Pub Date : 2025-12-11DOI: 10.1016/j.nucengdes.2025.114643
Siqin Hu , Chong Zhou , Jian Tian , Yuhan Fan , Shanwu Wang , Xiaohan Yu
The Small Modular Liquid Fuel Molten Salt Reactor (SM-TMSR), a novel type of fourth-generation reactor, combines the advantages of liquid fuel molten salt reactors and small modular reactors, offering enhanced safety and economic benefits. This study systematically investigates the thermal–hydraulic design and steady-state characteristics of a 150 MWt SM-TMSR core, based on the SM-TMSR concept and the design requirements for a 100 MW-class liquid fuel molten salt reactor. The study addresses the deterioration of narrow-gap heat transfer in traditional graphite assemblies and proposes an optimized novel dual-cooled assembly design based on a self-developed thermal–hydraulic analysis program, the results revealed that the External Fuel Volume Fraction (EVF) is a key parameter influencing the thermal–hydraulic performance of the assembly. The three-dimensional core analysis model is established using Computational Fluid Dynamics (CFD) tool Fluent, and a simulation study on core flow distribution design and core steady-state thermal transfer characteristics was conducted. A set of optimized core structural parameters meeting the design requirements was ultimately proposed.
{"title":"Novel dual-cooled assembly design and thermal hydraulic analysis for a 150 MWt small modular molten salt reactor core","authors":"Siqin Hu , Chong Zhou , Jian Tian , Yuhan Fan , Shanwu Wang , Xiaohan Yu","doi":"10.1016/j.nucengdes.2025.114643","DOIUrl":"10.1016/j.nucengdes.2025.114643","url":null,"abstract":"<div><div>The Small Modular Liquid Fuel Molten Salt Reactor (SM-TMSR), a novel type of fourth-generation reactor, combines the advantages of liquid fuel molten salt reactors and small modular reactors, offering enhanced safety and economic benefits. This study systematically investigates the thermal–hydraulic design and steady-state characteristics of a 150 MWt SM-TMSR core, based on the SM-TMSR concept and the design requirements for a 100 MW-class liquid fuel molten salt reactor. The study addresses the deterioration of narrow-gap heat transfer in traditional graphite assemblies and proposes an optimized novel dual-cooled assembly design based on a self-developed thermal–hydraulic analysis program, the results revealed that the External Fuel Volume Fraction (EVF) is a key parameter influencing the thermal–hydraulic performance of the assembly. The three-dimensional core analysis model is established using Computational Fluid Dynamics (CFD) tool Fluent, and a simulation study on core flow distribution design and core steady-state thermal transfer characteristics was conducted. A set of optimized core structural parameters meeting the design requirements was ultimately proposed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114643"},"PeriodicalIF":2.1,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734872","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 : 2025-12-11DOI: 10.1016/j.nucengdes.2025.114669
Liang Zhang, Wei Liu, Koji Morita
3D particle-based simulations were performed to investigate the melting behavior of solid boron carbide below its melting point, resulting from the eutectic reaction with liquid stainless steel. The experiments conducted by the Japan Atomic Energy Agency under isothermal conditions using three different temperatures were analyzed by simulation herein. The smoothed particle hydrodynamics (SPH) method was employed, with an implicit incompressible SPH pressure solver and a series of mathematical models for viscosity, surface tension, heat transfer, and mass diffusion based on Fick's law. The values of the diffusion coefficients that were applied to the simulations to obtain reaction rates similar to those of the experiments followed an exponential relationship with temperature. The simulations allowed us to quantify the reaction rates and examine the effect of convection, which appears to accelerate the eutectic reaction. In addition, the effect of particle size on the SPH simulation results was discussed theoretically.
{"title":"SPH simulations reproducing the eutectic melting behavior of solid boron carbide immersed in liquid stainless steel","authors":"Liang Zhang, Wei Liu, Koji Morita","doi":"10.1016/j.nucengdes.2025.114669","DOIUrl":"10.1016/j.nucengdes.2025.114669","url":null,"abstract":"<div><div>3D particle-based simulations were performed to investigate the melting behavior of solid boron carbide below its melting point, resulting from the eutectic reaction with liquid stainless steel. The experiments conducted by the Japan Atomic Energy Agency under isothermal conditions using three different temperatures were analyzed by simulation herein. The smoothed particle hydrodynamics (SPH) method was employed, with an implicit incompressible SPH pressure solver and a series of mathematical models for viscosity, surface tension, heat transfer, and mass diffusion based on Fick's law. The values of the diffusion coefficients that were applied to the simulations to obtain reaction rates similar to those of the experiments followed an exponential relationship with temperature. The simulations allowed us to quantify the reaction rates and examine the effect of convection, which appears to accelerate the eutectic reaction. In addition, the effect of particle size on the SPH simulation results was discussed theoretically.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114669"},"PeriodicalIF":2.1,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734871","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 : 2025-12-11DOI: 10.1016/j.nucengdes.2025.114645
Muthuraj T. , John Arul , Aditya Bhandari
Sodium void formation is a major source of instability in medium and large-sized sodium-cooled fast reactors (SFRs), as it introduces positive reactivity feedback that depends on the void location and propagation behavior. Such voids may arise during global power-to-flow imbalance events, including unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP), as well as during localized phenomena such as fission gas release, flow blockage, or coolant boiling. Previous bifurcation studies have shown that void propagation can lead to nonlinear power oscillations and even chaotic behavior, highlighting the importance of rigorous dynamic analysis under these conditions. Quantifying the perturbation worth of key state variables is therefore essential for establishing inherent safety margins, as reactivity coefficients govern the reactor’s dynamic response to changes in temperature, density, and coolant composition. Unlike thermal reactors — where negative fuel and moderator temperature coefficients strongly enhance stability — fast reactors exhibit different behavior due to the absence of a moderator. In this study, the Lyapunov function method is applied to assess the nonlinear stability of a 500 MWe MOX-fueled SFR under sodium voiding conditions, and to determine the domain of attraction (DOA) that defines safe perturbation limits for the reactor state variables.
{"title":"Determination of domain of attraction of sodium-cooled fast reactor core with sodium void propagation using Lyapunov function method and particle swarm optimization","authors":"Muthuraj T. , John Arul , Aditya Bhandari","doi":"10.1016/j.nucengdes.2025.114645","DOIUrl":"10.1016/j.nucengdes.2025.114645","url":null,"abstract":"<div><div>Sodium void formation is a major source of instability in medium and large-sized sodium-cooled fast reactors (SFRs), as it introduces positive reactivity feedback that depends on the void location and propagation behavior. Such voids may arise during global power-to-flow imbalance events, including unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP), as well as during localized phenomena such as fission gas release, flow blockage, or coolant boiling. Previous bifurcation studies have shown that void propagation can lead to nonlinear power oscillations and even chaotic behavior, highlighting the importance of rigorous dynamic analysis under these conditions. Quantifying the perturbation worth of key state variables is therefore essential for establishing inherent safety margins, as reactivity coefficients govern the reactor’s dynamic response to changes in temperature, density, and coolant composition. Unlike thermal reactors — where negative fuel and moderator temperature coefficients strongly enhance stability — fast reactors exhibit different behavior due to the absence of a moderator. In this study, the Lyapunov function method is applied to assess the nonlinear stability of a 500 MWe MOX-fueled SFR under sodium voiding conditions, and to determine the domain of attraction (DOA) that defines safe perturbation limits for the reactor state variables.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114645"},"PeriodicalIF":2.1,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734870","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 : 2025-12-10DOI: 10.1016/j.nucengdes.2025.114671
Junnan Zhang, Lu Li, Qingning Zuo, Xiaoying Xiong, Qiming Wei, Xiaoping Bai
Point-kinetics (PK) dynamics are stiff and, under real-time and hardware-in-the-loop (RT/HIL) constraints, necessitate fixed-step time integrators whose performance has been evaluated only sporadically in heterogeneous settings. This study introduces a unified benchmarking protocol and systematically compares explicit Runge–Kutta (RK) integrators (Heun, Ralston3, RK4) with the implicit -method (with between 0.5 and 1) across accuracy, stability, and real-time feasibility. The protocol specifies shared scenarios and events; common sampling and time alignment; normalized root-mean-square error (NRMSE); peak magnitude and timing errors; real-time factor (RTF); work–accuracy analysis; availability maps; and Newton-iteration statistics. Results indicate that, for smooth transients, explicit methods achieve their theoretical orders. The efficiency frontier shifts with the target error: explicit integrators dominate under moderate accuracy requirements and weak stiffness, whereas higher- implicit schemes (approaching Backward Euler) become preferable under stricter accuracy goals or stronger stiffness by allowing larger stable steps. Peak metrics are sensitive to sampling and interpolation; enforcing unified alignment and monotone-limited interpolation substantially reduces spurious peaks and timing bias. A practical step-size policy selects the larger of the order-derived step and the real-time-limited step to reconcile accuracy with real-time constraints. The protocol and findings provide reproducible guidance for integrator selection and parameterization in real-time reactor simulators and nuclear digital twins.
{"title":"Real-time point-kinetics: Empirical laws, frontiers, and energy consistency","authors":"Junnan Zhang, Lu Li, Qingning Zuo, Xiaoying Xiong, Qiming Wei, Xiaoping Bai","doi":"10.1016/j.nucengdes.2025.114671","DOIUrl":"10.1016/j.nucengdes.2025.114671","url":null,"abstract":"<div><div>Point-kinetics (PK) dynamics are stiff and, under real-time and hardware-in-the-loop (RT/HIL) constraints, necessitate fixed-step time integrators whose performance has been evaluated only sporadically in heterogeneous settings. This study introduces a unified benchmarking protocol and systematically compares explicit Runge–Kutta (RK) integrators (Heun, Ralston3, RK4) with the implicit <span><math><mi>θ</mi></math></span>-method (with <span><math><mi>θ</mi></math></span> between 0.5 and 1) across accuracy, stability, and real-time feasibility. The protocol specifies shared scenarios and events; common sampling and time alignment; normalized root-mean-square error (NRMSE); peak magnitude and timing errors; real-time factor (RTF); work–accuracy analysis; <span><math><mrow><mo>(</mo><mi>θ</mi><mo>,</mo><mi>Δ</mi><mi>t</mi><mo>)</mo></mrow></math></span> availability maps; and Newton-iteration statistics. Results indicate that, for smooth transients, explicit methods achieve their theoretical orders. The efficiency frontier shifts with the target error: explicit integrators dominate under moderate accuracy requirements and weak stiffness, whereas higher-<span><math><mi>θ</mi></math></span> implicit schemes (approaching Backward Euler) become preferable under stricter accuracy goals or stronger stiffness by allowing larger stable steps. Peak metrics are sensitive to sampling and interpolation; enforcing unified alignment and monotone-limited interpolation substantially reduces spurious peaks and timing bias. A practical step-size policy selects the larger of the order-derived step and the real-time-limited step to reconcile accuracy with real-time constraints. The protocol and findings provide reproducible guidance for integrator selection and parameterization in real-time reactor simulators and nuclear digital twins.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114671"},"PeriodicalIF":2.1,"publicationDate":"2025-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734884","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}
This study proposes an innovative reactor configuration termed the Liquid Metal Matrix-Based Heat Pipe-Cooled Reactor (LM-HPR). The design aims to address two primary challenges in advanced reactor systems: eliminating the additional solid-solid contact thermal resistance between solid matrix materials and high-temperature heat pipe (HTHP) surfaces, and concurrently mitigating the high-temperature dynamic corrosion issues commonly associated with lead‑bismuth fast reactors. The heat transfer performance of this novel configuration was investigated using a dedicated lead‑bismuth/heat pipe heat exchange prototype (LBE-HPHE Prototype). A specialized experimental platform was constructed to measure critical physical parameters of both the HTHPs and the Lead-Bismuth Eutectic (LBE), facilitating a detailed examination of the heat exchange dynamics between these two media. Initial tests conducted on potassium HTHPs successfully verified stable startup characteristics and excellent isothermal performance. A series of five steady-state experiments were performed under varying LBE temperature conditions. During the final testing phase, the HTHPs exhibited an average axial temperature gradient of 43.82 °C and demonstrated a heat extraction capacity of 2.42 kW from the LBE pool. The average equivalent thermal resistance of the HTHPs was calculated to be 0.0516 K·W−1, indicating good overall heat transfer characteristics alongside their confirmed isothermal performance. The average temperature difference measured between the bottom and top layers of the LBE was 5.90 °C. Analysis revealed that as the LBE temperature increased, the interlayer temperature difference initially decreased before subsequently increasing, a trend attributed to the evolving performance characteristics of the HTHPs under different operational conditions. With increasing LBE temperature, the heat exchange process between the HTHPs and the LBE intensified significantly. The natural convection heat transfer coefficient exhibited a substantial enhancement, rising from 327.32 W/(m2·°C) to 8431.65 W/(m2·°C). This improvement corresponded with increased heat transfer efficiency of the HTHPs and stronger natural convection heat transfer between the LBE and the HTHPs. Consequently, the Nusselt (Nu) number increased, while the Rayleigh (Ra) number also showed a gradual rise accompanying the elevation in LBE temperature. These experimental outcomes validate the fundamental feasibility of the liquid metal matrix-based heat pipe-cooled microreactor concept, providing crucial empirical data to support its further development and potential deployment.
{"title":"Experimental study on the heat transfer characteristics of a lead-bismuth/heat pipe heat exchange prototype for liquid metal matrix-based heat pipe-cooled microreactors","authors":"Jiarui Zhang, Zihao Hei, Chenglong Wang, Zeqin Zhang, Kailun Guo, Wenxi Tian, Suizheng Qiu, Guanghui Su","doi":"10.1016/j.nucengdes.2025.114656","DOIUrl":"10.1016/j.nucengdes.2025.114656","url":null,"abstract":"<div><div>This study proposes an innovative reactor configuration termed the Liquid Metal Matrix-Based Heat Pipe-Cooled Reactor (LM-HPR). The design aims to address two primary challenges in advanced reactor systems: eliminating the additional solid-solid contact thermal resistance between solid matrix materials and high-temperature heat pipe (HTHP) surfaces, and concurrently mitigating the high-temperature dynamic corrosion issues commonly associated with lead‑bismuth fast reactors. The heat transfer performance of this novel configuration was investigated using a dedicated lead‑bismuth/heat pipe heat exchange prototype (LBE-HPHE Prototype). A specialized experimental platform was constructed to measure critical physical parameters of both the HTHPs and the Lead-Bismuth Eutectic (LBE), facilitating a detailed examination of the heat exchange dynamics between these two media. Initial tests conducted on potassium HTHPs successfully verified stable startup characteristics and excellent isothermal performance. A series of five steady-state experiments were performed under varying LBE temperature conditions. During the final testing phase, the HTHPs exhibited an average axial temperature gradient of 43.82 °C and demonstrated a heat extraction capacity of 2.42 kW from the LBE pool. The average equivalent thermal resistance of the HTHPs was calculated to be 0.0516 K·W<sup>−1</sup>, indicating good overall heat transfer characteristics alongside their confirmed isothermal performance. The average temperature difference measured between the bottom and top layers of the LBE was 5.90 °C. Analysis revealed that as the LBE temperature increased, the interlayer temperature difference initially decreased before subsequently increasing, a trend attributed to the evolving performance characteristics of the HTHPs under different operational conditions. With increasing LBE temperature, the heat exchange process between the HTHPs and the LBE intensified significantly. The natural convection heat transfer coefficient exhibited a substantial enhancement, rising from 327.32 W/(m<sup>2</sup>·°C) to 8431.65 W/(m<sup>2</sup>·°C). This improvement corresponded with increased heat transfer efficiency of the HTHPs and stronger natural convection heat transfer between the LBE and the HTHPs. Consequently, the Nusselt (<em>Nu</em>) number increased, while the Rayleigh (<em>Ra</em>) number also showed a gradual rise accompanying the elevation in LBE temperature. These experimental outcomes validate the fundamental feasibility of the liquid metal matrix-based heat pipe-cooled microreactor concept, providing crucial empirical data to support its further development and potential deployment.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114656"},"PeriodicalIF":2.1,"publicationDate":"2025-12-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734880","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 : 2025-12-09DOI: 10.1016/j.nucengdes.2025.114672
Bryan Tan , Yu Duan , Michael J. Bluck , Matthew D. Eaton , Bo Liu
The novel coarse-mesh CFD tool Subchannel-CFD (SubChCFD) is coupled with the Monte Carlo transport code Serpent using a segregated multi-physics coupling algorithm. The multi-physics model is validated using the CASL VERA Core Physics Benchmark Progression Problem #6, showing good agreement of pin powers and exit coolant temperatures. The model is also used to simulate an assembly from the K-SMR soluble‑boron-free small modular reactor, which shows good agreement with a full-scale CFD model from the software STAR-CCM+, with the exception of a small gap region between two large control rod guide tube and burnable poison rods, which induces a periodic flow instability that SubChCFD cannot fully replicate. Nevertheless, the oscillatory characteristics of the instabilities modelled by both codes show some similarity.
{"title":"Validation and application of coupled thermal-hydraulics and neutronics model using sub-channel CFD and SERPENT","authors":"Bryan Tan , Yu Duan , Michael J. Bluck , Matthew D. Eaton , Bo Liu","doi":"10.1016/j.nucengdes.2025.114672","DOIUrl":"10.1016/j.nucengdes.2025.114672","url":null,"abstract":"<div><div>The novel coarse-mesh CFD tool Subchannel-CFD (SubChCFD) is coupled with the Monte Carlo transport code Serpent using a segregated multi-physics coupling algorithm. The multi-physics model is validated using the CASL VERA Core Physics Benchmark Progression Problem #6, showing good agreement of pin powers and exit coolant temperatures. The model is also used to simulate an assembly from the K-SMR soluble‑boron-free small modular reactor, which shows good agreement with a full-scale CFD model from the software STAR-CCM+, with the exception of a small gap region between two large control rod guide tube and burnable poison rods, which induces a periodic flow instability that SubChCFD cannot fully replicate. Nevertheless, the oscillatory characteristics of the instabilities modelled by both codes show some similarity.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114672"},"PeriodicalIF":2.1,"publicationDate":"2025-12-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145735253","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 : 2025-12-09DOI: 10.1016/j.nucengdes.2025.114664
Pierre-Louis Ardizzone, Clément Lafond, Yann De Carlan, Jérôme Garnier
The use of coated structural materials in Molten Salt Reactors (MSR) enables surfaces to be functionalized. This would allow to obtain a corrosion and irradiation resistant component. This study explores the mechanical and microstructural properties of a 9Cr-1Mo (grade 91) martensitic steel substrate coated with Inconel 625. Tensile tests were conducted at ambient (20 °C) and high temperatures (400 °C, 600 °C, and 700 °C). The interface’s tensile strength was evaluated, revealing that failures consistently occurred within the substrate rather than at the interface. This indicates strong adhesion between the coating and substrate. Microstructural analyses were performed, identifying five distinct regions within the Heat Affected Zone (HAZ). These regions included rapidly quenched martensite and residual ferrite, which influenced the mechanical properties observed during testing. Additionally, the solidification of the Inconel 625 coating is discussed. It is proposed that Inconel 625 grows epitaxially on the substrate (P91) while it is still in the austenitic phase during cooling of the molten zone. This could explain the good mechanical strength of the bond.
{"title":"Mechanical and microstructural characterization of 9Cr-1Mo steel with Inconel 625 thick coating","authors":"Pierre-Louis Ardizzone, Clément Lafond, Yann De Carlan, Jérôme Garnier","doi":"10.1016/j.nucengdes.2025.114664","DOIUrl":"10.1016/j.nucengdes.2025.114664","url":null,"abstract":"<div><div>The use of coated structural materials in Molten Salt Reactors (MSR) enables surfaces to be functionalized. This would allow to obtain a corrosion and irradiation resistant component. This study explores the mechanical and microstructural properties of a 9Cr-1Mo (grade 91) martensitic steel substrate coated with Inconel 625. Tensile tests were conducted at ambient (20 °C) and high temperatures (400 °C, 600 °C, and 700 °C). The interface’s tensile strength was evaluated, revealing that failures consistently occurred within the substrate rather than at the interface. This indicates strong adhesion between the coating and substrate. Microstructural analyses were performed, identifying five distinct regions within the Heat Affected Zone (HAZ). These regions included rapidly quenched martensite and residual <span><math><mi>δ</mi></math></span> ferrite, which influenced the mechanical properties observed during testing. Additionally, the solidification of the Inconel 625 coating is discussed. It is proposed that Inconel 625 grows epitaxially on the substrate (P91) while it is still in the austenitic phase during cooling of the molten zone. This could explain the good mechanical strength of the bond.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114664"},"PeriodicalIF":2.1,"publicationDate":"2025-12-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734883","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 : 2025-12-07DOI: 10.1016/j.nucengdes.2025.114646
Alberto Sáez Maderuelo , Francisco Javier Perosanz , Gaspar González-Doncel , Ricardo Fernández , Radek Novotny , Michal Novak , Thibaut de Terris
Small modular reactors (SMRs) have been positioned as a future alternative to provide electric power free of CO2 to the European grid. Among the different types of SMRs, the supercritical water-cooled reactors (SCWR) stand out. The pre-design of the supercritical water-SMR (SCW-SMR) has been funded by the European Commission through the ECC-SMART Project, highlighting the importance of this technology. A key pillar of SMRs, is the use of sustainable materials and advanced manufacturing procedures in their design. In this context, this study aims to explore the corrosion behavior of austenitic stainless steel AISI 316L, widely used in the manufacture of internal components of light water reactors (LWRs) currently in operation, under SCWR conditions. Because of its good performance under LWR operating conditions, it has been considered as a candidate structural material for the SCWR and the SCW-SMR.
A stainless steel AISI 316L 3 mm thick deposited layer manufactured by Cold Spray (CS) was studied using immersion corrosion tests. The tests were carried out for 1000 h in SCW at 380 °C and 23 MPa; i.e., in the vicinity of the critical point of water, of great interest from the perspective of material corrosion. Oxidation coupons were studied before and after the oxidation tests. The steel AISI 316L CS exhibits a complex microstructure with micrometric and nanometric grains, high roughness, and the presence of pores. No residual stresses were detected in the samples. After oxidation tests in supercritical conditions the material shows significantly lower weight gain compared to literature, indicating that the cold spray process may not have a negative impact oxidation resistance. An oxide layer with Fe on the outer and Cr in the inner side was observed with small areas where the oxide is detached from the metal base.
{"title":"Exploration of the corrosion behavior of cold-sprayed AISI 316L steel near the water critical point","authors":"Alberto Sáez Maderuelo , Francisco Javier Perosanz , Gaspar González-Doncel , Ricardo Fernández , Radek Novotny , Michal Novak , Thibaut de Terris","doi":"10.1016/j.nucengdes.2025.114646","DOIUrl":"10.1016/j.nucengdes.2025.114646","url":null,"abstract":"<div><div>Small modular reactors (SMRs) have been positioned as a future alternative to provide electric power free of CO<sub>2</sub> to the European grid. Among the different types of SMRs, the supercritical water-cooled reactors (SCWR) stand out. The pre-design of the supercritical water-SMR (SCW-SMR) has been funded by the European Commission through the ECC-SMART Project, highlighting the importance of this technology. A key pillar of SMRs, is the use of sustainable materials and advanced manufacturing procedures in their design. In this context, this study aims to explore the corrosion behavior of austenitic stainless steel AISI 316L, widely used in the manufacture of internal components of light water reactors (LWRs) currently in operation, under SCWR conditions. Because of its good performance under LWR operating conditions, it has been considered as a candidate structural material for the SCWR and the SCW-SMR.</div><div>A stainless steel AISI 316L 3 mm thick deposited layer manufactured by Cold Spray (CS) was studied using immersion corrosion tests. The tests were carried out for 1000 h in SCW at 380 °C and 23 MPa; <em>i.e.</em>, in the vicinity of the critical point of water, of great interest from the perspective of material corrosion. Oxidation coupons were studied before and after the oxidation tests. The steel AISI 316L CS exhibits a complex microstructure with micrometric and nanometric grains, high roughness, and the presence of pores. No residual stresses were detected in the samples. After oxidation tests in supercritical conditions the material shows significantly lower weight gain compared to literature, indicating that the cold spray process may not have a negative impact oxidation resistance. An oxide layer with Fe on the outer and Cr in the inner side was observed with small areas where the oxide is detached from the metal base.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114646"},"PeriodicalIF":2.1,"publicationDate":"2025-12-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145734881","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 : 2025-12-04DOI: 10.1016/j.nucengdes.2025.114620
John Beal, Spencer Fargusson, Ha Bui, Seyed Reihani, Hammad Khalid, Zahra Mohaghegh
The nuclear industry is investigating applications of Digital Twins (DTs) in nuclear power plants (NPPs), where real-time sensor data from a Physical Twin (PT) (i.e., a physical element) is utilized in the DT (i.e., a time-synchronized, virtual representation of the PT) to enhance operations and maintenance (O&M) decision-making and predictive maintenance. This paper develops an Integrated PRA (I-PRA) methodological framework, including a simulated high-fidelity PT model (i.e., Replicated PT), a physical degradation-based DT model, and a human model (i.e., simulation of O&M decision-making and a Human Reliability Analysis-based maintenance model). I-PRA incorporates simulation of the coupled human-PT-DT system into PRA, while characterizing and propagating epistemic and aleatory uncertainties associated with their inputs. Novel developments of I-PRA include: (i) bidirectional and dynamic interactions between the Replicated PT and human models, incorporating simulations of maintenance activities (e.g., repair and replacement) and their effects on the state of the Replicated PT in conjunction with the DT; and (ii) bidirectional integration of the coupled human-PT-DT system with PRA. Futuristic plant-level risk estimates are based on predicted component performance from DTs, providing forward-looking insights to inform both short-term actions (e.g., immediate maintenance) and long-term O&M decisions (e.g., Risk-Informed Performance-Based applications and regulation). Component reliabilities are estimated from the simulated output of the coupled human-PT-DT system, serving as inputs to the PRA and supporting short-term risk estimation (e.g., during plant mission time). A case study focuses on a NPP piping component subject to stress corrosion cracking using the Extremely Low Probability of Rupture (xLPR) Probabilistic Fracture Mechanics code.
{"title":"Developing temporal coupling of human performance, physical twin, and digital twin models for probabilistic risk assessment in nuclear power plants","authors":"John Beal, Spencer Fargusson, Ha Bui, Seyed Reihani, Hammad Khalid, Zahra Mohaghegh","doi":"10.1016/j.nucengdes.2025.114620","DOIUrl":"10.1016/j.nucengdes.2025.114620","url":null,"abstract":"<div><div>The nuclear industry is investigating applications of Digital Twins (DTs) in nuclear power plants (NPPs), where real-time sensor data from a Physical Twin (PT) (i.e., a physical element) is utilized in the DT (i.e., a time-synchronized, virtual representation of the PT) to enhance operations and maintenance (O&M) decision-making and predictive maintenance. This paper develops an Integrated PRA (I-PRA) methodological framework, including a simulated high-fidelity PT model (i.e., Replicated PT), a physical degradation-based DT model, and a human model (i.e., simulation of O&M decision-making and a Human Reliability Analysis-based maintenance model). I-PRA incorporates simulation of the coupled human-PT-DT system into PRA, while characterizing and propagating epistemic and aleatory uncertainties associated with their inputs. Novel developments of I-PRA include: (i) bidirectional and dynamic interactions between the Replicated PT and human models, incorporating simulations of maintenance activities (e.g., repair and replacement) and their effects on the state of the Replicated PT in conjunction with the DT; and (ii) bidirectional integration of the coupled human-PT-DT system with PRA. Futuristic plant-level risk estimates are based on predicted component performance from DTs, providing forward-looking insights to inform both short-term actions (e.g., immediate maintenance) and long-term O&M decisions (e.g., Risk-Informed Performance-Based applications and regulation). Component reliabilities are estimated from the simulated output of the coupled human-PT-DT system, serving as inputs to the PRA and supporting short-term risk estimation (e.g., during plant mission time). A case study focuses on a NPP piping component subject to stress corrosion cracking using the Extremely Low Probability of Rupture (xLPR) Probabilistic Fracture Mechanics code.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114620"},"PeriodicalIF":2.1,"publicationDate":"2025-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145659148","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 : 2025-12-04DOI: 10.1016/j.nucengdes.2025.114648
Fabrizio Aguzzi , Martín Armoa , Santiago M. Rabazzi , César Pairetti , Alejandro E. Albanesi
This work presents a modeling framework to represent the thermomechanical behavior of complex materials, based on micro mechanical dynamics. This tool is then applied to fuel rod elements, consisting of Zircaloy-2 cladding tubes and spacer grids, under typical Pressurized Water Reactor (PWR) conditions. The model incorporates thermal expansion and thermal creep through a VPSC–FEM coupling with the finite element method (FEM) solver Code_Aster, enabling analysis of in-reactor behavior under combined thermal, mechanical, and irradiation loading. The framework captures anisotropic deformation driven by crystallographic texture and prismatic slip activity under radial loading. Thermal creep, being stress-sensitive, contributes to early-stage stress relaxation and strain accumulation, leading to higher strain compared to the irradiation-only case. The interaction of thermal creep with irradiation mechanisms was found to modify the stress distribution and clearance (CLR) evolution, with relaxation governed by prismatic slip. For fuel rod elements, irradiation-induced mechanisms dominate the long-term CLR behavior, whereas thermal effects remain relevant in contact dynamics during thermal preloading. Furthermore, the stress–strain response was found to be more sensitive to micromechanics than to elasticity. This high-resolution formulation enables predictive modeling of spacer–cladding interaction and provides a basis for the development of reduced-order models.
{"title":"Multiphysics finite element modeling of irradiation and thermal behavior demonstrated on a fuel-assembly problem","authors":"Fabrizio Aguzzi , Martín Armoa , Santiago M. Rabazzi , César Pairetti , Alejandro E. Albanesi","doi":"10.1016/j.nucengdes.2025.114648","DOIUrl":"10.1016/j.nucengdes.2025.114648","url":null,"abstract":"<div><div>This work presents a modeling framework to represent the thermomechanical behavior of complex materials, based on micro mechanical dynamics. This tool is then applied to fuel rod elements, consisting of Zircaloy-2 cladding tubes and spacer grids, under typical Pressurized Water Reactor (PWR) conditions. The model incorporates thermal expansion and thermal creep through a VPSC–FEM coupling with the finite element method (FEM) solver Code_Aster, enabling analysis of in-reactor behavior under combined thermal, mechanical, and irradiation loading. The framework captures anisotropic deformation driven by crystallographic texture and prismatic slip activity under radial loading. Thermal creep, being stress-sensitive, contributes to early-stage stress relaxation and strain accumulation, leading to higher strain compared to the irradiation-only case. The interaction of thermal creep with irradiation mechanisms was found to modify the stress distribution and clearance (CLR) evolution, with relaxation governed by prismatic slip. For fuel rod elements, irradiation-induced mechanisms dominate the long-term CLR behavior, whereas thermal effects remain relevant in contact dynamics during thermal preloading. Furthermore, the stress–strain response was found to be more sensitive to micromechanics than to elasticity. This high-resolution formulation enables predictive modeling of spacer–cladding interaction and provides a basis for the development of reduced-order models.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"447 ","pages":"Article 114648"},"PeriodicalIF":2.1,"publicationDate":"2025-12-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145683749","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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