Pub Date : 2026-04-01Epub Date: 2026-02-06DOI: 10.1016/j.nucengdes.2026.114807
Yanlong Wen, Hongjian Zhang, Haiyan Xiao, Qing Zhu, Liguo Zhang, Hong Li
Accurate determination of uranium loading in spherical fuel elements of high-temperature gas-cooled reactors (HTGRs) is essential for nuclear material accounting and reactor safety. Conventional γ-ray spectrometry based on a single characteristic peak (185.7 keV) suffers from significant variability due to differences in fuel sphere dimensions and TRISO particle distributions, yielding a coefficient of variation (CV) of around 1%.
In this study, a machine learning–based error correction framework was developed to enhance UO2 loading measurements. Three representative algorithms—Random Forest (RF), Support Vector Regression (SVR), and Multilayer Perceptron (MLP)—were trained on Geant4-simulated γ-ray data, incorporating both raw counts and composite features derived from six characteristic γ lines. Results show that all models reduced the CV to 0.3–0.4%, with the MLP achieving the best performance (CV ≤ 0.35%).
Further comparative analysis indicated that the three algorithms emphasize different mathematical solution spaces: RF captures discrete threshold effects, SVR provides smooth nonlinear fitting via kernel methods, and MLP flexibly spans a richer function space to approximate the mixed discrete–continuous nature of γ-ray attenuation. The integration of these perspectives demonstrates not only the potential of machine learning for nuclear material measurement but also the complementarity of different models in balancing accuracy, robustness, and interpretability.
This work provides a practical framework for improving non-destructive UO2 loading measurements in HTGR fuel elements, supporting more reliable nuclear material accounting and contributing to advanced reactor fuel management.
{"title":"Comparative analysis of machine learning methods for correcting uranium loading measurement errors in pebble-bed HTGR spherical fuel elements","authors":"Yanlong Wen, Hongjian Zhang, Haiyan Xiao, Qing Zhu, Liguo Zhang, Hong Li","doi":"10.1016/j.nucengdes.2026.114807","DOIUrl":"10.1016/j.nucengdes.2026.114807","url":null,"abstract":"<div><div>Accurate determination of uranium loading in spherical fuel elements of high-temperature gas-cooled reactors (HTGRs) is essential for nuclear material accounting and reactor safety. Conventional γ-ray spectrometry based on a single characteristic peak (185.7 keV) suffers from significant variability due to differences in fuel sphere dimensions and TRISO particle distributions, yielding a coefficient of variation (CV) of around 1%.</div><div>In this study, a machine learning–based error correction framework was developed to enhance UO<sub>2</sub> loading measurements. Three representative algorithms—Random Forest (RF), Support Vector Regression (SVR), and Multilayer Perceptron (MLP)—were trained on Geant4-simulated γ-ray data, incorporating both raw counts and composite features derived from six characteristic γ lines. Results show that all models reduced the CV to 0.3–0.4%, with the MLP achieving the best performance (CV ≤ 0.35%).</div><div>Further comparative analysis indicated that the three algorithms emphasize different mathematical solution spaces: RF captures discrete threshold effects, SVR provides smooth nonlinear fitting via kernel methods, and MLP flexibly spans a richer function space to approximate the mixed discrete–continuous nature of γ-ray attenuation. The integration of these perspectives demonstrates not only the potential of machine learning for nuclear material measurement but also the complementarity of different models in balancing accuracy, robustness, and interpretability.</div><div>This work provides a practical framework for improving non-destructive UO<sub>2</sub> loading measurements in HTGR fuel elements, supporting more reliable nuclear material accounting and contributing to advanced reactor fuel management.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114807"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190734","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 : 2026-04-01Epub Date: 2026-02-06DOI: 10.1016/j.nucengdes.2026.114810
Rongjie Li , Wukun Zhu , Chengwen Qiang , Minghan He , Han Wang , Lu Liu , Sicheng Wang , Dajun Fan
The China Initiative Accelerator Driven System (CiADS) holds significant promise in the transmutation of long-lived nuclear waste and the efficient utilization of nuclear fuel. Understanding the coolant flow characteristics within reactor fuel assemblies is crucial for ensuring the safe and economical operation of nuclear facilities. In this study, a visual hydraulic experimental platform was constructed to investigate the flow velocity distribution within the fuel assembly of CiADS. Utilizing particle image velocimetry technology, the transverse velocity distribution at subchannel interfaces within a 19-pin wire-wrapped fuel assembly was systematically investigated. The results reveal a distinct cosine-type distribution of the normalized transverse velocity at the interfaces between internal subchannels, with key flow features demonstrating robust Reynolds-number independence. Notably, significant wall-induced asymmetries alter this distribution at the interface between the internal subchannel and the side-wall subchannel. A novel transverse mixing model is developed, which shows close agreement with experimental data, confirming its predictive capability for inter-subchannel mixing in wire-wrapped bundles. This research further confirms the periodic disturbance characteristics induced by wire-wraps on transverse flow and provides a theoretical basis for predicting normalized transverse velocity in wire-wrapped bundles of different sizes, thereby deepening the understanding of the transverse mixing mechanism within rod bundles.
{"title":"PIV study on subchannel cross-flow characteristics in 19-pin wire-wrapped bundle channels","authors":"Rongjie Li , Wukun Zhu , Chengwen Qiang , Minghan He , Han Wang , Lu Liu , Sicheng Wang , Dajun Fan","doi":"10.1016/j.nucengdes.2026.114810","DOIUrl":"10.1016/j.nucengdes.2026.114810","url":null,"abstract":"<div><div>The China Initiative Accelerator Driven System (CiADS) holds significant promise in the transmutation of long-lived nuclear waste and the efficient utilization of nuclear fuel. Understanding the coolant flow characteristics within reactor fuel assemblies is crucial for ensuring the safe and economical operation of nuclear facilities. In this study, a visual hydraulic experimental platform was constructed to investigate the flow velocity distribution within the fuel assembly of CiADS. Utilizing particle image velocimetry technology, the transverse velocity distribution at subchannel interfaces within a 19-pin wire-wrapped fuel assembly was systematically investigated. The results reveal a distinct cosine-type distribution of the normalized transverse velocity at the interfaces between internal subchannels, with key flow features demonstrating robust Reynolds-number independence. Notably, significant wall-induced asymmetries alter this distribution at the interface between the internal subchannel and the side-wall subchannel. A novel transverse mixing model is developed, which shows close agreement with experimental data, confirming its predictive capability for inter-subchannel mixing in wire-wrapped bundles. This research further confirms the periodic disturbance characteristics induced by wire-wraps on transverse flow and provides a theoretical basis for predicting normalized transverse velocity in wire-wrapped bundles of different sizes, thereby deepening the understanding of the transverse mixing mechanism within rod bundles.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114810"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190730","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 : 2026-04-01Epub Date: 2026-02-04DOI: 10.1016/j.nucengdes.2026.114798
S. Lopes , S. Keijers , L. Koloszar , J. Pacio , P. Planquart , K. Van Tichelen
This article is part of a series devoted to the analysis of pool thermal-hydraulic phenomena in heavy-liquid metal systems, combining experimental investigations with numerical simulations. The reference system is the Accelerator Driven System MYRRHA, currently under construction at SCK CEN (Belgium), which couples a particle accelerator with a pool-type research reactor cooled by lead-bismuth eutectic (LBE). An extensive experimental campaign has been carried out in the E-SCAPE facility, a 1/6-scaled LBE mockup of the primary vessel of MYRRHA, equipped with additional auxiliary circuits. In parallel, numerical studies based on computational fluid dynamics (CFD) are performed with the main objective of developing and validating robust CFD methodology capable of accurately simulating large-scale heavy-liquid metal experimental facilities. These studies allow the investigation of thermal-hydraulic phenomena and provide detailed insight into thermal and momentum fields. Model validation is achieved through systematic comparison between simulations and measurements obtained in E-SCAPE for a range of operating scenarios, including forced and natural circulation, symmetric and asymmetric loss-of-flow events, and asymmetric decay heat removal conditions.
In this paper, the numerical methodology to simulate the E-SCAPE facility under both steady-state and transient operating conditions is established. Since the natural circulation condition and the transition from forced to natural circulation are characterized by a strong coupling between thermal and momentum fields, detailed numerical description of the E-SCAPE core simulator, of the upper plena regions and of the external circuits are important. The main reason is that the flow rate in the core region is driven by the balance between the pressure drop and the thermal buoyancy effect and the flow rates in the external circuits are based on pressure losses through the valves, filters, and pumps. The impact of geometric simplifications of core plates and heaters, relative to the real configuration, is evaluated on the flow and on the temperature fields.
Furthermore, the effects of numerical mesh refinement and thermal boundary conditions are investigated under both iso-thermal and thermal forced circulation operating conditions. Simulated temperature distributions and pressure losses in the pool are compared against local experimental measurements. Finally, practical guidelines are proposed for CFD simulation of forced circulation, natural circulation, and transient conditions, addressing geometry simplifications, definition of fluid and solid domains, characterization of external circuits, and numerical strategies for transient initiation. The initial conditions for transient simulations are also computed and successfully compared with experimental data.
{"title":"CFD simulation methodology for large-scale heavy liquid metal facility relevant for MYRRHA: Application to E-SCAPE facility","authors":"S. Lopes , S. Keijers , L. Koloszar , J. Pacio , P. Planquart , K. Van Tichelen","doi":"10.1016/j.nucengdes.2026.114798","DOIUrl":"10.1016/j.nucengdes.2026.114798","url":null,"abstract":"<div><div>This article is part of a series devoted to the analysis of pool thermal-hydraulic phenomena in heavy-liquid metal systems, combining experimental investigations with numerical simulations. The reference system is the Accelerator Driven System MYRRHA, currently under construction at SCK CEN (Belgium), which couples a particle accelerator with a pool-type research reactor cooled by lead-bismuth eutectic (LBE). An extensive experimental campaign has been carried out in the E-SCAPE facility, a 1/6-scaled LBE mockup of the primary vessel of MYRRHA, equipped with additional auxiliary circuits. In parallel, numerical studies based on computational fluid dynamics (CFD) are performed with the main objective of developing and validating robust CFD methodology capable of accurately simulating large-scale heavy-liquid metal experimental facilities. These studies allow the investigation of thermal-hydraulic phenomena and provide detailed insight into thermal and momentum fields. Model validation is achieved through systematic comparison between simulations and measurements obtained in E-SCAPE for a range of operating scenarios, including forced and natural circulation, symmetric and asymmetric loss-of-flow events, and asymmetric decay heat removal conditions.</div><div>In this paper, the numerical methodology to simulate the E-SCAPE facility under both steady-state and transient operating conditions is established. Since the natural circulation condition and the transition from forced to natural circulation are characterized by a strong coupling between thermal and momentum fields, detailed numerical description of the E-SCAPE core simulator, of the upper plena regions and of the external circuits are important. The main reason is that the flow rate in the core region is driven by the balance between the pressure drop and the thermal buoyancy effect and the flow rates in the external circuits are based on pressure losses through the valves, filters, and pumps. The impact of geometric simplifications of core plates and heaters, relative to the real configuration, is evaluated on the flow and on the temperature fields.</div><div>Furthermore, the effects of numerical mesh refinement and thermal boundary conditions are investigated under both iso-thermal and thermal forced circulation operating conditions. Simulated temperature distributions and pressure losses in the pool are compared against local experimental measurements. Finally, practical guidelines are proposed for CFD simulation of forced circulation, natural circulation, and transient conditions, addressing geometry simplifications, definition of fluid and solid domains, characterization of external circuits, and numerical strategies for transient initiation. The initial conditions for transient simulations are also computed and successfully compared with experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114798"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190733","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 : 2026-04-01Epub Date: 2026-01-28DOI: 10.1016/j.nucengdes.2026.114790
P. Pfahl , B. Kędzierska , I. Kim , A. Chambon , L. Fischer , L. Bureš , J. Groth-Jensen , Y. Kim , A. Rineiski , B. Lauritzen
Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.
{"title":"Multi-Physics Benchmark for a Thermal Molten Salt Reactor","authors":"P. Pfahl , B. Kędzierska , I. Kim , A. Chambon , L. Fischer , L. Bureš , J. Groth-Jensen , Y. Kim , A. Rineiski , B. Lauritzen","doi":"10.1016/j.nucengdes.2026.114790","DOIUrl":"10.1016/j.nucengdes.2026.114790","url":null,"abstract":"<div><div>Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114790"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078926","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 : 2026-04-01Epub Date: 2026-01-25DOI: 10.1016/j.nucengdes.2026.114789
K. Khumsa-Ang , M. Fulger , A. Sáez Maderuelo , A. Toivonen , M. Sipova , D. Marusakova , L. Zhang , J. Macák , R. Novotny
This document presents a summary of the most relevant research and development (R&D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).
One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.
{"title":"A state-of-the-art review of R&D for the supercritical water-cooled reactor technology. Part II materials & chemistry","authors":"K. Khumsa-Ang , M. Fulger , A. Sáez Maderuelo , A. Toivonen , M. Sipova , D. Marusakova , L. Zhang , J. Macák , R. Novotny","doi":"10.1016/j.nucengdes.2026.114789","DOIUrl":"10.1016/j.nucengdes.2026.114789","url":null,"abstract":"<div><div>This document presents a summary of the most relevant research and development (R&D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).</div><div>One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114789"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078928","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 : 2026-04-01Epub Date: 2026-01-28DOI: 10.1016/j.nucengdes.2026.114792
Jin-Young Park, Tae-Hyun Kwon
Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.
{"title":"Prediction of leakage rates from containment buildings under ultimate internal pressure","authors":"Jin-Young Park, Tae-Hyun Kwon","doi":"10.1016/j.nucengdes.2026.114792","DOIUrl":"10.1016/j.nucengdes.2026.114792","url":null,"abstract":"<div><div>Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114792"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078930","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 : 2026-04-01Epub Date: 2026-01-12DOI: 10.1016/j.nucengdes.2025.114696
Quazi Md. Zobaer Shah , Md Mahabub Hasan Mousum , Debashis Datta , Md. Arefin Kowser , Mohammad Asaduzzaman Chowdhury , Md. Shojib Mia , Shahpara Sheikh Dola
Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.
{"title":"Integrated multiphysics assessment of grid-to-rod fretting mitigation: enhancing nuclear safety through improved fuel integrity","authors":"Quazi Md. Zobaer Shah , Md Mahabub Hasan Mousum , Debashis Datta , Md. Arefin Kowser , Mohammad Asaduzzaman Chowdhury , Md. Shojib Mia , Shahpara Sheikh Dola","doi":"10.1016/j.nucengdes.2025.114696","DOIUrl":"10.1016/j.nucengdes.2025.114696","url":null,"abstract":"<div><div>Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114696"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950226","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 : 2026-04-01Epub Date: 2026-01-30DOI: 10.1016/j.nucengdes.2026.114770
Nóra Vér , Róbert Farkas , Berta Bürger , Anna Pintér-Csordás , Tamás Novotny , Erzsébet Perez-Feró , Péter Szabó , Levente Illés , Zoltán Kovács , Dávid Cinger , Martin Ševeček , Zoltán Hózer
The CODEX-ATF integral bundle test was conducted within the framework of the IAEA Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS) project at the CODEX (COre Degradation Experiment) facility in Hungary. The electrically heated seven-rod bundle consisted of four Cr-coated and three uncoated Zr alloy cladding tubes, enabling a direct comparison of their behavior under high-temperature accident conditions. The experiment primarily aimed to investigate fuel failure and degradation mechanisms.
During the test, several rods exhibited ballooning and burst phenomena. The maximum temperature exceeded 1600 °C. The transient was terminated by a bottom-up water quench. The total hydrogen generation was approximately 3 g, indicating substantial oxidation of the zirconium-based components. Intensive Zr-Cr eutectic interaction was observed in the hottest region of the bundle on the Cr-coated claddings. Post-test examinations revealed pronounced deformation and failure in both coated and uncoated claddings.
{"title":"Side-by-side high-temperature accident performance of ATF and conventional claddings in the CODEX-ATF experiment","authors":"Nóra Vér , Róbert Farkas , Berta Bürger , Anna Pintér-Csordás , Tamás Novotny , Erzsébet Perez-Feró , Péter Szabó , Levente Illés , Zoltán Kovács , Dávid Cinger , Martin Ševeček , Zoltán Hózer","doi":"10.1016/j.nucengdes.2026.114770","DOIUrl":"10.1016/j.nucengdes.2026.114770","url":null,"abstract":"<div><div>The CODEX-ATF integral bundle test was conducted within the framework of the IAEA Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS) project at the CODEX (COre Degradation Experiment) facility in Hungary. The electrically heated seven-rod bundle consisted of four Cr-coated and three uncoated Zr alloy cladding tubes, enabling a direct comparison of their behavior under high-temperature accident conditions. The experiment primarily aimed to investigate fuel failure and degradation mechanisms.</div><div>During the test, several rods exhibited ballooning and burst phenomena. The maximum temperature exceeded 1600 °C. The transient was terminated by a bottom-up water quench. The total hydrogen generation was approximately 3 g, indicating substantial oxidation of the zirconium-based components. Intensive Zr-Cr eutectic interaction was observed in the hottest region of the bundle on the Cr-coated claddings. Post-test examinations revealed pronounced deformation and failure in both coated and uncoated claddings.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114770"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081254","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 : 2026-04-01Epub Date: 2026-01-27DOI: 10.1016/j.nucengdes.2026.114760
Hua Rong , Xuan Zhang , Yajing Shen , Shuai Tan , Xinglang Fan , Yang Du , Yifeng Feng
This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.
{"title":"Thermo-mechanical coupled analysis and probabilistic safety evaluation of the prestressed concrete containment vessel under accident pressure","authors":"Hua Rong , Xuan Zhang , Yajing Shen , Shuai Tan , Xinglang Fan , Yang Du , Yifeng Feng","doi":"10.1016/j.nucengdes.2026.114760","DOIUrl":"10.1016/j.nucengdes.2026.114760","url":null,"abstract":"<div><div>This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114760"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078929","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 : 2026-04-01Epub Date: 2026-01-21DOI: 10.1016/j.nucengdes.2026.114772
Yuchen Li, Yanmin Zhou, Haifeng Gu, Yichen Zhang, Shuolei Fan
Fuel cladding in lead‑bismuth eutectic (LBE) cooled reactors may develop micron-scale failures during long-term high-temperature operation. The flow regime of gas submerged jet at the defect orifice directly influences the scrubbing behavior of these fission products. This study investigated a micron-scale failures in LBE system by conducting visualization experiments using inert gas and deionized water as working fluids. The research first systematically compared the criteria for bubbling-jet regime transition between micron-scale and millimeter-scale orifices. The results revealed that the traditional criteria based on the liquid-phase Weber number or Mach number, which are applicable for millimeter-scale orifices failed at micron scale. The phenomenon was primarily attributed to the lower gas momentum from micron-scale orifice, which allowed for continued flow regime evolution even after critical flow was reached. Consequently, a new criterion defined as the product of a density correction factor and the liquid Weber number was proposed to accurately predict the flow regime transition. Furthermore, an empirical correlation for predicting the Sauter mean diameter of the gas bubbles was established based on the SPARC90 bubble size prediction model, with a prediction error within the range of −15% to +15%. Finally, the results were extrapolated to a prototypical lead‑bismuth environment through scaling analysis, which verified a good predictive capability for bubble size generated from micron-scale orifices of this correlation. The research provided theoretical support for two-phase flow regime identification and bubble behavior prediction in the safety analysis of lead‑bismuth reactors.
{"title":"Gas-liquid flow characteristics through a Micron scale orifice of failed fuel pin in Lead-bismuth cooled reactors","authors":"Yuchen Li, Yanmin Zhou, Haifeng Gu, Yichen Zhang, Shuolei Fan","doi":"10.1016/j.nucengdes.2026.114772","DOIUrl":"10.1016/j.nucengdes.2026.114772","url":null,"abstract":"<div><div>Fuel cladding in lead‑bismuth eutectic (LBE) cooled reactors may develop micron-scale failures during long-term high-temperature operation. The flow regime of gas submerged jet at the defect orifice directly influences the scrubbing behavior of these fission products. This study investigated a micron-scale failures in LBE system by conducting visualization experiments using inert gas and deionized water as working fluids. The research first systematically compared the criteria for bubbling-jet regime transition between micron-scale and millimeter-scale orifices. The results revealed that the traditional criteria based on the liquid-phase Weber number or Mach number, which are applicable for millimeter-scale orifices failed at micron scale. The phenomenon was primarily attributed to the lower gas momentum from micron-scale orifice, which allowed for continued flow regime evolution even after critical flow was reached. Consequently, a new criterion defined as the product of a density correction factor and the liquid Weber number was proposed to accurately predict the flow regime transition. Furthermore, an empirical correlation for predicting the Sauter mean diameter of the gas bubbles was established based on the SPARC90 bubble size prediction model, with a prediction error within the range of −15% to +15%. Finally, the results were extrapolated to a prototypical lead‑bismuth environment through scaling analysis, which verified a good predictive capability for bubble size generated from micron-scale orifices of this correlation. The research provided theoretical support for two-phase flow regime identification and bubble behavior prediction in the safety analysis of lead‑bismuth reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114772"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036026","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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