Pub Date : 2026-04-01Epub Date: 2026-01-24DOI: 10.1016/j.nucengdes.2026.114781
Zelong Zhao , Honghang Chi , Yuchen Xie , Yahui Wang , Yu Ma
Ultra-real-time simulation is crucial for ensuring the safe operation and control of nuclear power plants, as it enables rapid prediction and response to thermal-hydraulic behavior under accident conditions. This study proposes an ultra-real-time thermal-hydraulic modeling approach for the reactor primary circuit based on an intrusive reduced-order model (ROM). The governing equations of all components are discretized using a finite difference scheme, and variables for establishing ROMs are selected from these discretized equations to eliminate nonlinear terms. The transient solutions obtained from the initial 5% of time steps, calculated by the full-order model, served as snapshots, from which characteristic modes are extracted using the proper orthogonal decomposition. By projecting the discretized governing equations of each component onto the characteristic mode space, ultra-real-time thermal-hydraulic ROMs are constructed for each component. The integration of these ROMs for all components resulted in a comprehensive ultra-real-time model (URTM) of the primary circuit, capable of predicting system evolution. Simulation results demonstrated that the URTM achieves ultra-real-time performance while maintaining a maximum relative error of less than 0.15% for key thermal-hydraulic parameters.
{"title":"Ultra-real-time model reduction for nuclear reactor primary circuit calculation","authors":"Zelong Zhao , Honghang Chi , Yuchen Xie , Yahui Wang , Yu Ma","doi":"10.1016/j.nucengdes.2026.114781","DOIUrl":"10.1016/j.nucengdes.2026.114781","url":null,"abstract":"<div><div>Ultra-real-time simulation is crucial for ensuring the safe operation and control of nuclear power plants, as it enables rapid prediction and response to thermal-hydraulic behavior under accident conditions. This study proposes an ultra-real-time thermal-hydraulic modeling approach for the reactor primary circuit based on an intrusive reduced-order model (ROM). The governing equations of all components are discretized using a finite difference scheme, and variables for establishing ROMs are selected from these discretized equations to eliminate nonlinear terms. The transient solutions obtained from the initial 5% of time steps, calculated by the full-order model, served as snapshots, from which characteristic modes are extracted using the proper orthogonal decomposition. By projecting the discretized governing equations of each component onto the characteristic mode space, ultra-real-time thermal-hydraulic ROMs are constructed for each component. The integration of these ROMs for all components resulted in a comprehensive ultra-real-time model (URTM) of the primary circuit, capable of predicting system evolution. Simulation results demonstrated that the URTM achieves ultra-real-time performance while maintaining a maximum relative error of less than 0.15% for key thermal-hydraulic parameters.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114781"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036028","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: 2026-02-02DOI: 10.1016/j.nucengdes.2026.114793
Makuteswara Srinivasan
{"title":"Erratum to “An analysis of the brittleness indices of SiC layer in the TRISO fuel,” Nucl. Eng. Design 446 (2026) 114614.","authors":"Makuteswara Srinivasan","doi":"10.1016/j.nucengdes.2026.114793","DOIUrl":"10.1016/j.nucengdes.2026.114793","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114793"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189363","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 corrosion resistance and structural mechanical properties of neutron shielding materials such as B4C/Al and boron‑aluminum alloy are insufficient. The 316L stainless steel commonly used in the nuclear industry has excellent structural mechanical properties and corrosion resistance, but its neutron shielding performance is poor. The above-mentioned materials are unable to meet the multi-dimensional functional requirements of the core shielding layer of fast reactors, the structural components within the reactor, and the storage of spent fuel, which include high load-bearing capacity, efficient neutron shielding, and excellent corrosion resistance. Therefore, 316L-based neutron shielding materials are designed and prepared using 316L stainless steel as the substrate and B4C, Gd, Sm2O3, and Eu2O3 as neutron shielding enhancers. Firstly, the relationships between the areal densities of B4C, Gd, Sm2O3, and Eu2O3, and the neutron shielding rate and secondary γ-ray production rate of the corresponding shielding materials are established. Secondly, using the established relational equation, the required contents of B4C, Gd, Sm2O3, and Eu2O3 in 316L stainless steel are calculated and compared using the neutron shielding rate of 30% B4C/Al composite as the design basis. Finally, Gd/316L neutron shielding materials are prepared by directed energy deposition additive manufacturing (DED-AM) process, and their micro-morphology and mechanical properties are analyzed. The results show that Gd is more suitable as a neutron shielding enhancer for 316L stainless steel. The 1.9% Gd/316L exhibits good mechanical properties, while a further increase in Gd content degrades the mechanical performance of the material
{"title":"Design of 316L-based neutron shielding materials and preparation and characterization of Gd/316L","authors":"Chunguang Qiao , Zhonghua Wang , Xinpeng Wei , Dehui Wu , Chao Jiang","doi":"10.1016/j.nucengdes.2026.114808","DOIUrl":"10.1016/j.nucengdes.2026.114808","url":null,"abstract":"<div><div>The corrosion resistance and structural mechanical properties of neutron shielding materials such as B<sub>4</sub>C/Al and boron‑aluminum alloy are insufficient. The 316L stainless steel commonly used in the nuclear industry has excellent structural mechanical properties and corrosion resistance, but its neutron shielding performance is poor. The above-mentioned materials are unable to meet the multi-dimensional functional requirements of the core shielding layer of fast reactors, the structural components within the reactor, and the storage of spent fuel, which include high load-bearing capacity, efficient neutron shielding, and excellent corrosion resistance. Therefore, 316L-based neutron shielding materials are designed and prepared using 316L stainless steel as the substrate and B<sub>4</sub>C, Gd, Sm<sub>2</sub>O<sub>3</sub>, and Eu<sub>2</sub>O<sub>3</sub> as neutron shielding enhancers. Firstly, the relationships between the areal densities of B<sub>4</sub>C, Gd, Sm<sub>2</sub>O<sub>3</sub>, and Eu<sub>2</sub>O<sub>3</sub>, and the neutron shielding rate and secondary γ-ray production rate of the corresponding shielding materials are established. Secondly, using the established relational equation, the required contents of B<sub>4</sub>C, Gd, Sm<sub>2</sub>O<sub>3</sub>, and Eu<sub>2</sub>O<sub>3</sub> in 316L stainless steel are calculated and compared using the neutron shielding rate of 30% B<sub>4</sub>C/Al composite as the design basis. Finally, Gd/316L neutron shielding materials are prepared by directed energy deposition additive manufacturing (DED-AM) process, and their micro-morphology and mechanical properties are analyzed. The results show that Gd is more suitable as a neutron shielding enhancer for 316L stainless steel. The 1.9% Gd/316L exhibits good mechanical properties, while a further increase in Gd content degrades the mechanical performance of the material</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114808"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190736","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: 2026-01-14DOI: 10.1016/j.nucengdes.2026.114755
Jeehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho
The purpose of this study is to develop a robustness assessment methodology for performance evaluation considering the performance characteristics of passive safety systems being introduced in light water reactors and to propose safety analysis guidelines for passive safety systems by evaluating their impact on various performance degradation factors. To develop the methodology, the concerns with the introduction of passive safety systems and the current technical standards for passive safety systems from regulatory bodies around the world were analyzed. Since passive safety systems have less existing operating experience, there is uncertainty about the performance of the system, and it is necessary to prove the applicability of existing system analysis codes. In addition, since a passive safety system does not use devices such as pumps, it is more likely than a conventional safety system that the performance of the system will be degraded by changes in the internal or external environment. Therefore, this study developed a robustness assessment methodology consisting of seven steps to evaluate the impact of issues on the introduction of a passive safety system and to demonstrate the ability of the passive system to perform safety functions.
{"title":"Development of robustness assessment methodology for passive safety system against potential performance issue","authors":"Jeehee Lee , Seong-Su Jeon , Ju-Yeop Park , Hyoung Kyu Cho","doi":"10.1016/j.nucengdes.2026.114755","DOIUrl":"10.1016/j.nucengdes.2026.114755","url":null,"abstract":"<div><div>The purpose of this study is to develop a robustness assessment methodology for performance evaluation considering the performance characteristics of passive safety systems being introduced in light water reactors and to propose safety analysis guidelines for passive safety systems by evaluating their impact on various performance degradation factors. To develop the methodology, the concerns with the introduction of passive safety systems and the current technical standards for passive safety systems from regulatory bodies around the world were analyzed. Since passive safety systems have less existing operating experience, there is uncertainty about the performance of the system, and it is necessary to prove the applicability of existing system analysis codes. In addition, since a passive safety system does not use devices such as pumps, it is more likely than a conventional safety system that the performance of the system will be degraded by changes in the internal or external environment. Therefore, this study developed a robustness assessment methodology consisting of seven steps to evaluate the impact of issues on the introduction of a passive safety system and to demonstrate the ability of the passive system to perform safety functions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114755"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981790","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}
A novel high-efficiency diaphragm hydrodiode (i.e., fluidic diode) for NPP safety circuits is proposed. To achieve maximum diodicity, a multi-parameter optimization of its geometry is performed using a machine-learning-aided surrogate model. Training the surrogate model is performed using the quasi-random sampling, while the exact diodicity values were provided by the CFD simulations based on the Reynolds-Averaged Navier-Stokes equations closed with the turbulence model. Iterative complementation of the sampling is employed to further increase the surrogate model accuracy. Genetic and Trust-Region optimization algorithms are executed on top of the surrogate model to arrive at the optimal hydrodiode configuration. The maximum diodicity value reported by both CFD and the surrogate model is , while the experimentally confirmed diodicity of the optimal diode configuration is found to be . Such a high diodicity value for the diaphragm hydrodiode is reported for the first time, thus constituting an achievement in the field. The proposed design and optimization methodology open up possibilities for constructing compact and reliable passive components for safety systems.
提出了一种用于核电站安全电路的新型高效膜片水二极管（即流体二极管）。为了实现最大的二度性，使用机器学习辅助代理模型对其几何形状进行多参数优化。采用准随机抽样方法对代理模型进行训练，而基于k−ω湍流模型封闭的reynolds - average Navier-Stokes方程的CFD模拟提供了精确的二度值。采用采样的迭代互补，进一步提高代理模型的精度。在代理模型的基础上执行遗传算法和信任域优化算法，得到最优的水二极管结构。CFD和替代模型所报告的最大二度值均为DCFD≈2.74，而实验证实的最佳二极管配置的二度值为Dexp=2.59。本文首次报道了膜片式水二极管如此高的双极性值，这是该领域的一项成就。提出的设计和优化方法为构建紧凑可靠的安全系统被动元件提供了可能性。
{"title":"Multidimensional optimization of the high-diodicity diaphragm hydrodiode for passive safety systems of nuclear power plants","authors":"Victor Shcherba , Anatoliy Khait , Sergey Kaigorodov , Ksenia Sokirko , Evgeniy Pavlyuchenko","doi":"10.1016/j.nucengdes.2026.114803","DOIUrl":"10.1016/j.nucengdes.2026.114803","url":null,"abstract":"<div><div>A novel high-efficiency diaphragm hydrodiode (i.e., fluidic diode) for NPP safety circuits is proposed. To achieve maximum diodicity, a multi-parameter optimization of its geometry is performed using a machine-learning-aided surrogate model. Training the surrogate model is performed using the quasi-random sampling, while the exact diodicity values were provided by the CFD simulations based on the Reynolds-Averaged Navier-Stokes equations closed with the <span><math><mi>k</mi><mo>−</mo><mi>ω</mi></math></span> turbulence model. Iterative complementation of the sampling is employed to further increase the surrogate model accuracy. Genetic and Trust-Region optimization algorithms are executed on top of the surrogate model to arrive at the optimal hydrodiode configuration. The maximum diodicity value reported by both CFD and the surrogate model is <span><math><msub><mi>D</mi><mi>CFD</mi></msub><mo>≈</mo><mn>2.74</mn></math></span>, while the experimentally confirmed diodicity of the optimal diode configuration is found to be <span><math><msub><mi>D</mi><mi>exp</mi></msub><mo>=</mo><mn>2.59</mn></math></span>. Such a high diodicity value for the diaphragm hydrodiode is reported for the first time, thus constituting an achievement in the field. The proposed design and optimization methodology open up possibilities for constructing compact and reliable passive components for safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114803"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081253","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: 2026-02-02DOI: 10.1016/j.nucengdes.2026.114811
Junqiang Zheng , Jingyu Nie , Ya'nan He , Yingwei Wu , Jing Zhang , Yuanming Li , Shichao Liu , G.H. Su
This study presents a comprehensive multi-physics coupled analysis of TRISO-based Ceramic matrix Dispersion Microencapsulated (CDM) fuels, focusing on the debonding behavior at the TRISO-matrix interface under irradiation and thermal conditions, as well as the influence of TRISO-matrix interactions on TRISO particle performance within fuel pellets. A two-dimensional axisymmetric representative unit cell model was developed to investigate the effects of interfacial debonding characteristics under varying bonding parameters. Based on these findings, a three-dimensional pellet-scale multi-physics coupled model incorporating the Cohesive Zone Model was established to simulate interfacial debonding between TRISO particles and the matrix and its impact on structural integrity and failure probability. The model integrates irradiation-induced deformation, thermomechanical behavior, fission gas release, and interfacial damage evolution, enabling detailed evaluation of stress distribution and failure probability in both TRISO particles and SiC matrices.
Key results reveal that early-stage debonding at the TRISO-matrix interface is primarily driven by thermal expansion mismatch, with irradiation-induced deformation of the PyC layer contributing minimally. Smaller initial damage displacements accelerate interface failure, while larger failure displacements delay debonding but increase coating layer stresses, particularly in the SiC layer. At low temperatures, internal gas pressure within TRISO particles remains insufficient to induce critical tensile stress in the SiC layer, whereas PyC layer irradiation-induced deformation dominates the failure probability of the SiC coating. Additionally, spatial particle distribution and the surrounding matrix state significantly influence SiC coating failure, with localized high tensile stresses in narrow matrix regions posing risks for microcrack initiation.
Comparative analysis between pellet-scale CDM models and single TRISO particle models highlights the amplification of PyC layer hoop stresses due to particle-matrix interactions. The maximum failure probability of SiC coatings in pellet-scale models is approximately one order of magnitude higher than in single-particle simulations, underscoring the critical role of matrix confinement and interparticle interactions. These findings emphasize the necessity of considering high-packing-fraction configurations and interfacial bonding parameters in TRISO-based fuel performance assessments. Future work will focus on quantifying the effects of particle-matrix interactions on TRISO particle performance.
{"title":"Initial study on the effects of TRISO-matrix interactions on TRISO particle performance in ceramic matrix dispersion microencapsulated fuel pellet","authors":"Junqiang Zheng , Jingyu Nie , Ya'nan He , Yingwei Wu , Jing Zhang , Yuanming Li , Shichao Liu , G.H. Su","doi":"10.1016/j.nucengdes.2026.114811","DOIUrl":"10.1016/j.nucengdes.2026.114811","url":null,"abstract":"<div><div>This study presents a comprehensive multi-physics coupled analysis of TRISO-based Ceramic matrix Dispersion Microencapsulated (CDM) fuels, focusing on the debonding behavior at the TRISO-matrix interface under irradiation and thermal conditions, as well as the influence of TRISO-matrix interactions on TRISO particle performance within fuel pellets. A two-dimensional axisymmetric representative unit cell model was developed to investigate the effects of interfacial debonding characteristics under varying bonding parameters. Based on these findings, a three-dimensional pellet-scale multi-physics coupled model incorporating the Cohesive Zone Model was established to simulate interfacial debonding between TRISO particles and the matrix and its impact on structural integrity and failure probability. The model integrates irradiation-induced deformation, thermomechanical behavior, fission gas release, and interfacial damage evolution, enabling detailed evaluation of stress distribution and failure probability in both TRISO particles and SiC matrices.</div><div>Key results reveal that early-stage debonding at the TRISO-matrix interface is primarily driven by thermal expansion mismatch, with irradiation-induced deformation of the PyC layer contributing minimally. Smaller initial damage displacements accelerate interface failure, while larger failure displacements delay debonding but increase coating layer stresses, particularly in the SiC layer. At low temperatures, internal gas pressure within TRISO particles remains insufficient to induce critical tensile stress in the SiC layer, whereas PyC layer irradiation-induced deformation dominates the failure probability of the SiC coating. Additionally, spatial particle distribution and the surrounding matrix state significantly influence SiC coating failure, with localized high tensile stresses in narrow matrix regions posing risks for microcrack initiation.</div><div>Comparative analysis between pellet-scale CDM models and single TRISO particle models highlights the amplification of PyC layer hoop stresses due to particle-matrix interactions. The maximum failure probability of SiC coatings in pellet-scale models is approximately one order of magnitude higher than in single-particle simulations, underscoring the critical role of matrix confinement and interparticle interactions. These findings emphasize the necessity of considering high-packing-fraction configurations and interfacial bonding parameters in TRISO-based fuel performance assessments. Future work will focus on quantifying the effects of particle-matrix interactions on TRISO particle performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114811"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190259","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.2026.114813
Ahmed Jasim , Mauricio Maestri , Abdullah Al Zubaidi , Omar Farid , Muthanna Al-Dahhan
Accurate characterization of void-fraction distributions in pebble-bed reactors (PBRs) is essential for predicting flow, heat transfer, and neutronic behavior. High-fidelity experimental benchmark data for validating such predictions remain scarce, largely due to the challenges of non-invasive measurements. In this study, gamma-ray computed tomography (CT) was employed to measure radial and cross-sectional porosity in a laboratory-scale pebble bed containing graphite pebbles. A Discrete Element Method (DEM) simulation was implemented and validated against these measurements, then applied to the full-scale geometry of the Xe-100 high-temperature gas-cooled pebble-bed reactor. Analyses included radial and axial void-fraction profiles in the cylindrical section and conical base, with particular attention to near-wall oscillations at multiple axial levels. Both axially averaged profiles, integrating over extended bed sections, and locally resolved profiles, capturing fine-scale oscillations, were evaluated. Additional analyses examined cross-sectional void distributions and the effect of pebble recirculation. The DEM results reproduced the expected near-wall oscillatory layering with a characteristic wavelength of ∼1 and bulk void fractions near and further showed that oscillatory patterns persist into the conical region, where the first trough shifts outward, and a broader near-wall gap develops. Recirculation studies, corresponding to and complete bed inventory cycles, showed that structural rearrangements occur mainly during the initial passes, after which the bed attains a quasi-steady configuration. Recirculation intensified near-wall oscillations, particularly in the lower regions, but had negligible impact on bulk porosity in the cylindrical section. In the cone, however, the void fraction was elevated during dynamic operation due to pebble drainage and upward void propagation. The findings support improved neutronic and thermal-hydraulic modeling and contribute to the design and safety assessment of next-generation pebble-bed systems.
{"title":"Near-wall void distribution characterization in pebble bed reactor using gamma-ray CT and DEM simulation","authors":"Ahmed Jasim , Mauricio Maestri , Abdullah Al Zubaidi , Omar Farid , Muthanna Al-Dahhan","doi":"10.1016/j.nucengdes.2026.114813","DOIUrl":"10.1016/j.nucengdes.2026.114813","url":null,"abstract":"<div><div>Accurate characterization of void-fraction distributions in pebble-bed reactors (PBRs) is essential for predicting flow, heat transfer, and neutronic behavior. High-fidelity experimental benchmark data for validating such predictions remain scarce, largely due to the challenges of non-invasive measurements. In this study, gamma-ray computed tomography (CT) was employed to measure radial and cross-sectional porosity in a laboratory-scale pebble bed containing <span><math><mrow><mn>6</mn><mspace></mspace><mi>cm</mi></mrow></math></span> graphite pebbles. A Discrete Element Method (DEM) simulation was implemented and validated against these measurements, then applied to the full-scale geometry of the Xe-100 high-temperature gas-cooled pebble-bed reactor. Analyses included radial and axial void-fraction profiles in the cylindrical section and conical base, with particular attention to near-wall oscillations at multiple axial levels. Both axially averaged profiles, integrating over extended bed sections, and locally resolved profiles, capturing fine-scale oscillations, were evaluated. Additional analyses examined cross-sectional void distributions and the effect of pebble recirculation. The DEM results reproduced the expected near-wall oscillatory layering with a characteristic wavelength of ∼1 <span><math><mrow><mi>dp</mi></mrow></math></span> and bulk void fractions near <span><math><mrow><mn>0.40</mn></mrow></math></span> and further showed that oscillatory patterns persist into the conical region, where the first trough shifts outward, and a broader near-wall gap develops. Recirculation studies, corresponding to <span><math><mrow><mn>5</mn><mo>,</mo><mn>10</mn><mo>,</mo></mrow></math></span> and <span><math><mrow><mn>15</mn></mrow></math></span> complete bed inventory cycles, showed that structural rearrangements occur mainly during the initial passes, after which the bed attains a quasi-steady configuration. Recirculation intensified near-wall oscillations, particularly in the lower regions, but had negligible impact on bulk porosity in the cylindrical section. In the cone, however, the void fraction was elevated during dynamic operation due to pebble drainage and upward void propagation. The findings support improved neutronic and thermal-hydraulic modeling and contribute to the design and safety assessment of next-generation pebble-bed systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114813"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190258","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: 2026-02-02DOI: 10.1016/j.nucengdes.2026.114797
Alif Imran Mohd Shuhaimi, Mohd Syukri Yahya, Shamsul Amri Sulaiman, Eli Syafiqah Aziman
Achieving Net Zero Emissions (NZE) by 2050 necessitates a strategic combination of renewable energy sources and dependable, clean baseload power. Nuclear energy is increasingly recognized as a critical element of this transition, with recent global policy endorsements calling for a tripling of nuclear capacity, underscoring its growing importance. Substantial expansion is anticipated not only in established nuclear markets but also in newcomer countries. However, initiating nuclear power programs poses significant challenges for newcomers, particularly due to massive capital investment, long-term infrastructure commitments, and the complexity of managing associated development risks. Reactor Technology Assessment (RTA) is therefore an essential process to ensure nuclear projects are delivered on time, within budget, safely, and to specifications. The International Atomic Energy Agency (IAEA) supports this process through its comprehensive framework of 19 infrastructure issues, addressing key dimensions such as national position, technical, and economic factors. This study presents a systematic, multi-criteria decision-making framework for conducting RTA, focusing on Small Modular Reactors (SMRs) as viable options for deployment in newcomer countries. The framework integrates the IAEA's infrastructure guidance with a PESTEL-based evaluation model. A case study is conducted for Malaysia to identify suitable SMRs for a Nuclear Hydrogen Hybrid System (NHHS). The assessment framework is validated using the Analytic Hierarchy Process (AHP), ensuring robustness and consistency in decision-making. The findings provide insights into aligning SMR technology selections with national requirements and risk profiles, offering a replicable methodology to support evidence-based decision-making in emerging nuclear energy programs.
{"title":"Analytic hierarchy process-based integrated assessment of IAEA'S 19 infrastructure criteria for small modular reactor technology assessment","authors":"Alif Imran Mohd Shuhaimi, Mohd Syukri Yahya, Shamsul Amri Sulaiman, Eli Syafiqah Aziman","doi":"10.1016/j.nucengdes.2026.114797","DOIUrl":"10.1016/j.nucengdes.2026.114797","url":null,"abstract":"<div><div>Achieving Net Zero Emissions (NZE) by 2050 necessitates a strategic combination of renewable energy sources and dependable, clean baseload power. Nuclear energy is increasingly recognized as a critical element of this transition, with recent global policy endorsements calling for a tripling of nuclear capacity, underscoring its growing importance. Substantial expansion is anticipated not only in established nuclear markets but also in newcomer countries. However, initiating nuclear power programs poses significant challenges for newcomers, particularly due to massive capital investment, long-term infrastructure commitments, and the complexity of managing associated development risks. Reactor Technology Assessment (RTA) is therefore an essential process to ensure nuclear projects are delivered on time, within budget, safely, and to specifications. The International Atomic Energy Agency (IAEA) supports this process through its comprehensive framework of 19 infrastructure issues, addressing key dimensions such as national position, technical, and economic factors. This study presents a systematic, multi-criteria decision-making framework for conducting RTA, focusing on Small Modular Reactors (SMRs) as viable options for deployment in newcomer countries. The framework integrates the IAEA's infrastructure guidance with a PESTEL-based evaluation model. A case study is conducted for Malaysia to identify suitable SMRs for a Nuclear Hydrogen Hybrid System (NHHS). The assessment framework is validated using the Analytic Hierarchy Process (AHP), ensuring robustness and consistency in decision-making. The findings provide insights into aligning SMR technology selections with national requirements and risk profiles, offering a replicable methodology to support evidence-based decision-making in emerging nuclear energy programs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114797"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190256","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: 2026-02-11DOI: 10.1016/j.nucengdes.2026.114809
S. Ozerov , N. Boyle , D. DiPrete , R.P. Taleyarkhan
This paper addresses a need for spectroscopic monitoring for the illicit movement of stationary/moving special nuclear materials (SNMs) via high-efficiency real-time detection from tell-tale neutron radiation signatures, especially while in motion, and with 100% background gamma-beta blindness. Studies were conducted in a direct one-on-one comparison vs current state-of-the-art technology, which conventionally employs He-3 detectors, which require bulky shielding for neutron down-scattering. The advanced Economical Acoustically Tensioned Metastable Fluid Detector [E-ATMFD(Ver.1)] developed at Purdue University promises to offer 10× improved efficiency of SNM detection of equivalent form factor He-3 detectors. This paper details the design and performance of the advanced E-ATMFD(Ver.1) sensor system using borated sensing fluid, which demonstrated absolute efficiencies x10–20 greater than a state-of-the-art moderated Ludlum-42-49B (He-3) detector of equivalent size, when compared side-by-side for detecting and tracking a shielded and unshielded moving Plutonium-Beryllium (SNM surrogate of similar intensity) neutron source. An additional factor of 2× improvement was obtained by tailoring the internal and external static pressure fields, leading to enhanced dynamic tensioned (negative) pressure (Pneg) metastability states within the E-ATMFD(Ver.1). By sweeping the Pneg states, the E-ATMFD (Ver.1) was also shown capable of spectroscopically distinguishing a Cf-252 fission multiplicity-bearing neutron source from a non-multiplicity AmBe (α,n) random neutron source.
{"title":"Advanced acoustically tensioned metastable fluid detectors for 10-15× over He-3 detector efficiency, real-time monitoring of moving SNMS, and spectroscopic identification of fission vs. alpha-n neutron sources","authors":"S. Ozerov , N. Boyle , D. DiPrete , R.P. Taleyarkhan","doi":"10.1016/j.nucengdes.2026.114809","DOIUrl":"10.1016/j.nucengdes.2026.114809","url":null,"abstract":"<div><div>This paper addresses a need for spectroscopic monitoring for the illicit movement of stationary/moving special nuclear materials (SNMs) via high-efficiency real-time detection from tell-tale neutron radiation signatures, especially while in motion, and with 100% background gamma-beta blindness. Studies were conducted in a direct one-on-one comparison vs current state-of-the-art technology, which conventionally employs He-3 detectors, which require bulky shielding for neutron down-scattering. The advanced Economical Acoustically Tensioned Metastable Fluid Detector [<em>E</em>-ATMFD(Ver.1)] developed at Purdue University promises to offer 10× improved efficiency of SNM detection of equivalent form factor He-3 detectors. This paper details the design and performance of the advanced <em>E</em>-ATMFD(Ver.1) sensor system using borated sensing fluid, which demonstrated absolute efficiencies x10–20 greater than a state-of-the-art moderated Ludlum-42-49B (He-3) detector of equivalent size, when compared side-by-side for detecting and tracking a shielded and unshielded moving Plutonium-Beryllium (SNM surrogate of similar intensity) neutron source. An additional factor of 2× improvement was obtained by tailoring the internal and external static pressure fields, leading to enhanced dynamic tensioned (negative) pressure (Pneg) metastability states within the <em>E</em>-ATMFD(Ver.1). By sweeping the Pneg states, the E-ATMFD (Ver.1) was also shown capable of spectroscopically distinguishing a Cf-252 fission multiplicity-bearing neutron source from a non-multiplicity Am<img>Be (α,n) random neutron source.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114809"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190252","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: 2026-01-31DOI: 10.1016/j.nucengdes.2026.114805
Ismail Erol , Ahmet Oztel , Ihsan Tolga Medeni , Ilker Murat Ar
Small modular reactors (SMRs) promise reduced upfront costs, faster construction, and enhanced safety compared to traditional reactors. However, widespread adoption is hindered by challenges such as high capital costs, regulatory delays, supply chain inefficiencies, cybersecurity risks, nuclear waste management, and public skepticism. Despite qualitative studies highlighting these barriers, quantitative analyses remain scarce, necessitating systematic frameworks to model interdependencies and guide solutions. The goal of this study is to scrutinize SMR adoption challenges using a novel multi-criteria decision-making (MCDM) approach. Drawing on a literature review from Web of Science, Scopus, and various reports from international institutions, 13 key challenges were identified. A panel of 58 experts—academics, government officials, and cybersecurity specialists—provided inputs via pairwise comparisons. The methodology used in this study integrates Fermatean Fuzzy Interpretive Structural Modeling (FFISM) with the cross-impact matrix multiplication applied to classification (MICMAC) analysis. Fermatean fuzzy sets extend traditional ISM by accommodating higher uncertainty in expert judgments through expanded membership/non-membership degrees. Validation involved 10,000 simulations comparing FFISM to conventional fuzzy ISM. Results reveal a six-level hierarchy: licensing/regulatory constraints and lack of proven technology/FOAK units as top challenges, influencing linkage challenges such as supply chain effectiveness, cybersecurity risks, and waste management. Dependent challenges include perceived investment risk, cost estimation, and public opinion. Policy recommendations include risk-informed licensing to cut timelines, blockchain for traceability addressing fuel availability, and public-private partnerships with green bonds to mitigate risks. This research provides actionable strategies for policymakers and stakeholders to accelerate SMR deployment, strengthening nuclear energy's role in global decarbonization.
与传统反应堆相比，小型模块化反应堆（smr）有望降低前期成本，加快建设速度，提高安全性。然而，高资本成本、监管延误、供应链效率低下、网络安全风险、核废料管理和公众怀疑等挑战阻碍了这种技术的广泛采用。尽管定性研究强调了这些障碍，但定量分析仍然很少，需要系统框架来模拟相互依赖关系并指导解决方案。本研究的目的是使用一种新颖的多标准决策（MCDM）方法来审视SMR采用的挑战。通过对Web of Science、Scopus的文献综述以及国际机构的各种报告，确定了13个关键挑战。一个由58名专家组成的小组——学者、政府官员和网络安全专家——通过两两比较提供了意见。本研究采用Fermatean Fuzzy Interpretive Structural Modeling （FFISM）与cross-impact matrix multiplication应用于分类分析（MICMAC）相结合的方法。Fermatean模糊集通过扩展隶属度/非隶属度来适应专家判断的高不确定性，从而扩展了传统的模糊集。验证涉及10,000个模拟，将FFISM与传统模糊ISM进行比较。结果显示了六个层次结构：许可/监管约束和缺乏成熟的技术/FOAK单元是最大的挑战，影响供应链有效性、网络安全风险和废物管理等联系挑战。相关挑战包括可感知的投资风险、成本估算和公众意见。政策建议包括风险知情许可以缩短时间表，区块链可追溯性解决燃料可用性，以及与绿色债券建立公私合作伙伴关系以降低风险。本研究为政策制定者和利益相关者提供了可操作的战略，以加速小型反应堆的部署，加强核能在全球脱碳中的作用。
{"title":"Enhancing small modular reactor adoption for sustainable energy transition: a Fermatean fuzzy ISM-MICMAC framework for analyzing challenges","authors":"Ismail Erol , Ahmet Oztel , Ihsan Tolga Medeni , Ilker Murat Ar","doi":"10.1016/j.nucengdes.2026.114805","DOIUrl":"10.1016/j.nucengdes.2026.114805","url":null,"abstract":"<div><div>Small modular reactors (SMRs) promise reduced upfront costs, faster construction, and enhanced safety compared to traditional reactors. However, widespread adoption is hindered by challenges such as high capital costs, regulatory delays, supply chain inefficiencies, cybersecurity risks, nuclear waste management, and public skepticism. Despite qualitative studies highlighting these barriers, quantitative analyses remain scarce, necessitating systematic frameworks to model interdependencies and guide solutions. The goal of this study is to scrutinize SMR adoption challenges using a novel multi-criteria decision-making (MCDM) approach. Drawing on a literature review from Web of Science, Scopus, and various reports from international institutions, 13 key challenges were identified. A panel of 58 experts—academics, government officials, and cybersecurity specialists—provided inputs via pairwise comparisons. The methodology used in this study integrates Fermatean Fuzzy Interpretive Structural Modeling (FFISM) with the cross-impact matrix multiplication applied to classification (MICMAC) analysis. Fermatean fuzzy sets extend traditional ISM by accommodating higher uncertainty in expert judgments through expanded membership/non-membership degrees. Validation involved 10,000 simulations comparing FFISM to conventional fuzzy ISM. Results reveal a six-level hierarchy: licensing/regulatory constraints and lack of proven technology/FOAK units as top challenges, influencing linkage challenges such as supply chain effectiveness, cybersecurity risks, and waste management. Dependent challenges include perceived investment risk, cost estimation, and public opinion. Policy recommendations include risk-informed licensing to cut timelines, blockchain for traceability addressing fuel availability, and public-private partnerships with green bonds to mitigate risks. This research provides actionable strategies for policymakers and stakeholders to accelerate SMR deployment, strengthening nuclear energy's role in global decarbonization.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114805"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190729","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号