Pub 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":"2026-01-23","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 : 2026-01-22DOI: 10.1016/j.nucengdes.2026.114788
Xueyan Zhang, Yixuan Zhang, Jun Yang
Employing system-code simulations, the study investigates scaling effects on transient simulations, uncertainty and sensitivity quantification, and data applicability analysis between the Purdue University Multi-Dimensional Integral Test Assembly and the Economic Simplified Boiling Water Reactor. The analysis focuses on refining Best Estimate Plus Uncertainty methodologies for nuclear reactor safety by evaluating the impact of scaling between test facilities and their prototypes. Transient simulation results highlight the importance of precise scaling in replicating system behaviours during small break loss-of-coolant accidents, with particular emphasis on the discrepancies in dynamic responses. In uncertainty and sensitivity quantification, variations due to scaling notably influence the magnitude and correlation of critical parameters such as the Depressurization Valve discharge coefficient and core thermal hydraulic diameter, underscoring the necessity for accurate model scaling in safety assessments. Furthermore, data applicability analysis, bolstered by dimensionless number evaluations, reveals essential insights into the extent to which scaled experimental data can mirror prototype phenomena, thereby emphasizing the pivotal role of scaling in experimental setups for nuclear safety analysis. Collectively, these findings advance the accuracy of predictive safety evaluations and contribute significantly to the enhancement of nuclear safety standards and methodologies.
{"title":"Scaling sensitivity and data applicability analysis of small-scale integral test to the ESBWR small break loss-of-coolant accidents","authors":"Xueyan Zhang, Yixuan Zhang, Jun Yang","doi":"10.1016/j.nucengdes.2026.114788","DOIUrl":"10.1016/j.nucengdes.2026.114788","url":null,"abstract":"<div><div>Employing system-code simulations, the study investigates scaling effects on transient simulations, uncertainty and sensitivity quantification, and data applicability analysis between the Purdue University Multi-Dimensional Integral Test Assembly and the Economic Simplified Boiling Water Reactor. The analysis focuses on refining Best Estimate Plus Uncertainty methodologies for nuclear reactor safety by evaluating the impact of scaling between test facilities and their prototypes. Transient simulation results highlight the importance of precise scaling in replicating system behaviours during small break loss-of-coolant accidents, with particular emphasis on the discrepancies in dynamic responses. In uncertainty and sensitivity quantification, variations due to scaling notably influence the magnitude and correlation of critical parameters such as the Depressurization Valve discharge coefficient and core thermal hydraulic diameter, underscoring the necessity for accurate model scaling in safety assessments. Furthermore, data applicability analysis, bolstered by dimensionless number evaluations, reveals essential insights into the extent to which scaled experimental data can mirror prototype phenomena, thereby emphasizing the pivotal role of scaling in experimental setups for nuclear safety analysis. Collectively, these findings advance the accuracy of predictive safety evaluations and contribute significantly to the enhancement of nuclear safety standards and methodologies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114788"},"PeriodicalIF":2.1,"publicationDate":"2026-01-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036027","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-21DOI: 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":"2026-01-21","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 : 2026-01-21DOI: 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":"2026-01-21","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}
Fully Ceramic Microencapsulated (FCM®) fuels produced by sintering has been demonstrated in a method that does not damage fuel particles, due to assembling particles into layers. Pilot manufacturing data (N = 32) of the designed fuel is presented. Screening of SiC powders indicated that sintered densities are a function of powder size and compaction pressure. Abiding particles are postulated to prevent interparticle infiltration of powder. Image analysis of X-ray Computed Tomography (XCT) and optical microscopy cross-sections show a trend of decreasing matrix density between particles, and density is underestimated by metallographic methods. Interparticle porosity was always observed, limiting matrix density (86–97 %). Appropriate green forming can ensure all fired porosity is closed against He, water, and O2 penetration. TRISO packing fractions were between 36 and 38 % and the packing has room for improvement. For industry adoption, development of radiation-stable, pressureless sintering of SiC is recommended.
{"title":"Design and pilot production of Fully Ceramic Microencapsulated (FCM®) fuels fabricated by sintering","authors":"Cameron Hilliard , Ethan Deters , Mason Phillips , Caen Ang","doi":"10.1016/j.nucengdes.2025.114745","DOIUrl":"10.1016/j.nucengdes.2025.114745","url":null,"abstract":"<div><div>Fully Ceramic Microencapsulated (FCM®) fuels produced by sintering has been demonstrated in a method that does not damage fuel particles, due to assembling particles into layers. Pilot manufacturing data (<em>N</em> = 32) of the designed fuel is presented. Screening of SiC powders indicated that sintered densities are a function of powder size and compaction pressure. Abiding particles are postulated to prevent interparticle infiltration of powder. Image analysis of X-ray Computed Tomography (XCT) and optical microscopy cross-sections show a trend of decreasing matrix density between particles, and density is underestimated by metallographic methods. Interparticle porosity was always observed, limiting matrix density (86–97 %). Appropriate green forming can ensure all fired porosity is closed against He, water, and O<sub>2</sub> penetration. TRISO packing fractions were between 36 and 38 % and the packing has room for improvement. For industry adoption, development of radiation-stable, pressureless sintering of SiC is recommended.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114745"},"PeriodicalIF":2.1,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036025","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-20DOI: 10.1016/j.nucengdes.2026.114768
Fatima Ghandour , Salah Hamieh , Ziad Francis
This study investigates the neutronic performance of dual cooled annular fuel rods in the CAREM 25 integral Pressurized Water Reactor (iPWR), a small modular reactor (SMR), using MCNP5 Monte Carlo simulations. The motivation is to reduce power peaking factors (PPFs) and enhance thermal-hydraulic safety margins by adopting an annular fuel geometry with internal and external cooling. Three annular fuel configurations with 100%, 95%, and 93% fuel loading were analyzed and compared to the conventional solid fuel design. Geometric transformations were performed analytically—introducing, for the first time, closed-form equations for the inner and outer radii of annular fuel rods—to maintain the fuel-to-coolant volume ratio while limiting fuel mass reduction to ≤10%. The results show that the total PPF decreased by up to 27.45% in the 95% fuel loading case, dropping from 2.404 (solid design) to 1.744. Additionally, the effective multiplication factor (Keff) was reduced from 1.12445 to 1.09512, enhancing reactor controllability. The 95% loading configuration emerged as the optimal design, balancing neutronic performance and safety. These findings demonstrate that annular fuel can significantly flatten the power distribution and improve the safety profile of iPWR SMRs without compromising core performance.
{"title":"Optimizing iPWR SMR core design: a power peaking factor analysis of annular fuel rods using MCNP5","authors":"Fatima Ghandour , Salah Hamieh , Ziad Francis","doi":"10.1016/j.nucengdes.2026.114768","DOIUrl":"10.1016/j.nucengdes.2026.114768","url":null,"abstract":"<div><div>This study investigates the neutronic performance of dual cooled annular fuel rods in the CAREM 25 integral Pressurized Water Reactor (iPWR), a small modular reactor (SMR), using MCNP5 Monte Carlo simulations. The motivation is to reduce power peaking factors (PPFs) and enhance thermal-hydraulic safety margins by adopting an annular fuel geometry with internal and external cooling. Three annular fuel configurations with 100%, 95%, and 93% fuel loading were analyzed and compared to the conventional solid fuel design. Geometric transformations were performed analytically—introducing, for the first time, closed-form equations for the inner and outer radii of annular fuel rods—to maintain the fuel-to-coolant volume ratio while limiting fuel mass reduction to ≤10%. The results show that the total PPF decreased by up to 27.45% in the 95% fuel loading case, dropping from 2.404 (solid design) to 1.744. Additionally, the effective multiplication factor (Keff) was reduced from 1.12445 to 1.09512, enhancing reactor controllability. The 95% loading configuration emerged as the optimal design, balancing neutronic performance and safety. These findings demonstrate that annular fuel can significantly flatten the power distribution and improve the safety profile of iPWR SMRs without compromising core performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114768"},"PeriodicalIF":2.1,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036073","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2026-01-20DOI: 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":"2026-01-20","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 : 2026-01-20DOI: 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":"2026-01-20","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 : 2026-01-20DOI: 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":"2026-01-20","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 : 2026-01-19DOI: 10.1016/j.nucengdes.2026.114782
Chengrui Zhang , Juan Chen
The conjugate heat transfer characteristics of ballooning accident-tolerant fuel (ATF) rods in supercritical water were systematically investigated. A two-dimensional numerical multi-physics conjugate heat transfer program was developed for a 2 × 2 rod bundle with blocking sleeve and spacer grids, based on a comprehensive review of experimental and simulation studies on blocked flow, with particular emphasis on correlations for the friction coefficient and convective heat transfer coefficient. The program integrates a two-dimensional thermal conductivity model for ballooning fuel rod, solved by the finite difference method, with the simulation of convective heat transfer between the deformed cladding and coolant using various heat transfer coefficient correlations. Flow distribution between blocked and unblocked channels containing spacer grids is also simulated under supercritical pressure conditions, also employing the finite difference method with different friction coefficient correlations applied. Validation against SWAMUP experimental data shows that the relative error of the calculated axial cladding surface temperature remains within 1.6 %. Subsequently, the effects of blockage ratio, heat flux, and system pressure on heat transfer performance were analyzed. Key findings include: 1) the convective heat transfer coefficient downstream of the blockage region generally increases owing to enhanced turbulence, particularly at low swelling ratio. 2) When the swelling ratio exceeds 55 %, friction pressure drop contributes up to 87.1 % of the total pressure drop, leading to heat transfer deterioration. 3) Higher system pressures result in a greater total pressure drop and more uniform flow distribution between blocked and unblocked channels. 4) Heat transfer deterioration can occur when the inlet temperature approaches the pseudo-critical temperature. This research provides theoretical support for the design optimization and safe operation of SCWR.
{"title":"Development of conjugate heat transfer coupling model for supercritical water flowing in 2 × 2 ballooning ATF rod bundles","authors":"Chengrui Zhang , Juan Chen","doi":"10.1016/j.nucengdes.2026.114782","DOIUrl":"10.1016/j.nucengdes.2026.114782","url":null,"abstract":"<div><div>The conjugate heat transfer characteristics of ballooning accident-tolerant fuel (ATF) rods in supercritical water were systematically investigated. A two-dimensional numerical multi-physics conjugate heat transfer program was developed for a 2 × 2 rod bundle with blocking sleeve and spacer grids, based on a comprehensive review of experimental and simulation studies on blocked flow, with particular emphasis on correlations for the friction coefficient and convective heat transfer coefficient. The program integrates a two-dimensional thermal conductivity model for ballooning fuel rod, solved by the finite difference method, with the simulation of convective heat transfer between the deformed cladding and coolant using various heat transfer coefficient correlations. Flow distribution between blocked and unblocked channels containing spacer grids is also simulated under supercritical pressure conditions, also employing the finite difference method with different friction coefficient correlations applied. Validation against SWAMUP experimental data shows that the relative error of the calculated axial cladding surface temperature remains within 1.6 %. Subsequently, the effects of blockage ratio, heat flux, and system pressure on heat transfer performance were analyzed. Key findings include: 1) the convective heat transfer coefficient downstream of the blockage region generally increases owing to enhanced turbulence, particularly at low swelling ratio. 2) When the swelling ratio exceeds 55 %, friction pressure drop contributes up to 87.1 % of the total pressure drop, leading to heat transfer deterioration. 3) Higher system pressures result in a greater total pressure drop and more uniform flow distribution between blocked and unblocked channels. 4) Heat transfer deterioration can occur when the inlet temperature approaches the pseudo-critical temperature. This research provides theoretical support for the design optimization and safe operation of SCWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114782"},"PeriodicalIF":2.1,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036055","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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