Pub Date : 2026-06-01Epub Date: 2026-01-30DOI: 10.1016/j.anucene.2026.112173
Huidong Shi , Taikun Guo , Yu Hu , Jie Feng , Ruifeng Tian , Jiming Wen
In the operation of nuclear power systems, leakage incidents in condensers can cause significant threats to safe operation. The intrusion of seawater into the secondary loop, containing chloride ions (Cl-) and other insoluble impurities, may corrode heat transfer tubes and lead to severe consequences. This study focuses on investigating the migration and precipitation of Cl- and other insoluble impurities in the secondary side of the steam generator. The heat transfer and flow characteristics were numerically simulated, and the distributions of secondary side temperature, void fraction, and velocity were analysed. On this basis, the study investigated the migration of Cl-, their precipitation induced by evaporation and crystallization processes, as well as the migration and precipitation of insoluble impurities. The results demonstrate that the secondary side inlet flow velocity, temperature, and primary side flow velocity all exhibit significant influences on chloride ion concentration. Cl- accumulates on the hot side of the conical expansion support plate, reaching a peak concentration of 30.3 parts per million (ppm). The insoluble impurities are primarily deposited on the flow distribution plate.
{"title":"Investigation on ion migration in the secondary side of steam generators based on porous medium method","authors":"Huidong Shi , Taikun Guo , Yu Hu , Jie Feng , Ruifeng Tian , Jiming Wen","doi":"10.1016/j.anucene.2026.112173","DOIUrl":"10.1016/j.anucene.2026.112173","url":null,"abstract":"<div><div>In the operation of nuclear power systems, leakage incidents in condensers can cause significant threats to safe operation. The intrusion of seawater into the secondary loop, containing chloride ions (Cl-) and other insoluble impurities, may corrode heat transfer tubes and lead to severe consequences. This study focuses on investigating the migration and precipitation of Cl- and other insoluble impurities in the secondary side of the steam generator. The heat transfer and flow characteristics were numerically simulated, and the distributions of secondary side temperature, void fraction, and velocity were analysed. On this basis, the study investigated the migration of Cl-, their precipitation induced by evaporation and crystallization processes, as well as the migration and precipitation of insoluble impurities. The results demonstrate that the secondary side inlet flow velocity, temperature, and primary side flow velocity all exhibit significant influences on chloride ion concentration. Cl- accumulates on the hot side of the conical expansion support plate, reaching a peak concentration of 30.3 parts per million (ppm). The insoluble impurities are primarily deposited on the flow distribution plate.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112173"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074818","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 : 2026-06-01Epub Date: 2026-01-14DOI: 10.1016/j.anucene.2025.112107
Cristian G. de Oliveira , Alirio J.S. Piña , Antonella L. Costa , Claubia Pereira , Clarysson A.M. Silva , Damian E. Ramajo , Dario M. Godino
Small Modular Reactors (SMRs) are increasingly pursued to provide reliable low-carbon power, with Pressurized Water Reactors (PWRs) offering particular advantages due to extensive operational experience, well-characterized coolant properties, and an established industrial infrastructure. In this context, the present study develops a detailed conjugate heat-transfer (CHT) model of the hottest fuel assembly in a PWR-type SMR using the open-source CFD toolbox OpenFOAM. Exploiting geometric symmetry, the fuel and coolant regions were resolved with finite volumes, while the thin helium gap and cladding layers were modeled as lumped thermal resistances. A four-case mesh sensitivity study confirmed mesh-independent predictions of the average outlet coolant temperature and quantified the influence of spatial refinement on hydraulic-loss estimation. Steady-state CHT simulations provided axial and radial temperature distributions in both the fuel and coolant, together with detailed velocity and pressure fields. The results captured key physical features, including spacer-grid-induced velocity reductions and the coupling between coolant heating, density decrease, and local flow acceleration. The predicted average outlet coolant temperature (607 K) was consistent with reference operating conditions, while the total pressure drop across the assembly was found to be on the order of 20 kPa, reflecting the combined effect of frictional losses and spacer-grid resistance. A transient Loss-of-Flow Accident (LOFA) was also simulated by imposing a 50% reduction in the inlet mass flow rate. The analysis revealed oscillations in the outlet temperature driven by variable-density effects, with an average increase of approximately 15 K relative to steady-state conditions. Fuel centerline temperatures increased accordingly but remained safely below the UO melting point. Overall, the proposed methodology demonstrates the capability of open-source CFD tools to predict coupled neutronic–thermal–hydraulic behavior at the fuel-assembly scale in PWR-type SMRs, providing a solid foundation for future extensions to multiphase modeling and beyond-design-basis accident scenarios.
{"title":"Development of a local power peaking analysis methodology using OpenFOAM for a pressurized water-cooled small modular reactor","authors":"Cristian G. de Oliveira , Alirio J.S. Piña , Antonella L. Costa , Claubia Pereira , Clarysson A.M. Silva , Damian E. Ramajo , Dario M. Godino","doi":"10.1016/j.anucene.2025.112107","DOIUrl":"10.1016/j.anucene.2025.112107","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) are increasingly pursued to provide reliable low-carbon power, with Pressurized Water Reactors (PWRs) offering particular advantages due to extensive operational experience, well-characterized coolant properties, and an established industrial infrastructure. In this context, the present study develops a detailed conjugate heat-transfer (CHT) model of the hottest fuel assembly in a PWR-type SMR using the open-source CFD toolbox OpenFOAM. Exploiting geometric symmetry, the fuel and coolant regions were resolved with finite volumes, while the thin helium gap and cladding layers were modeled as lumped thermal resistances. A four-case mesh sensitivity study confirmed mesh-independent predictions of the average outlet coolant temperature and quantified the influence of spatial refinement on hydraulic-loss estimation. Steady-state CHT simulations provided axial and radial temperature distributions in both the fuel and coolant, together with detailed velocity and pressure fields. The results captured key physical features, including spacer-grid-induced velocity reductions and the coupling between coolant heating, density decrease, and local flow acceleration. The predicted average outlet coolant temperature (607 K) was consistent with reference operating conditions, while the total pressure drop across the assembly was found to be on the order of 20 kPa, reflecting the combined effect of frictional losses and spacer-grid resistance. A transient Loss-of-Flow Accident (LOFA) was also simulated by imposing a 50% reduction in the inlet mass flow rate. The analysis revealed oscillations in the outlet temperature driven by variable-density effects, with an average increase of approximately 15 K relative to steady-state conditions. Fuel centerline temperatures increased accordingly but remained safely below the UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> melting point. Overall, the proposed methodology demonstrates the capability of open-source CFD tools to predict coupled neutronic–thermal–hydraulic behavior at the fuel-assembly scale in PWR-type SMRs, providing a solid foundation for future extensions to multiphase modeling and beyond-design-basis accident scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112107"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975067","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 : 2026-06-01Epub Date: 2026-02-03DOI: 10.1016/j.anucene.2026.112161
Shamsuddeen Lawal , Chenyang Wang , Minjun Peng
The global imperative for decarbonization has reaffirmed the nuclear energy’s role as a low-carbon baseload source, contingent on robust safety assurances. The shift toward Passive System System (PSS), which utilize natural phenomena like gravity and natural circulation enhances resilience but poses unique reliability challenges. Conventional Probabilistic Risk Assessment (PRA) inadequately models functional failure, where performance degrades due to uncertain physical phenomena despite all components operational, and struggles with dominant epistemic uncertainties in novel designs. This review synthesizes methodological advances, tracing the evaluation from computationally intensive first-generation framework (e.g., RMPS/ASPRA) to machine learning-driven paradigm integrating AI-based surrogate models (e.g., Kriging, Polynomial Chaos Expansion, Physics-Informed Neural Networks). These enable efficient quantification of functional failure probabilities, epistemic uncertainty mapping via Bayesian and adaptive sampling, and revelation of time-dependent risk pathways via Dynamic PRA (DPRA) invisible to static methods. However, the irreplaceable role of machine learning in addressing computational bottleneck introduces new issues, including “black-box” opacity, regulatory challeges for licensing, hybrid active–passive system integration, data scarcity for Gene III+, SMR, Gen-IV designs, and long-term material degradation effects. We conclude that PSS reliability hinges on Explainable AI (XAI) to demystify models, standardized validation protocol, integrated cyber-physical-security framework. This transformation, particularly through Physics-Informed Machine Learning tools like PINNs, is essential to generate the rigorous, regulatory-acceptance evidence needed for licensing and deploying advanced reactors.
{"title":"Reliability of passive safety system in nuclear power plants: advances, emerging technologies, and persistent challenges","authors":"Shamsuddeen Lawal , Chenyang Wang , Minjun Peng","doi":"10.1016/j.anucene.2026.112161","DOIUrl":"10.1016/j.anucene.2026.112161","url":null,"abstract":"<div><div>The global imperative for decarbonization has reaffirmed the nuclear energy’s role as a low-carbon baseload source, contingent on robust safety assurances. The shift toward Passive System System (PSS), which utilize natural phenomena like gravity and natural circulation enhances resilience but poses unique reliability challenges. Conventional Probabilistic Risk Assessment (PRA) inadequately models <em>functional failure,</em> where performance degrades due to uncertain physical phenomena despite all components operational, and struggles with dominant epistemic uncertainties in novel designs. This review synthesizes methodological advances, tracing the evaluation from computationally intensive first-generation framework (e.g., RMPS/ASPRA) to machine learning-driven paradigm integrating AI-based surrogate models (e.g., Kriging, Polynomial Chaos Expansion, Physics-Informed Neural Networks). These enable efficient quantification of functional failure probabilities, epistemic uncertainty mapping via Bayesian and adaptive sampling, and revelation of time-dependent risk pathways via Dynamic PRA (DPRA) invisible to static methods. However, the irreplaceable role of machine learning in addressing computational bottleneck introduces new issues, including “black-box” opacity, regulatory challeges for licensing, hybrid active–passive system integration, data scarcity for Gene III+, SMR, Gen-IV designs, and long-term material degradation effects. We conclude that PSS reliability hinges on Explainable AI (XAI) to demystify models, standardized validation protocol, integrated cyber-physical-security framework. This transformation, particularly through Physics-Informed Machine Learning tools like PINNs, is essential to generate the rigorous, regulatory-acceptance evidence needed for licensing and deploying advanced reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112161"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146098724","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 : 2026-06-01Epub Date: 2026-02-13DOI: 10.1016/j.anucene.2026.112205
Xiaoqin Du , Zhuojun Jiang , Wan Sun , Zhuhai Zhong , Yan Wei , Liangming Pan
The steam turbine is one of the most critical energy conversion components in power generation systems. Unlike thermal power plants, nuclear steam turbines operate in the wet steam region in most of their stages, with the inlet steam often near saturation, resulting in more severe non-equilibrium condensation. This study aims to investigate the non-equilibrium condensation flow characteristics and efficiency of a nuclear steam turbine. Firstly, a suitable condensation model was selected and validated by comparing it with the existing experimental data on nozzles. Then, a thermodynamic analysis of non-equilibrium condensation in a nine-stage nuclear turbine was performed using the Euler–Euler method. Due to the saturated inlet conditions and a low supercooling degree, nucleation was suppressed by pre-existing droplets, and condensation was dominated by droplet growth. Furthermore, the effects of steam extraction ports and shaft seals on condensation were examined. Results indicate that although extraction promotes nucleation, the extracted flow, accounting for approximately 15% of the inlet mass flow, removes part of the droplets, thereby reducing overall humidity. Finally, internal efficiency and various types of losses were analyzed. The turbine efficiency decreased from 91% to around 80%, with wet steam loss being the most significant, reaching about 15% at the last stage. The findings provide insights for improving the design and operation of nuclear steam turbines, enhancing their economic performance and operational reliability.
{"title":"Numerical study on the flow field characteristics and efficiency losses in nuclear power turbines based on the non-equilibrium condensation model","authors":"Xiaoqin Du , Zhuojun Jiang , Wan Sun , Zhuhai Zhong , Yan Wei , Liangming Pan","doi":"10.1016/j.anucene.2026.112205","DOIUrl":"10.1016/j.anucene.2026.112205","url":null,"abstract":"<div><div>The steam turbine is one of the most critical energy conversion components in power generation systems. Unlike thermal power plants, nuclear steam turbines operate in the wet steam region in most of their stages, with the inlet steam often near saturation, resulting in more severe non-equilibrium condensation. This study aims to investigate the non-equilibrium condensation flow characteristics and efficiency of a nuclear steam turbine. Firstly, a suitable condensation model was selected and validated by comparing it with the existing experimental data on nozzles. Then, a thermodynamic analysis of non-equilibrium condensation in a nine-stage nuclear turbine was performed using the Euler–Euler method. Due to the saturated inlet conditions and a low supercooling degree, nucleation was suppressed by pre-existing droplets, and condensation was dominated by droplet growth. Furthermore, the effects of steam extraction ports and shaft seals on condensation were examined. Results indicate that although extraction promotes nucleation, the extracted flow, accounting for approximately 15% of the inlet mass flow, removes part of the droplets, thereby reducing overall humidity. Finally, internal efficiency and various types of losses were analyzed. The turbine efficiency decreased from 91% to around 80%, with wet steam loss being the most significant, reaching about 15% at the last stage. The findings provide insights for improving the design and operation of nuclear steam turbines, enhancing their economic performance and operational reliability.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112205"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186618","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 : 2026-06-01Epub Date: 2026-02-13DOI: 10.1016/j.anucene.2026.112200
Ji Hwan Lim , Seung-Hwan Yu , Kyoung-Sik Bang , Gyung-sun Chae , Kyung-Wook Shin , Nam-Hee Lee
The objective of this study is to quantify and interpret residual-water evaporation/retention behavior in a simulated damaged spent-fuel cladding during vacuum drying as a function of key operational and geometric parameters. A lab-scale vacuum-drying facility was used to conduct a parametric investigation by varying vacuum pump suction capacity (100–600 L/min), defect diameter (0.3–2.0 mm), and initial water temperature (20–80 °C), and by evaluating water removal after the vacuum-drying criterion was satisfied. The results show that increasing suction capacity to 600 L/min reduced the time to meet the drying criterion by more than 50% compared with 100 L/min, while rapid cooling at high suction could hinder complete removal. Circular defect geometry strongly governed bulk discharge: a 2.0 mm defect achieved up to 75% removal, whereas a 0.3 mm defect yielded ∼ 15%. Higher initial water temperature (80 °C) did not improve removal as expected; instead, vapor entrapment associated with buoyancy reduced removal efficiency relative to 20 °C. Contour mapping indicated residual-water retention spanning 24.80–88.80% across the tested conditions, with defect size as the dominant factor. These findings provide experimentally grounded guidance for interpreting and optimizing vacuum-drying performance for damaged fuel configurations, including conditions relevant to high-burnup thermal loads.
{"title":"In-depth analysis of water evaporation in damaged spent nuclear fuel after vacuum drying","authors":"Ji Hwan Lim , Seung-Hwan Yu , Kyoung-Sik Bang , Gyung-sun Chae , Kyung-Wook Shin , Nam-Hee Lee","doi":"10.1016/j.anucene.2026.112200","DOIUrl":"10.1016/j.anucene.2026.112200","url":null,"abstract":"<div><div>The objective of this study is to quantify and interpret residual-water evaporation/retention behavior in a simulated damaged spent-fuel cladding during vacuum drying as a function of key operational and geometric parameters. A lab-scale vacuum-drying facility was used to conduct a parametric investigation by varying vacuum pump suction capacity (100–600 L/min), defect diameter (0.3–2.0 mm), and initial water temperature (20–80 °C), and by evaluating water removal after the vacuum-drying criterion was satisfied. The results show that increasing suction capacity to 600 L/min reduced the time to meet the drying criterion by more than 50% compared with 100 L/min, while rapid cooling at high suction could hinder complete removal. Circular defect geometry strongly governed bulk discharge: a 2.0 mm defect achieved up to 75% removal, whereas a 0.3 mm defect yielded ∼ 15%. Higher initial water temperature (80 °C) did not improve removal as expected; instead, vapor entrapment associated with buoyancy reduced removal efficiency relative to 20 °C. Contour mapping indicated residual-water retention spanning 24.80–88.80% across the tested conditions, with defect size as the dominant factor. These findings provide experimentally grounded guidance for interpreting and optimizing vacuum-drying performance for damaged fuel configurations, including conditions relevant to high-burnup thermal loads.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112200"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186619","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 : 2026-06-01Epub Date: 2026-02-10DOI: 10.1016/j.anucene.2026.112191
Adolphus Lye , Scott Ferson , Sicong Xiao
The paper proposes a stochastic model updating framework which possesses two key features: (1) the entropy-based Jensen–Shannon divergence as the distance function for the Approximate Bayesian Computation procedure, owing to its strengths in – (a) effectively capturing the discrepancy in both the relative mean and the variance information between two distributions, and (b) returning bounded finite values which avoids the issue of computational instability; and (2) the implementation of the Transitional Ensemble Markov Chain Monte Carlo to obtain posterior samples under affine-invariance. To the best of knowledge, the Jensen–Shannon divergence implementation for Approximate Bayesian Computation is under studied, providing an opportunity to study its robustness towards performing model calibration under varying data size. In addition, the challenge of performing a stochastic reliability analysis involving a system of coupled-equations is not widely investigated within the existing literature which further motivates such study. The proposed framework is validated through the 2008 SANDIA thermal problem involving a reactor slab material, whose thermal property reliability under a specific temperature condition is assessed. The robustness of the proposed framework is evaluated and verified against published results.
{"title":"A distribution-free stochastic physics-guided reliability analysis under polymorphic uncertainty","authors":"Adolphus Lye , Scott Ferson , Sicong Xiao","doi":"10.1016/j.anucene.2026.112191","DOIUrl":"10.1016/j.anucene.2026.112191","url":null,"abstract":"<div><div>The paper proposes a stochastic model updating framework which possesses two key features: (1) the entropy-based Jensen–Shannon divergence as the distance function for the Approximate Bayesian Computation procedure, owing to its strengths in – (a) effectively capturing the discrepancy in both the relative mean and the variance information between two distributions, and (b) returning bounded finite values which avoids the issue of computational instability; and (2) the implementation of the Transitional Ensemble Markov Chain Monte Carlo to obtain posterior samples under affine-invariance. To the best of knowledge, the Jensen–Shannon divergence implementation for Approximate Bayesian Computation is under studied, providing an opportunity to study its robustness towards performing model calibration under varying data size. In addition, the challenge of performing a stochastic reliability analysis involving a system of coupled-equations is not widely investigated within the existing literature which further motivates such study. The proposed framework is validated through the 2008 SANDIA thermal problem involving a reactor slab material, whose thermal property reliability under a specific temperature condition is assessed. The robustness of the proposed framework is evaluated and verified against published results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112191"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186622","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 : 2026-06-01Epub Date: 2026-02-11DOI: 10.1016/j.anucene.2026.112199
Ruiyang Zhong , Wenhai Qu , Jun Yang , Li Zhan
The wire-wrapped rod fuel assembly has advantage of self-supporting and heat transfer improvement. Turbulent flow in wire-wrapped rod bundle should be studied to reveal the helical wire effect on flow. However, the wrapped wires introduce laser refraction problem for visualization measurement. In this study, cross-flow fields in a 7 wire-wrapped rod bundle (15 mm in rod diameter, 4 mm in wire diameter and 150 mm in screw length) were studied by time-resolved particle image velocimetry (TR-PIV) assisted by high-precision matched index of refraction technology of acrylic and sodium iodide solution. Based on experimental data, time-averaged cross-flow structures and Reynolds stresses were analyzed. The helical wires spatially periodically produce shear flow and vortex, improve Reynolds stresses and generate turbulent eddies with relatively small cross correlation length scales in wire-wrapped rod bundle. The Reynolds number has no effect on turbulence in wire-wrapped rod bundle at 30000. The experimental data help to understanding the helical wire effect on turbulent flow in wire-wrapped rod bundle.
{"title":"Experimental study of cross-flow in a 7 wire-wrapped rod bundle by PIV","authors":"Ruiyang Zhong , Wenhai Qu , Jun Yang , Li Zhan","doi":"10.1016/j.anucene.2026.112199","DOIUrl":"10.1016/j.anucene.2026.112199","url":null,"abstract":"<div><div>The wire-wrapped rod fuel assembly has advantage of self-supporting and heat transfer improvement. Turbulent flow in wire-wrapped rod bundle should be studied to reveal the helical wire effect on flow. However, the wrapped wires introduce laser refraction problem for visualization measurement. In this study, cross-flow fields in a 7 wire-wrapped rod bundle (15 mm in rod diameter, 4 mm in wire diameter and 150 mm in screw length) were studied by time-resolved particle image velocimetry (TR-PIV) assisted by high-precision matched index of refraction technology of acrylic and sodium iodide solution. Based on experimental data, time-averaged cross-flow structures and Reynolds stresses were analyzed. The helical wires spatially periodically produce shear flow and vortex, improve Reynolds stresses and generate turbulent eddies with relatively small cross correlation length scales in wire-wrapped rod bundle. The Reynolds number has no effect on turbulence in wire-wrapped rod bundle at 30000. The experimental data help to understanding the helical wire effect on turbulent flow in wire-wrapped rod bundle.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112199"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186633","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 : 2026-06-01Epub Date: 2026-02-09DOI: 10.1016/j.anucene.2026.112183
Yi Ren , Ming Jin , Shize Ouyang , Chao Lian , Chaodong Zhang , Jieqiong Jiang , FDS Consortium
Steam Generator Tube Rupture (SGTR) accident is one of the critical safety accidents in lead-based reactors. In this study, a 200MWt small lead-based reactor was taken as the research object. A computational domain was established using a 1/2 geometric model of the primary loop system of the reactor. Combined with the Eulerian-Lagrangian CFD Discrete Phase Model and the drag coefficient correlation, the bubble migration characteristics were analyzed, and the law of bubble migration characteristics was obtained. Additionally, the possibility of bubbles entering the reactor core was evaluated. The results show that: for accidents with lower break location, the probability of bubbles entering the core is higher; besides, the smaller the bubble diameter, the higher the probability of entering the core. Based on the calculation results, a baffle-based mitigation measure was designed along the flow path from the heat exchanger outlet to the pump inlet. The numerical calculation results demonstrate that the baffle can effectively reduce the probability of bubbles entering the reactor core. This research has certain guiding significance for the study of bubble migration characteristics in SGTR of small lead-based reactors, and provides a reference for judging accident impacts and designing mitigation measures.
{"title":"Analysis of bubble migration characteristics and design of mitigation measures after a steam generator tube rupture accident in a small lead-based reactor","authors":"Yi Ren , Ming Jin , Shize Ouyang , Chao Lian , Chaodong Zhang , Jieqiong Jiang , FDS Consortium","doi":"10.1016/j.anucene.2026.112183","DOIUrl":"10.1016/j.anucene.2026.112183","url":null,"abstract":"<div><div>Steam Generator Tube Rupture (SGTR) accident is one of the critical safety accidents in lead-based reactors. In this study, a 200MWt small lead-based reactor was taken as the research object. A computational domain was established using a 1/2 geometric model of the primary loop system of the reactor. Combined with the Eulerian-Lagrangian CFD Discrete Phase Model and the drag coefficient correlation, the bubble migration characteristics were analyzed, and the law of bubble migration characteristics was obtained. Additionally, the possibility of bubbles entering the reactor core was evaluated. The results show that: for accidents with lower break location, the probability of bubbles entering the core is higher; besides, the smaller the bubble diameter, the higher the probability of entering the core. Based on the calculation results, a baffle-based mitigation measure was designed along the flow path from the heat exchanger outlet to the pump inlet. The numerical calculation results demonstrate that the baffle can effectively reduce the probability of bubbles entering the reactor core. This research has certain guiding significance for the study of bubble migration characteristics in SGTR of small lead-based reactors, and provides a reference for judging accident impacts and designing mitigation measures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112183"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186634","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 : 2026-06-01Epub Date: 2026-01-19DOI: 10.1016/j.anucene.2026.112148
Kaiyuan Wang, Yayun Luo, Yanan Zhu, Xin Jin
Cr2O3-doped large-grain UO2 pellets enhance fission gas retention and fuel rod accident tolerance. In this paper, drawing upon the behavior mechanism of fission gas, a mechanistic fission gas release model has been developed to evaluate fission gas release in the engineering design of large-grained UO2 fuel rods. The model integrates diffusion coefficients derived from free-energy cluster dynamics and couples with JASMINE from the NATENE package. Based on the new model-based software, a detailed validation was performed by combining the experimental data from international projects. The results show excellent agreement between simulation results and experimental measurements for key parameters, including fission gas release rate, internal pressure, and temperature. A prediction deviation of the fission gas release rate remains within ±35%. In summary, the new model-based software demonstrates the capability to simulate in-reactor fission gas behavior for Cr2O3-doped large-grained UO2 nuclear fuel.
{"title":"Development and verification of a new fission gas release model for large-grained UO2 pellets","authors":"Kaiyuan Wang, Yayun Luo, Yanan Zhu, Xin Jin","doi":"10.1016/j.anucene.2026.112148","DOIUrl":"10.1016/j.anucene.2026.112148","url":null,"abstract":"<div><div>Cr<sub>2</sub>O<sub>3</sub>-doped large-grain UO<sub>2</sub> pellets enhance fission gas retention and fuel rod accident tolerance. In this paper, drawing upon the behavior mechanism of fission gas, a mechanistic fission gas release model has been developed to evaluate fission gas release in the engineering design of large-grained UO<sub>2</sub> fuel rods. The model integrates diffusion coefficients derived from free-energy cluster dynamics and couples with JASMINE from the NATENE package. Based on the new model-based software, a detailed validation was performed by combining the experimental data from international projects. The results show excellent agreement between simulation results and experimental measurements for key parameters, including fission gas release rate, internal pressure, and temperature. A prediction deviation of the fission gas release rate remains within ±35%. In summary, the new model-based software demonstrates the capability to simulate in-reactor fission gas behavior for Cr<sub>2</sub>O<sub>3</sub>-doped large-grained UO<sub>2</sub> nuclear fuel.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112148"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035570","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 : 2026-06-01Epub Date: 2026-01-20DOI: 10.1016/j.anucene.2026.112158
Xinling Dai , Dechang Cai , Jin Cai
Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize . This method significantly reduces computational effort while maintaining a slight deviation of less than 2.62. This approach enables rapid yet rigorous criticality safety assessments.
{"title":"Efficient critical safety analysis model for PWR fuel assembly under transport accidents by neutron worth iteration","authors":"Xinling Dai , Dechang Cai , Jin Cai","doi":"10.1016/j.anucene.2026.112158","DOIUrl":"10.1016/j.anucene.2026.112158","url":null,"abstract":"<div><div>Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span>. This method significantly reduces computational effort while maintaining a slight <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> deviation of less than 2.62<span><math><mo>‰</mo></math></span>. This approach enables rapid yet rigorous criticality safety assessments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112158"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035575","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