> layer at the interface between the outer and inner oxides. This interfacial SiO<sub>2</sub> impeded the outward diffusion of Fe and Cr while facilitating Al outward transport, leading to denser (Fe,Al)<sub>3</sub>O<sub>4</sub> spinels in the outermost layer and a reduced defect density within the Al<sub>2</sub>O<sub>3</sub>-rich inner layer. A dissolution–oxidation–redeposition mechanism was proposed to rationalize the evolution of the multi-layered scales. These findings demonstrate that minor Si additions enhance the compactness and protectiveness of oxide scales on ODS alloys, offering guidance for the design of corrosion-resistant structural materials for LBE-cooled systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106226"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927079","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 : 2026-03-01Epub Date: 2025-12-27DOI: 10.1016/j.pnucene.2025.106210
Weihua Cai , Jian Zhang , Zequan Huang , Desheng Jin , Jianchuang Sun , Gongqing Wang , Wenchao Zhang
Fuel rods can enhance the heat exchange capacity of small reactors. In this paper, four new arrangements are designed to enhance heat transfer by changing the spiral direction of the ribs around petal-shaped fuel rods. Based on numerical simulations, the flow and heat transfer characteristics of petal-shaped fuel assemblies under different arrangements are compared, and the mechanism of the heat transfer enhanced by the configuration with the strongest comprehensive heat transfer performance is deduced. The study finds that the pressure drops of the four new arrangements are almost the same as those of the original arrangement. Based on the comprehensive performance evaluation index, it is determined that the fuel rod spacer arrangement with clockwise and counterclockwise twist in the diagonal direction has the best performance among the four arrangements. A further comparison of the heat transfer performance within each sub-channel of this staggered arrangement and the original arrangement reveals that the heat transfer coefficients within the four types of sub-channels are significantly enhanced. This is because the fuel rod spacer arrangement with clockwise and counterclockwise twist in the diagonal direction can significantly enhance the coolant flow on the leeward side of fuel rods, thereby reducing the thickness of velocity boundary layer and thermal boundary layer, strengthening the mixing of hot and cold fluids, and thus greatly improving the convective heat transfer performance in the leeward area. The relevant research results have laid a theoretical foundation for the arrangement scheme of the petal-shaped fuel rods within the reactor.
{"title":"Numerical study on the optimization of petal-shaped fuel rod arrangement and heat transfer characteristics","authors":"Weihua Cai , Jian Zhang , Zequan Huang , Desheng Jin , Jianchuang Sun , Gongqing Wang , Wenchao Zhang","doi":"10.1016/j.pnucene.2025.106210","DOIUrl":"10.1016/j.pnucene.2025.106210","url":null,"abstract":"<div><div>Fuel rods can enhance the heat exchange capacity of small reactors. In this paper, four new arrangements are designed to enhance heat transfer by changing the spiral direction of the ribs around petal-shaped fuel rods. Based on numerical simulations, the flow and heat transfer characteristics of petal-shaped fuel assemblies under different arrangements are compared, and the mechanism of the heat transfer enhanced by the configuration with the strongest comprehensive heat transfer performance is deduced. The study finds that the pressure drops of the four new arrangements are almost the same as those of the original arrangement. Based on the comprehensive performance evaluation index, it is determined that the fuel rod spacer arrangement with clockwise and counterclockwise twist in the diagonal direction has the best performance among the four arrangements. A further comparison of the heat transfer performance within each sub-channel of this staggered arrangement and the original arrangement reveals that the heat transfer coefficients within the four types of sub-channels are significantly enhanced. This is because the fuel rod spacer arrangement with clockwise and counterclockwise twist in the diagonal direction can significantly enhance the coolant flow on the leeward side of fuel rods, thereby reducing the thickness of velocity boundary layer and thermal boundary layer, strengthening the mixing of hot and cold fluids, and thus greatly improving the convective heat transfer performance in the leeward area. The relevant research results have laid a theoretical foundation for the arrangement scheme of the petal-shaped fuel rods within the reactor.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106210"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841093","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 : 2026-03-01Epub Date: 2025-12-25DOI: 10.1016/j.pnucene.2025.106207
Yuanlin Yao , Qianlong Zuo , Penghui Zhang , Jian Deng , Dan Wu , Jiayue Zhou , Deqi Chen
Reflooding serves as an essential emergency cooling strategy following reactor core exposure due to a loss of coolant accident. The compact architecture of advanced nuclear core, in contrast to conventional fuel rods, elevates the risk of localized heat flux peaks and imposes more stringent demands on reflooding heat transfer efficiency. This work experimentally investigates heat transfer enhancement methods during reflooding in narrow rectangular channels. The reflooding characteristic was firstly analyzed, then the effects of inlet parameters, nanofluids, and microstructured surfaces on the minimum film boiling temperature and cooling time were quantitatively evaluated. Experimental results demonstrate that increasing the coolant inlet velocity and subcooling significantly improve the minimum film boiling temperature and reduce the cooling time. In contrast, initial wall temperature and input heat flux exhibit negligible influence on minimum film boiling temperature but are positively correlated with cooling time. Nanofluid (Fe3O4 nanoparticles <1 wt%) exhibit negligible heat transfer enhancement due to the vapor film preventing their surface deposition. The microstructured surface achieved the minimum film boiling temperature improvements of 4.8 %–21.1 % relative to the flat surface across the tested inlet velocities and 6.5 %–12.6 % under varying subcooling conditions, while simultaneously reducing the cooling time by 1.8 %–2.7 %.
{"title":"Experimental study on two-phase heat transfer enhancement in narrow rectangular channels during reflooding process","authors":"Yuanlin Yao , Qianlong Zuo , Penghui Zhang , Jian Deng , Dan Wu , Jiayue Zhou , Deqi Chen","doi":"10.1016/j.pnucene.2025.106207","DOIUrl":"10.1016/j.pnucene.2025.106207","url":null,"abstract":"<div><div>Reflooding serves as an essential emergency cooling strategy following reactor core exposure due to a loss of coolant accident. The compact architecture of advanced nuclear core, in contrast to conventional fuel rods, elevates the risk of localized heat flux peaks and imposes more stringent demands on reflooding heat transfer efficiency. This work experimentally investigates heat transfer enhancement methods during reflooding in narrow rectangular channels. The reflooding characteristic was firstly analyzed, then the effects of inlet parameters, nanofluids, and microstructured surfaces on the minimum film boiling temperature and cooling time were quantitatively evaluated. Experimental results demonstrate that increasing the coolant inlet velocity and subcooling significantly improve the minimum film boiling temperature and reduce the cooling time. In contrast, initial wall temperature and input heat flux exhibit negligible influence on minimum film boiling temperature but are positively correlated with cooling time. Nanofluid (Fe<sub>3</sub>O<sub>4</sub> nanoparticles <1 wt%) exhibit negligible heat transfer enhancement due to the vapor film preventing their surface deposition. The microstructured surface achieved the minimum film boiling temperature improvements of 4.8 %–21.1 % relative to the flat surface across the tested inlet velocities and 6.5 %–12.6 % under varying subcooling conditions, while simultaneously reducing the cooling time by 1.8 %–2.7 %.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106207"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841095","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}
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