Pub Date : 2026-04-01Epub Date: 2026-01-20DOI: 10.1016/j.nucengdes.2026.114767
Armando Nava Dominguez , Chukwudi Azih , Alberto D'Ansi Mendoza España , Hussam Zahlan , Guido Mazzini , Alis Musa-Ruscak , Sara Kassem , Andrea Pucciarelli , Walter Ambrosini , Fabian Wiltschko , Ivan Otic , Tamás Varju , Attila Kiss , Pan Wu , Elena Poplavskaia
This study 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 is operated at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of higher-grade process heat. Several collaborations have been established to support the development of this technology under the GIF umbrella, as well as through other international avenues. Therefore, the development is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. Globally, the R&D programs have been methodologically executed in phases, namely: fundamental R&D, validation and verification of assumptions used in R&D analyses, and pre-conceptualization of supercritical water-cooled reactors (SCWRs).
This article summarizes the recent R&D work performed to support the development of the SCWR technology on thermalhydraulics and safety, economics and licensing. Furthermore, R&D highlights, identified knowledge gaps, conclusions, and recommendations are presented.
{"title":"A state-of-the-art review of R&D for the super critical water-cooled reactor technology. Part I economics, thermalhydraulics, safety and licensing","authors":"Armando Nava Dominguez , Chukwudi Azih , Alberto D'Ansi Mendoza España , Hussam Zahlan , Guido Mazzini , Alis Musa-Ruscak , Sara Kassem , Andrea Pucciarelli , Walter Ambrosini , Fabian Wiltschko , Ivan Otic , Tamás Varju , Attila Kiss , Pan Wu , Elena Poplavskaia","doi":"10.1016/j.nucengdes.2026.114767","DOIUrl":"10.1016/j.nucengdes.2026.114767","url":null,"abstract":"<div><div>This study 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 is operated at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of higher-grade process heat. Several collaborations have been established to support the development of this technology under the GIF umbrella, as well as through other international avenues. Therefore, the development is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. Globally, the R&D programs have been methodologically executed in phases, namely: fundamental R&D, validation and verification of assumptions used in R&D analyses, and pre-conceptualization of supercritical water-cooled reactors (SCWRs).</div><div>This article summarizes the recent R&D work performed to support the development of the SCWR technology on thermalhydraulics and safety, economics and licensing. Furthermore, R&D highlights, identified knowledge gaps, conclusions, and recommendations are presented.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114767"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036056","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: 2026-01-20DOI: 10.1016/j.nucengdes.2026.114786
Shusheng Dai , Xiaochang Li , Yu Zhang , Ruifeng Tian , Jiming Wen , Sichao Tan
Wire-wrapped fuel rod bundles are commonly employed in Generation IV fast reactors. However, the swirl and perturbations induced by the helical wires can significantly enhance fluid-structure interaction and increase the risk of instability, and traditional models still struggle to capture the response characteristics. This study establishes a high-order theoretical model of fluid-structure interaction for wire-wrapped fuel rods in axial flow, grounded in the fundamental principles of Newtonian mechanics. Building upon the dynamic framework of a bare rod, the model incorporates the additional effects of wire-wrap mass, stiffness, and drag, enabling a concise yet systematic representation of wire-wrap effects. A dimensionless parameter system is established for nondimensionalization, and the Galerkin method is applied for discretization. The resulting system matrix is solved computationally to obtain the first wet-mode natural frequency and the critical instability velocity. The predicted results show good agreement with numerical simulations, with relative frequency errors below 8%, validating the model's reliability. Parameter sensitivity analysis is further performed to elucidate the effects of wire-wrap diameter and pitch on the critical instability velocity, and its coupled regulatory mechanisms on system stability under different rod diameters and lengths. The results indicate that increasing the wire-wrap diameter intensifies flow disturbances and reduces the critical instability velocity; a critical range of wire-wrap pitch exists that leads to the lowest system stability, displaying non-monotonic characteristics; and variations in rod parameters influence the wire-wrap effect on system stability, with smaller diameters or longer rods promoting instability, diameter reduction enhancing coupling, and length increase weakening it.
{"title":"Theoretical model of fluid-structure interaction and prediction of fluidelastic instability for wire-wrapped fuel rods in axial flow","authors":"Shusheng Dai , Xiaochang Li , Yu Zhang , Ruifeng Tian , Jiming Wen , Sichao Tan","doi":"10.1016/j.nucengdes.2026.114786","DOIUrl":"10.1016/j.nucengdes.2026.114786","url":null,"abstract":"<div><div>Wire-wrapped fuel rod bundles are commonly employed in Generation IV fast reactors. However, the swirl and perturbations induced by the helical wires can significantly enhance fluid-structure interaction and increase the risk of instability, and traditional models still struggle to capture the response characteristics. This study establishes a high-order theoretical model of fluid-structure interaction for wire-wrapped fuel rods in axial flow, grounded in the fundamental principles of Newtonian mechanics. Building upon the dynamic framework of a bare rod, the model incorporates the additional effects of wire-wrap mass, stiffness, and drag, enabling a concise yet systematic representation of wire-wrap effects. A dimensionless parameter system is established for nondimensionalization, and the Galerkin method is applied for discretization. The resulting system matrix is solved computationally to obtain the first wet-mode natural frequency and the critical instability velocity. The predicted results show good agreement with numerical simulations, with relative frequency errors below 8%, validating the model's reliability. Parameter sensitivity analysis is further performed to elucidate the effects of wire-wrap diameter and pitch on the critical instability velocity, and its coupled regulatory mechanisms on system stability under different rod diameters and lengths. The results indicate that increasing the wire-wrap diameter intensifies flow disturbances and reduces the critical instability velocity; a critical range of wire-wrap pitch exists that leads to the lowest system stability, displaying non-monotonic characteristics; and variations in rod parameters influence the wire-wrap effect on system stability, with smaller diameters or longer rods promoting instability, diameter reduction enhancing coupling, and length increase weakening it.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114786"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036024","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.2025.114733
María González-Alvear , Mariano Lázaro , Daniel Alvear , Eugenia Morgado , Miguel Ángel Jiménez , David Lázaro
Fire Dynamics Simulator (FDS) is a well-known fire computer model, which has been widely applied for different scenarios. In particular, several standards and guidelines support its use in fire safety engineering approaches in nuclear plants. Although the uncertainty of the FDS model has been analysed and collected in literature, the influence of each input parameter has not yet been fully addressed.
Some of these previous contributions were based on the Benchmark Exercise No. 3 of the International Collaborative Fire Model Project (NUREG 6905). Moreover, the Best Practice Guidelines of NEA/CSNI/R (2014)11 were used as the reference to analyse the influence of the boundary conditions on simulation results. This work aims to study the impact of selected key fire dynamics parameters on the simulations of that scenario, updating previous findings.
Since this fire scenario involves three horizontal cable trays and one vertical cable tray, it is of special interest for nuclear power plants. Moreover, it is relevant to analyse the influence of input parameters on cable ignition. A sensitivity analysis was conducted, to evaluate the most important parameters for the selected scenario, focussing on ventilation and the thermal properties of the cables such as conductivity, specific heat, density, emissivity.
The results show the influence of each parameter in the surface temperature and heat flux in the different cable trays. Consequently, this enables the authors to formulate some recommendations for defining fire scenarios when applying fire safety engineering principles in nuclear power plants.
{"title":"Effect of FDS uncertainty in fire simulations of nuclear power plants under different ventilation conditions","authors":"María González-Alvear , Mariano Lázaro , Daniel Alvear , Eugenia Morgado , Miguel Ángel Jiménez , David Lázaro","doi":"10.1016/j.nucengdes.2025.114733","DOIUrl":"10.1016/j.nucengdes.2025.114733","url":null,"abstract":"<div><div>Fire Dynamics Simulator (FDS) is a well-known fire computer model, which has been widely applied for different scenarios. In particular, several standards and guidelines support its use in fire safety engineering approaches in nuclear plants. Although the uncertainty of the FDS model has been analysed and collected in literature, the influence of each input parameter has not yet been fully addressed.</div><div>Some of these previous contributions were based on the Benchmark Exercise No. 3 of the International Collaborative Fire Model Project (NUREG 6905). Moreover, the Best Practice Guidelines of NEA/CSNI/R (2014)11 were used as the reference to analyse the influence of the boundary conditions on simulation results. This work aims to study the impact of selected key fire dynamics parameters on the simulations of that scenario, updating previous findings.</div><div>Since this fire scenario involves three horizontal cable trays and one vertical cable tray, it is of special interest for nuclear power plants. Moreover, it is relevant to analyse the influence of input parameters on cable ignition. A sensitivity analysis was conducted, to evaluate the most important parameters for the selected scenario, focussing on ventilation and the thermal properties of the cables such as conductivity, specific heat, density, emissivity.</div><div>The results show the influence of each parameter in the surface temperature and heat flux in the different cable trays. Consequently, this enables the authors to formulate some recommendations for defining fire scenarios when applying fire safety engineering principles in nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114733"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036030","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: 2026-02-03DOI: 10.1016/j.nucengdes.2025.114741
K. Sergeenko , A. Krutikov , L. Golibrodo , A. Zefirova , M. Shishlenin
Liquid metal coolants are considered in various fast neutron reactors. This paper presents the results of DNS calculations of heat and mass transfer processes of liquid metal coolants in the following regions: a parallel-plate channel, a rod bundle at cross-flow, and a channel with a step. The regions are selected in such a way as to describe as comprehensively as possible the thermal hydraulics of the fuel assembly, in which the spacing of the rods is carried out by the spacer grids. Two Reynolds numbers (Re = 2500 and 5000) and three Prandtl numbers (Pr = 0.0043, 0.0154 and 0.0324) are considered. The Prandtl numbers considered correspond to the liquid metal coolants used in fast neutron reactors: sodium, lead and lead bismuth, respectively.
The results of DNS simulations were compared with the data obtained using various RANS turbulence models (k-ε realizable two layer, k-ω SST, k-ε EB and k-ε V2F). It is shown that the k-ε V2F turbulence model describes the thermal-hydraulic processes in the best way. It recommends use this model for thermal-hydraulic calculations of fuel assemblies in reactor plants with liquid metal coolants at low Reynolds numbers.
It is shown that for all considered RANS turbulence models, the error in determining the heat transfer increases with increasing Peclet number in the parallel-plate channel. This is due to the features of the mechanisms of turbulent heat transfer for coolants at low Prandtl number.
The temperature profile is also compared with theoretical data. Satisfactory agreement is shown for all simulations in the area up to y+ ≈ 60. This can be used in experimental studies focused on determining the Nusselt number.
{"title":"Selection of RANS turbulence model for calculating thermal hydraulics of fuel assemblies of LMC reactors at low Reynolds numbers","authors":"K. Sergeenko , A. Krutikov , L. Golibrodo , A. Zefirova , M. Shishlenin","doi":"10.1016/j.nucengdes.2025.114741","DOIUrl":"10.1016/j.nucengdes.2025.114741","url":null,"abstract":"<div><div>Liquid metal coolants are considered in various fast neutron reactors. This paper presents the results of DNS calculations of heat and mass transfer processes of liquid metal coolants in the following regions: a parallel-plate channel, a rod bundle at cross-flow, and a channel with a step. The regions are selected in such a way as to describe as comprehensively as possible the thermal hydraulics of the fuel assembly, in which the spacing of the rods is carried out by the spacer grids. Two Reynolds numbers (<em>Re</em> = 2500 and 5000) and three Prandtl numbers (<em>Pr</em> = 0.0043, 0.0154 and 0.0324) are considered. The Prandtl numbers considered correspond to the liquid metal coolants used in fast neutron reactors: sodium, lead and lead bismuth, respectively.</div><div>The results of DNS simulations were compared with the data obtained using various RANS turbulence models (k-ε realizable two layer, k-ω SST, k-ε EB and k-ε V2F). It is shown that the k-ε V2F turbulence model describes the thermal-hydraulic processes in the best way. It recommends use this model for thermal-hydraulic calculations of fuel assemblies in reactor plants with liquid metal coolants at low Reynolds numbers.</div><div>It is shown that for all considered RANS turbulence models, the error in determining the heat transfer increases with increasing Peclet number in the parallel-plate channel. This is due to the features of the mechanisms of turbulent heat transfer for coolants at low Prandtl number.</div><div>The temperature profile is also compared with theoretical data. Satisfactory agreement is shown for all simulations in the area up to y<sup>+</sup> ≈ 60. This can be used in experimental studies focused on determining the Nusselt number.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114741"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190732","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 high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China represents the world's first demonstration power plant of its kind. As one of the Generation IV reactors, it is designed with inherent safety characteristics and offers potential for optimization in the structural integrity management of its reactor pressure vessel (RPV). Key advantages of the HTR-PM vessel, such as lower fast neutron flux than conventional pressurized water reactors, significantly slow thermal transients and reversed thermal gradients, collectively contribute to reducing both the risk and consequence of its failure. Using probabilistic fracture mechanics (PFM), a quantitative model has been developed to assess flaw initiation, crack propagation, and fracture of the RPV under typical HTR-PM transient and test conditions. This model enables assessment of the RPV failure probability, thereby establishing a basis for risk-informed safety evaluation and for optimization of both manufacturing and in-service inspection (ISI) requirements. Simulation results show that the RPV failure probability of HTR-PM is extremely low in the conservative case of no ISIs during its entire service life. Furthermore, flaws in circumferential weld regions contribute only minimally to the overall failure probability, suggesting potential for further extending inspection intervals or even exempting these areas from periodic inspection.
{"title":"Risk-informed structural integrity assessment of the HTR-PM reactor pressure vessel","authors":"Bowen Li, Haitao Wang, Yanhua Zheng, Zhengming Zhang, Heng Peng","doi":"10.1016/j.nucengdes.2026.114764","DOIUrl":"10.1016/j.nucengdes.2026.114764","url":null,"abstract":"<div><div>The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China represents the world's first demonstration power plant of its kind. As one of the Generation IV reactors, it is designed with inherent safety characteristics and offers potential for optimization in the structural integrity management of its reactor pressure vessel (RPV). Key advantages of the HTR-PM vessel, such as lower fast neutron flux than conventional pressurized water reactors, significantly slow thermal transients and reversed thermal gradients, collectively contribute to reducing both the risk and consequence of its failure. Using probabilistic fracture mechanics (PFM), a quantitative model has been developed to assess flaw initiation, crack propagation, and fracture of the RPV under typical HTR-PM transient and test conditions. This model enables assessment of the RPV failure probability, thereby establishing a basis for risk-informed safety evaluation and for optimization of both manufacturing and in-service inspection (ISI) requirements. Simulation results show that the RPV failure probability of HTR-PM is extremely low in the conservative case of no ISIs during its entire service life. Furthermore, flaws in circumferential weld regions contribute only minimally to the overall failure probability, suggesting potential for further extending inspection intervals or even exempting these areas from periodic inspection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114764"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981783","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: 2026-01-13DOI: 10.1016/j.nucengdes.2026.114763
J. Schäfer , S. Taş , U. Hampel
This work presents a novel training approach for machine learning based instance segmentation, without the need for manual annotated datasets. With applications in experimental investigations in bubbly flows in reactor safety research. This semi-supervised process consists of two different neural networks, a conditional generative adversarial network used for data generation and a U-net style convolutional neural network for instance segmentation. We validated the approach using a fully automated experimental setup creating layered bubble curtains, which enables an evaluation of the bubble size distribution with and without overlapping bubbles. The final neural network is able to measure the average bubble size as well as to recreate the bubble size distribution accurately.
{"title":"Gas bubble detection and segmentation using a machine learning approach leveraging semi-supervised training","authors":"J. Schäfer , S. Taş , U. Hampel","doi":"10.1016/j.nucengdes.2026.114763","DOIUrl":"10.1016/j.nucengdes.2026.114763","url":null,"abstract":"<div><div>This work presents a novel training approach for machine learning based instance segmentation, without the need for manual annotated datasets. With applications in experimental investigations in bubbly flows in reactor safety research. This semi-supervised process consists of two different neural networks, a conditional generative adversarial network used for data generation and a U-net style convolutional neural network for instance segmentation. We validated the approach using a fully automated experimental setup creating layered bubble curtains, which enables an evaluation of the bubble size distribution with and without overlapping bubbles. The final neural network is able to measure the average bubble size as well as to recreate the bubble size distribution accurately.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114763"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981787","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: 2026-02-14DOI: 10.1016/j.nucengdes.2025.114716
J.C. De la Rosa Blul , M.Á. Hernández-Ceballos , O. Parera Villacampa , G. Magrotti , A. Guglielmelli , C. Fazio
In situations where radiological risk does not originate from a single ongoing accident but from a widespread degradation of the operational context conditions affecting several nuclear installations, traditional event-oriented emergency preparedness may become insufficient. Under such conflict-related circumstances, protective planning requires a broader framework capable of capturing the aggregated and latent risk arising from multiple facilities and locations.
This paper presents the Diagnosis And Prognosis of Hazards in Nuclear Emergencies (DAPHNE) methodology (developed by the European Commission's Joint Research Centre), designed to support European Commission services and EU Member States in nuclear emergency preparedness and response. The methodology integrates MAAP, HYSPLIT, and the JRODOS decision-support system—together with dedicated in-house release models where appropriate—to provide a coherent assessment of both onsite accident progression and offsite radiological consequences.
Building on the conceptual framework introduced in Part I, this paper applies DAPHNE to a conflict-scenario case study to quantify integrated latent radiological risk and illustrate how the methodology enhances the identification of geographical areas where risk accumulates. By providing a harmonised numerical workflow for scenario selection, source-term calculation, dispersion modelling, and risk mapping, DAPHNE improves upon other methodologies and offers a scientifically grounded basis for designing conflict-related Emergency Planning Zones that complement existing site-specific arrangements.
Through its capacity to integrate multiple risk sources and their aggregated contribution to radiological exposure, DAPHNE strengthens preparedness planning under complex conflict scenarios and supports more informed decision-making across affected territories.
{"title":"Integrated latent risk of radiological damage under conflict situation – Part II: Application and results","authors":"J.C. De la Rosa Blul , M.Á. Hernández-Ceballos , O. Parera Villacampa , G. Magrotti , A. Guglielmelli , C. Fazio","doi":"10.1016/j.nucengdes.2025.114716","DOIUrl":"10.1016/j.nucengdes.2025.114716","url":null,"abstract":"<div><div>In situations where radiological risk does not originate from a single ongoing accident but from a widespread degradation of the operational context conditions affecting several nuclear installations, traditional event-oriented emergency preparedness may become insufficient. Under such conflict-related circumstances, protective planning requires a broader framework capable of capturing the aggregated and latent risk arising from multiple facilities and locations.</div><div>This paper presents the Diagnosis And Prognosis of Hazards in Nuclear Emergencies (DAPHNE) methodology (developed by the European Commission's Joint Research Centre), designed to support European Commission services and EU Member States in nuclear emergency preparedness and response. The methodology integrates MAAP, HYSPLIT, and the JRODOS decision-support system—together with dedicated in-house release models where appropriate—to provide a coherent assessment of both onsite accident progression and offsite radiological consequences.</div><div>Building on the conceptual framework introduced in Part I, this paper applies DAPHNE to a conflict-scenario case study to quantify integrated latent radiological risk and illustrate how the methodology enhances the identification of geographical areas where risk accumulates. By providing a harmonised numerical workflow for scenario selection, source-term calculation, dispersion modelling, and risk mapping, DAPHNE improves upon other methodologies and offers a scientifically grounded basis for designing conflict-related Emergency Planning Zones that complement existing site-specific arrangements.</div><div>Through its capacity to integrate multiple risk sources and their aggregated contribution to radiological exposure, DAPHNE strengthens preparedness planning under complex conflict scenarios and supports more informed decision-making across affected territories.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114716"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190813","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: 2026-01-23DOI: 10.1016/j.nucengdes.2026.114787
Satoshi Abe, Ari Hamdani, Shu Soma, Ryosuke Hangai, Masashi Ohmori, Akihiko Ohwada, Toshihito Ohmiya, Yasuteru Sibamoto
The Fukushima Daiichi accident underscored the urgent need to understand complex thermal-hydraulic phenomena governing containment integrity and gas mixture distribution during a severe accident. In response, the Japan Atomic Energy Agency (JAEA) established the CIGMA (Containment InteGral Measurement Apparatus) facility, a flagship large-scale installation capable of high-temperature, high-pressure experiments with a steam-air‑helium gas mixture. This paper presents key findings from a comprehensive experimental campaign with CIGMA. The JT-SJ series demonstrated the effectiveness of external surface cooling in suppressing top head flange overheating. The CC-SP series revealed spray-induced mixing mechanisms that rapidly homogenize flammable stratifications. The CC-PL series identified condensation processes of the gas mixture that are decisive for containment cooling strategies. Finally, the CC-SJ series provided insights into inter-compartment gas transport relevant to the multi-stage explosions in Unit 3 of Fukushima Daiichi. These results establish a high-fidelity experimental database, offering benchmarks for CFD validation and advancing the development of robust hydrogen mitigation and accident management strategies worldwide.
{"title":"CIGMA experiments on integral phenomena related to thermal hydraulics in a reactor containment vessel and building during a severe accident","authors":"Satoshi Abe, Ari Hamdani, Shu Soma, Ryosuke Hangai, Masashi Ohmori, Akihiko Ohwada, Toshihito Ohmiya, Yasuteru Sibamoto","doi":"10.1016/j.nucengdes.2026.114787","DOIUrl":"10.1016/j.nucengdes.2026.114787","url":null,"abstract":"<div><div>The Fukushima Daiichi accident underscored the urgent need to understand complex thermal-hydraulic phenomena governing containment integrity and gas mixture distribution during a severe accident. In response, the Japan Atomic Energy Agency (JAEA) established the CIGMA (Containment InteGral Measurement Apparatus) facility, a flagship large-scale installation capable of high-temperature, high-pressure experiments with a steam-air‑helium gas mixture. This paper presents key findings from a comprehensive experimental campaign with CIGMA. The JT-SJ series demonstrated the effectiveness of external surface cooling in suppressing top head flange overheating. The CC-SP series revealed spray-induced mixing mechanisms that rapidly homogenize flammable stratifications. The CC-PL series identified condensation processes of the gas mixture that are decisive for containment cooling strategies. Finally, the CC-SJ series provided insights into inter-compartment gas transport relevant to the multi-stage explosions in Unit 3 of Fukushima Daiichi. These results establish a high-fidelity experimental database, offering benchmarks for CFD validation and advancing the development of robust hydrogen mitigation and accident management strategies worldwide.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114787"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036057","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: 2026-01-20DOI: 10.1016/j.nucengdes.2026.114773
Jacob A. Hirschhorn, Mustafa K. Jaradat, Ryan T. Sweet, Nicolas E. Woolstenhulme, Paul A. Demkowicz, David A. Reger, Paolo Balestra, Gerhard Strydom
The United States nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles’ silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.
{"title":"Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT","authors":"Jacob A. Hirschhorn, Mustafa K. Jaradat, Ryan T. Sweet, Nicolas E. Woolstenhulme, Paul A. Demkowicz, David A. Reger, Paolo Balestra, Gerhard Strydom","doi":"10.1016/j.nucengdes.2026.114773","DOIUrl":"10.1016/j.nucengdes.2026.114773","url":null,"abstract":"<div><div>The United States nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles’ silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114773"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036058","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: 2026-02-09DOI: 10.1016/j.nucengdes.2026.114814
Xinhe Qu, Yi Huang, Wei Peng, Heng Xie
Many research reactors use a nitrogen pressurization system, where nitrogen is in direct contact with the primary coolant in the volume compensator, and the dissolved nitrogen molecules and the water molecules undergo radiolytic reactions such as ionization and excitation under the action of high-energy rays. The radiolytic products may affect the analysis and control of the primary coolant chemistry. Therefore, it is necessary to determine the solubility of nitrogen in the primary circuit. This study employed Aspen Plus to investigate and analyze nitrogen dissolution behavior in the primary circuit coolant of high flux reactors. First, the accuracy of the calculation model was validated by calculating the saturated solubility of nitrogen. Furthermore, steady-state simulations were performed for the primary circuit of a high flux reactor including a jet degassing process. Moreover, the key factors affecting the amount of nitrogen dissolved in the primary circuit, including the fluctuating flow rate in the surge line, the operating pressure of the degasser, the mass flow in the purification and degasification circuit (PDC), and the amount of nitrogen replenishment, were analyzed. The results indicate that nitrogen ingress from the volume compensator into the reactor pressure vessel and primary circuit occurs predominantly through convective flow, with diffusion playing a negligible role. The fluctuating flow rate in the surge line and the flow rate in the PDC have a significant influence on the amount of nitrogen dissolved in the primary circuit. Under typical analysis conditions, with a degasser pressure of 0.26 bar, a PDC flow rate at 1% of the primary circuit flow rate, and a fluctuating flow rate in the surge line of 0.1 kg/s, the nitrogen concentration in the primary circuit reaches 4.10 ppm—significantly below the saturated nitrogen solubility level. The results of this study have significant importance for the design of coolant pH control and the safety analysis of the high-flux reactors.
{"title":"Study on nitrogen dissolution characteristics in the nitrogen pressurization system of high flux reactors","authors":"Xinhe Qu, Yi Huang, Wei Peng, Heng Xie","doi":"10.1016/j.nucengdes.2026.114814","DOIUrl":"10.1016/j.nucengdes.2026.114814","url":null,"abstract":"<div><div>Many research reactors use a nitrogen pressurization system, where nitrogen is in direct contact with the primary coolant in the volume compensator, and the dissolved nitrogen molecules and the water molecules undergo radiolytic reactions such as ionization and excitation under the action of high-energy rays. The radiolytic products may affect the analysis and control of the primary coolant chemistry. Therefore, it is necessary to determine the solubility of nitrogen in the primary circuit. This study employed Aspen Plus to investigate and analyze nitrogen dissolution behavior in the primary circuit coolant of high flux reactors. First, the accuracy of the calculation model was validated by calculating the saturated solubility of nitrogen. Furthermore, steady-state simulations were performed for the primary circuit of a high flux reactor including a jet degassing process. Moreover, the key factors affecting the amount of nitrogen dissolved in the primary circuit, including the fluctuating flow rate in the surge line, the operating pressure of the degasser, the mass flow in the purification and degasification circuit (PDC), and the amount of nitrogen replenishment, were analyzed. The results indicate that nitrogen ingress from the volume compensator into the reactor pressure vessel and primary circuit occurs predominantly through convective flow, with diffusion playing a negligible role. The fluctuating flow rate in the surge line and the flow rate in the PDC have a significant influence on the amount of nitrogen dissolved in the primary circuit. Under typical analysis conditions, with a degasser pressure of 0.26 bar, a PDC flow rate at 1% of the primary circuit flow rate, and a fluctuating flow rate in the surge line of 0.1 kg/s, the nitrogen concentration in the primary circuit reaches 4.10 ppm—significantly below the saturated nitrogen solubility level. The results of this study have significant importance for the design of coolant pH control and the safety analysis of the high-flux reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114814"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190255","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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