Pub Date : 2024-11-16DOI: 10.1016/j.pnucene.2024.105519
V. Zeman, Z. Hlaváč, Š. Dyk
The paper presents a comprehensive method for modelling contact interactions between two different types of fuel assemblies within a mixed core. Particular emphasis is placed on the hexagonal fuel assemblies commonly found in VVER reactors. The presence of bowed fuel assemblies may induce mutual contact, resulting in temporary or permanent contact conditions. This interaction, in conjunction with the vibration of core components, can lead to fretting wear of the load-bearing skeletons of fuel assemblies. By using the proposed computational model and experimentally estimated fretting wear parameters, the method allows detailed analysis of fretting wear as a function of various factors. These factors include operating conditions, distribution of fuel assembly types within the mixed core, burn-up state of contacting assemblies and the shape of their static deformation.
{"title":"Contact phenomena between load-bearing skeletons of bowed fuel assemblies in PWRs with mixed cores","authors":"V. Zeman, Z. Hlaváč, Š. Dyk","doi":"10.1016/j.pnucene.2024.105519","DOIUrl":"10.1016/j.pnucene.2024.105519","url":null,"abstract":"<div><div>The paper presents a comprehensive method for modelling contact interactions between two different types of fuel assemblies within a mixed core. Particular emphasis is placed on the hexagonal fuel assemblies commonly found in VVER reactors. The presence of bowed fuel assemblies may induce mutual contact, resulting in temporary or permanent contact conditions. This interaction, in conjunction with the vibration of core components, can lead to fretting wear of the load-bearing skeletons of fuel assemblies. By using the proposed computational model and experimentally estimated fretting wear parameters, the method allows detailed analysis of fretting wear as a function of various factors. These factors include operating conditions, distribution of fuel assembly types within the mixed core, burn-up state of contacting assemblies and the shape of their static deformation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105519"},"PeriodicalIF":3.3,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653787","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 : 2024-11-14DOI: 10.1016/j.pnucene.2024.105518
Andrea M. Jokisaari , Stephen Taller , Yiren Chen , Wei-Ying Chen , Rongjie Song
As the needs for the nuclear energy industry continue to evolve in the century, timely adoption of new technological solutions acceptable to regulatory agencies is critical. Quantitative prediction of radiation damage in materials and its impact on mechanical properties is a key component of licensing and regulatory decisions regarding nuclear power plants. Accelerated testing methodologies such as combined ion and neutron irradiation data sets are crucial for the development and deployment of new materials and new manufacturing methods (e.g., additive manufacturing). However, regulatory acceptance of accelerated testing methodologies is necessary for their adoption. The present work discusses the fundamental basis for comparing ion- and neutron-induced material microstructures, the theory behind interpreting radiation damage across length and time scales and radiation types, and the codes, standards, and quality assurance concerns surrounding different modeling methods and software. In particular, recommendations are given as to the path forward that will enable national laboratories, academia, and industry to develop the modeling and software basis for regulatory acceptance of the combined use of ion and neutron irradiation for material performance evaluation.
{"title":"Promoting regulatory acceptance of combined ion and neutron irradiation testing of nuclear reactor materials: Modeling and software considerations","authors":"Andrea M. Jokisaari , Stephen Taller , Yiren Chen , Wei-Ying Chen , Rongjie Song","doi":"10.1016/j.pnucene.2024.105518","DOIUrl":"10.1016/j.pnucene.2024.105518","url":null,"abstract":"<div><div>As the needs for the nuclear energy industry continue to evolve in the <span><math><mrow><mn>21</mn><mi>st</mi></mrow></math></span> century, timely adoption of new technological solutions acceptable to regulatory agencies is critical. Quantitative prediction of radiation damage in materials and its impact on mechanical properties is a key component of licensing and regulatory decisions regarding nuclear power plants. Accelerated testing methodologies such as combined ion and neutron irradiation data sets are crucial for the development and deployment of new materials and new manufacturing methods (e.g., additive manufacturing). However, regulatory acceptance of accelerated testing methodologies is necessary for their adoption. The present work discusses the fundamental basis for comparing ion- and neutron-induced material microstructures, the theory behind interpreting radiation damage across length and time scales and radiation types, and the codes, standards, and quality assurance concerns surrounding different modeling methods and software. In particular, recommendations are given as to the path forward that will enable national laboratories, academia, and industry to develop the modeling and software basis for regulatory acceptance of the combined use of ion and neutron irradiation for material performance evaluation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105518"},"PeriodicalIF":3.3,"publicationDate":"2024-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653998","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 : 2024-11-13DOI: 10.1016/j.pnucene.2024.105526
Şeyma Bozkaya , Stephen Taiwo Onifade , Mahmut Sami Duran
As the global quest for clean energy grows, the environmentally friendly nature of nuclear energy as a potential non-fossil energy source is generating interest around the world. Therefore, we examine whether nuclear energy utilization has significantly driven carbon emission reduction among the utilizing states. Empirical analyses were conducted using second-generation techniques. The analyses conducted also incorporated testing the EKC theory, as well as examining the effects of natural resources and economic growth on emissions in the sample countries. The empirical analyses cover data from 2000 to 2020 for a total of 27 nuclear energy-using countries as obtained from the Statistical Review of World Energy (Bp, 2021). The findings show that neither the use of nuclear energy nor natural resources significantly reduces carbon emissions across the countries. Additionally, the EKC hypothesis of reduction in emission levels as income expands beyond a certain threshold does not hold for the countries. Moreover, the causality analysis shows that there is a one-way causality from emissions to nuclear energy use. These findings thus highlight the need for more research on how to minimize the indirect carbon footprint that is associated with nuclear energy utilization.
{"title":"Nuclear energy utilization and the expectations of emission-reduction gains: Empirical evidence from economic trajectory of selected utilizing states","authors":"Şeyma Bozkaya , Stephen Taiwo Onifade , Mahmut Sami Duran","doi":"10.1016/j.pnucene.2024.105526","DOIUrl":"10.1016/j.pnucene.2024.105526","url":null,"abstract":"<div><div>As the global quest for clean energy grows, the environmentally friendly nature of nuclear energy as a potential non-fossil energy source is generating interest around the world. Therefore, we examine whether nuclear energy utilization has significantly driven carbon emission reduction among the utilizing states. Empirical analyses were conducted using second-generation techniques. The analyses conducted also incorporated testing the EKC theory, as well as examining the effects of natural resources and economic growth on emissions in the sample countries. The empirical analyses cover data from 2000 to 2020 for a total of 27 nuclear energy-using countries as obtained from the Statistical Review of World Energy (Bp, 2021). The findings show that neither the use of nuclear energy nor natural resources significantly reduces carbon emissions across the countries. Additionally, the EKC hypothesis of reduction in emission levels as income expands beyond a certain threshold does not hold for the countries. Moreover, the causality analysis shows that there is a one-way causality from emissions to nuclear energy use. These findings thus highlight the need for more research on how to minimize the indirect carbon footprint that is associated with nuclear energy utilization.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105526"},"PeriodicalIF":3.3,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653788","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-13DOI: 10.1016/j.pnucene.2024.105530
Jing Wei , Guangchun Zhang
The accurate and efficient solution of the neutron transport equation is essential in nuclear reactor physics for understanding reactor kinetics, stability, and safety. This study investigates the application of the Krylov-Schur method to three-dimensional neutron transport criticality calculations within the Marvin code, comparing its performance to the traditional Power Iteration (PI) method. Using the Takeda benchmark problems, the Krylov-Schur solver demonstrated high accuracy in eigenvalue calculations and neutron flux distributions, closely matching reference Monte Carlo (MC) results. Additionally, the parallel efficiency of the Krylov-Schur method was evaluated, showing significant speed-up and better scalability compared to the PI method, particularly in large-scale computations. However, the method requires a larger memory footprint due to the need to store multiple Krylov subspace vectors and Schur decompositions. Despite this, the findings highlight the Krylov-Schur method's robustness and computational efficiency, making it a promising tool for neutron transport simulations in complex reactor configurations. Future work will focus on investigating the subtraction of high-order eigenvalues and eigenvectors using the Krylov-Schur method to further enhance neutron transport simulations.
在核反应堆物理学中，准确高效地求解中子输运方程对于理解反应堆动力学、稳定性和安全性至关重要。本研究调查了克雷洛夫-舒尔方法在 Marvin 代码三维中子输运临界计算中的应用，并将其性能与传统的功率迭代（PI）方法进行了比较。利用武田基准问题，克雷洛夫-舒尔求解器在特征值计算和中子通量分布方面表现出很高的精度，与参考蒙特卡罗（MC）结果非常接近。此外，还对 Krylov-Schur 方法的并行效率进行了评估，结果表明与 PI 方法相比，Krylov-Schur 方法的速度显著提高，可扩展性更好，尤其是在大规模计算中。不过，由于需要存储多个克雷洛夫子空间向量和舒尔分解，该方法需要占用较大的内存。尽管如此，研究结果还是凸显了克雷洛夫-舒尔方法的稳健性和计算效率，使其成为复杂反应堆配置中子输运模拟的理想工具。未来的工作将侧重于研究使用 Krylov-Schur 方法减去高阶特征值和特征向量，以进一步增强中子输运模拟。
{"title":"Application of the Krylov-Schur method in three-dimensional nuclear reactor discrete ordinates criticality calculations","authors":"Jing Wei , Guangchun Zhang","doi":"10.1016/j.pnucene.2024.105530","DOIUrl":"10.1016/j.pnucene.2024.105530","url":null,"abstract":"<div><div>The accurate and efficient solution of the neutron transport equation is essential in nuclear reactor physics for understanding reactor kinetics, stability, and safety. This study investigates the application of the Krylov-Schur method to three-dimensional neutron transport criticality calculations within the Marvin code, comparing its performance to the traditional Power Iteration (PI) method. Using the Takeda benchmark problems, the Krylov-Schur solver demonstrated high accuracy in eigenvalue calculations and neutron flux distributions, closely matching reference Monte Carlo (MC) results. Additionally, the parallel efficiency of the Krylov-Schur method was evaluated, showing significant speed-up and better scalability compared to the PI method, particularly in large-scale computations. However, the method requires a larger memory footprint due to the need to store multiple Krylov subspace vectors and Schur decompositions. Despite this, the findings highlight the Krylov-Schur method's robustness and computational efficiency, making it a promising tool for neutron transport simulations in complex reactor configurations. Future work will focus on investigating the subtraction of high-order eigenvalues and eigenvectors using the Krylov-Schur method to further enhance neutron transport simulations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105530"},"PeriodicalIF":3.3,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653773","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 : 2024-11-12DOI: 10.1016/j.pnucene.2024.105503
Mauricio E. Tano, Samuel A. Walker, Abdalla Abou-Jaoude, Robin Roper, Toni Karlsson, Mikael C.F. Karlsson, Parikshit Bajpai, Rodrigo de Oliveira, Ramiro Freile, Vasileios Kyriakopoulos, Mustafa K. Jaradat, Piyush Sabharwall
This study presents a computational methodology for analyzing isotopic evolution and associated uncertainties in molten salt reactors (MSRs), focusing on both fluoride- and chloride-based fuel salts. The primary goal is to enhance the understanding of isotopic behavior in MSRs and provide data to support future experimental efforts. The methodology integrates transport-coupled depletion calculations using OpenMC, equilibrium thermodynamics modeling with Thermochimica, and a corrosion model. Sensitivity analyses are performed to evaluate the impact of power density, air ingress, and humidity content on isotopic evolution in MSR concepts. This study examines representative F- and Cl-based MSR designs, highlighting the dominant influence of power density on isotopic composition, which significantly affects isotope production and depletion rates, accounting for approximately 76% of the observed variance in element concentration. Air ingress and humidity content also affect the redox potential, solubility of heavier elements, and corrosion rates, thereby altering the expected isotopic evolution in the reactor. On average, air ingress accounts for around 17% of the variance in element concentrations, while humidity explains the remaining 7%. These variances differ significantly from element to element, depending on the element’s role in depletion, redox potential evolution, and galvanic corrosion. The findings indicate that power density, air ingress, and humidity content are all critical factors for optimizing reactor design and operational strategies. Furthermore, the study provides expected ranges for key impurities in the fuel salt, which are crucial for guiding future experimental studies and refining MSR designs. Finally, this study demonstrates the importance of modeling depletion coupled with the evolution of redox potential and chemical interactions in MSR fuel salts.
{"title":"Coupled neutronics, thermochemistry, corrosion modeling and sensitivity analyses for isotopic evolution in molten salt reactors","authors":"Mauricio E. Tano, Samuel A. Walker, Abdalla Abou-Jaoude, Robin Roper, Toni Karlsson, Mikael C.F. Karlsson, Parikshit Bajpai, Rodrigo de Oliveira, Ramiro Freile, Vasileios Kyriakopoulos, Mustafa K. Jaradat, Piyush Sabharwall","doi":"10.1016/j.pnucene.2024.105503","DOIUrl":"10.1016/j.pnucene.2024.105503","url":null,"abstract":"<div><div>This study presents a computational methodology for analyzing isotopic evolution and associated uncertainties in molten salt reactors (MSRs), focusing on both fluoride- and chloride-based fuel salts. The primary goal is to enhance the understanding of isotopic behavior in MSRs and provide data to support future experimental efforts. The methodology integrates transport-coupled depletion calculations using OpenMC, equilibrium thermodynamics modeling with Thermochimica, and a corrosion model. Sensitivity analyses are performed to evaluate the impact of power density, air ingress, and humidity content on isotopic evolution in MSR concepts. This study examines representative F- and Cl-based MSR designs, highlighting the dominant influence of power density on isotopic composition, which significantly affects isotope production and depletion rates, accounting for approximately 76% of the observed variance in element concentration. Air ingress and humidity content also affect the redox potential, solubility of heavier elements, and corrosion rates, thereby altering the expected isotopic evolution in the reactor. On average, air ingress accounts for around 17% of the variance in element concentrations, while humidity explains the remaining 7%. These variances differ significantly from element to element, depending on the element’s role in depletion, redox potential evolution, and galvanic corrosion. The findings indicate that power density, air ingress, and humidity content are all critical factors for optimizing reactor design and operational strategies. Furthermore, the study provides expected ranges for key impurities in the fuel salt, which are crucial for guiding future experimental studies and refining MSR designs. Finally, this study demonstrates the importance of modeling depletion coupled with the evolution of redox potential and chemical interactions in MSR fuel salts.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105503"},"PeriodicalIF":3.3,"publicationDate":"2024-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653774","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-12DOI: 10.1016/j.pnucene.2024.105521
Shiqiao Liu , Zifei Zhu , Xinwen Zhao , Yangguang Wang , Xiang Sun , Lei Yu
The abnormal state detection in nuclear reactors constitutes a critical concern within the broader context of Nuclear Power Plants (NPPs) safety management. Deep learning techniques have exhibited exceptional performance in addressing issues pertaining to NPPs safety control. However, acquiring the large amount of labeled data required by supervised learning methodologies poses a significant challenge in practical applications. This paper addresses a key challenge in NPPs safety—abnormal state detection in nuclear reactors. Leveraging unsupervised learning due to the limited availability of labeled data, we propose an anomaly detection method using the Denoising Diffusion Probabilistic Model (DDPM) with a noise-to-noise training strategy. Comparative evaluation against AE, VAE, and GAN shows that DDPM outperforms in all metrics, demonstrating strong potential for NPPs anomaly diagnosis. Experimental results suggest that a feature count of 50 optimizes DDPM performance for NPPs anomaly detection, while the noise-to-noise training strategy improves model robustness.
{"title":"Unsupervised anomaly detection for Nuclear Power Plants based on Denoising Diffusion Probabilistic Models","authors":"Shiqiao Liu , Zifei Zhu , Xinwen Zhao , Yangguang Wang , Xiang Sun , Lei Yu","doi":"10.1016/j.pnucene.2024.105521","DOIUrl":"10.1016/j.pnucene.2024.105521","url":null,"abstract":"<div><div>The abnormal state detection in nuclear reactors constitutes a critical concern within the broader context of Nuclear Power Plants (NPPs) safety management. Deep learning techniques have exhibited exceptional performance in addressing issues pertaining to NPPs safety control. However, acquiring the large amount of labeled data required by supervised learning methodologies poses a significant challenge in practical applications. This paper addresses a key challenge in NPPs safety—abnormal state detection in nuclear reactors. Leveraging unsupervised learning due to the limited availability of labeled data, we propose an anomaly detection method using the Denoising Diffusion Probabilistic Model (DDPM) with a noise-to-noise training strategy. Comparative evaluation against AE, VAE, and GAN shows that DDPM outperforms in all metrics, demonstrating strong potential for NPPs anomaly diagnosis. Experimental results suggest that a feature count of 50 optimizes DDPM performance for NPPs anomaly detection, while the noise-to-noise training strategy improves model robustness.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105521"},"PeriodicalIF":3.3,"publicationDate":"2024-11-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653786","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 : 2024-11-08DOI: 10.1016/j.pnucene.2024.105529
YaoDi Li , Mei Huang , Boxue Wang , Xiangyuan Meng , YanTing Cheng
In this study, thermal hydraulic behaviors in a 19-pin bundle fuel assembly with nonuniform wire pitches is investigated by combing CFD with the Kriging method. To optimize the design, two geometric variables—the ratio of inner pitch to reference pitch (Pi/P) and the ratio of outer pitch to reference pitch (Po/P)—are selected, and the design space is sampled using Latin Hypercube Sampling (LHS). The sampled points are then subjected to CFD analysis. Convergence is considered achieved when the residuals of all variables are below 1e-5. The optimization problem aims to minimize the objective function, which is a linear combination of the cross-sectional temperature difference and friction factor. Sequential Quadratic Programming (SQP) is employed to search for the optimal point using a constructed meta-model. When compared to the reference shape, the optimal shape exhibits higher axial velocity in the inner channel, higher average temperature, smaller temperature difference at the outlet section, and reduced pressure drop in the fuel assembly. The Kriging model accurately predicts the cross-sectional temperature difference and friction coefficient for the optimal shape, consistent with the CFD calculation results. This confirms the accuracy and feasibility of the Kriging model in fuel assembly optimization.
{"title":"Numerical study on the hydrothermal characteristics of a wire-wrapped rod bundle with nonuniform wire pitches","authors":"YaoDi Li , Mei Huang , Boxue Wang , Xiangyuan Meng , YanTing Cheng","doi":"10.1016/j.pnucene.2024.105529","DOIUrl":"10.1016/j.pnucene.2024.105529","url":null,"abstract":"<div><div>In this study, thermal hydraulic behaviors in a 19-pin bundle fuel assembly with nonuniform wire pitches is investigated by combing CFD with the Kriging method. To optimize the design, two geometric variables—the ratio of inner pitch to reference pitch (Pi/P) and the ratio of outer pitch to reference pitch (Po/P)—are selected, and the design space is sampled using Latin Hypercube Sampling (LHS). The sampled points are then subjected to CFD analysis. Convergence is considered achieved when the residuals of all variables are below 1e-5. The optimization problem aims to minimize the objective function, which is a linear combination of the cross-sectional temperature difference and friction factor. Sequential Quadratic Programming (SQP) is employed to search for the optimal point using a constructed meta-model. When compared to the reference shape, the optimal shape exhibits higher axial velocity in the inner channel, higher average temperature, smaller temperature difference at the outlet section, and reduced pressure drop in the fuel assembly. The Kriging model accurately predicts the cross-sectional temperature difference and friction coefficient for the optimal shape, consistent with the CFD calculation results. This confirms the accuracy and feasibility of the Kriging model in fuel assembly optimization.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105529"},"PeriodicalIF":3.3,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653772","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 : 2024-11-08DOI: 10.1016/j.pnucene.2024.105516
Rolando Calabrese , Shun Hirooka
Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors (FBRs) confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.
{"title":"Comparison of correlations for thermal creep of FBR MOX","authors":"Rolando Calabrese , Shun Hirooka","doi":"10.1016/j.pnucene.2024.105516","DOIUrl":"10.1016/j.pnucene.2024.105516","url":null,"abstract":"<div><div>Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors (FBRs) confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105516"},"PeriodicalIF":3.3,"publicationDate":"2024-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653767","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-07DOI: 10.1016/j.pnucene.2024.105528
Zijian Huang , Hongkang Tian , Mengke Cai , Tenglong Cong , Yao Xiao , Hanyang Gu
Helical cruciform fuel (HCF) has the advantages of larger heat transfer area, enhanced coolant mixing and self-supporting, which contribute to increasing power density and safety margins. Compared with the square lattice configuration, the hexagonal arrangement of HCF assembly is more compact, which can help achieve a higher power density. In this paper, the flow characteristics and heat transfer behaviors of HCF in hexagonal lattice were predicted at high and low vapor quality during boiling crisis based on Eulerian two-fluid model. The influence of twist pitches and cross-sections of the fuel rod on heat transfer efficiency and fuel temperature was also studied. The cross-flow intensity changed periodically with a 30° cycle at low vapor quality, and did not fluctuate periodically at high vapor quality, which decreased with the increase of flow resistance. The highest heat flux of HCF rod was the at the blade root and the lowest was at the blade tip, and the maximum to average heat flux ratio was about 1.8. The peak vapor fraction and temperature occurred at leeside side of the fuel rods. The increase of the twist pitch reduced the critical heat flux (CHF), and the increase of blade length enhanced the non-uniformity of heat flux distribution. During boiling crisis, the maximum temperature of the fuel was lower than the phase transition temperature of U-50 wt%Zr alloy, which means the cladding meltdown caused by boiling crisis will occur before phase transition of the fuel.
{"title":"Numerical investigation on boiling crisis characteristic of a 7-rod HCF assembly in hexagonal lattice","authors":"Zijian Huang , Hongkang Tian , Mengke Cai , Tenglong Cong , Yao Xiao , Hanyang Gu","doi":"10.1016/j.pnucene.2024.105528","DOIUrl":"10.1016/j.pnucene.2024.105528","url":null,"abstract":"<div><div>Helical cruciform fuel (HCF) has the advantages of larger heat transfer area, enhanced coolant mixing and self-supporting, which contribute to increasing power density and safety margins. Compared with the square lattice configuration, the hexagonal arrangement of HCF assembly is more compact, which can help achieve a higher power density. In this paper, the flow characteristics and heat transfer behaviors of HCF in hexagonal lattice were predicted at high and low vapor quality during boiling crisis based on Eulerian two-fluid model. The influence of twist pitches and cross-sections of the fuel rod on heat transfer efficiency and fuel temperature was also studied. The cross-flow intensity changed periodically with a 30° cycle at low vapor quality, and did not fluctuate periodically at high vapor quality, which decreased with the increase of flow resistance. The highest heat flux of HCF rod was the at the blade root and the lowest was at the blade tip, and the maximum to average heat flux ratio was about 1.8. The peak vapor fraction and temperature occurred at leeside side of the fuel rods. The increase of the twist pitch reduced the critical heat flux (CHF), and the increase of blade length enhanced the non-uniformity of heat flux distribution. During boiling crisis, the maximum temperature of the fuel was lower than the phase transition temperature of U-50 wt%Zr alloy, which means the cladding meltdown caused by boiling crisis will occur before phase transition of the fuel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105528"},"PeriodicalIF":3.3,"publicationDate":"2024-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142653768","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 : 2024-11-06DOI: 10.1016/j.pnucene.2024.105517
Wei Liu , Jin Shi , Yang Liu , Yuhang Chen , Pan Wu , Kun Hou , Xuelin Li , Ying Zhang , Maogang He
Helium is a commonly used circulating working fluid in high-temperature gas-cooled reactors (HTGR). The thermophysical properties of helium are crucial for HTGR design and operation. The isobaric specific heat capacity, viscosity and thermal conductivity of helium were determined in this study based on flow method, capillary method and dynamic light scattering (DLS) method, respectively. To fill the data gap, the measurements were conducted over a temperature range of 293 K∼773 K and at pressures up to 7 MPa. The relative uncertainty estimates for the experimental apparatuses of isobaric specific heat capacity, viscosity, and thermal conductivity are less than 0.9%, 1.4%, and 2.2%, respectively. Based on the experimental data, the deviation of the existing calculation models for isobaric specific heat capacity, viscosity and thermal conductivity were analyzed. The calculation model posted by the Nuclear Safety Standards Commission (KTA) was modified to improve the reliability in the target p-T region.
{"title":"Measurement of helium thermophysical properties and modification of the calculation models in the KTA 3102.1 report","authors":"Wei Liu , Jin Shi , Yang Liu , Yuhang Chen , Pan Wu , Kun Hou , Xuelin Li , Ying Zhang , Maogang He","doi":"10.1016/j.pnucene.2024.105517","DOIUrl":"10.1016/j.pnucene.2024.105517","url":null,"abstract":"<div><div>Helium is a commonly used circulating working fluid in high-temperature gas-cooled reactors (HTGR). The thermophysical properties of helium are crucial for HTGR design and operation. The isobaric specific heat capacity, viscosity and thermal conductivity of helium were determined in this study based on flow method, capillary method and dynamic light scattering (DLS) method, respectively. To fill the data gap, the measurements were conducted over a temperature range of 293 K∼773 K and at pressures up to 7 MPa. The relative uncertainty estimates for the experimental apparatuses of isobaric specific heat capacity, viscosity, and thermal conductivity are less than 0.9%, 1.4%, and 2.2%, respectively. Based on the experimental data, the deviation of the existing calculation models for isobaric specific heat capacity, viscosity and thermal conductivity were analyzed. The calculation model posted by the Nuclear Safety Standards Commission (KTA) was modified to improve the reliability in the target <em>p</em>-<em>T</em> region.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105517"},"PeriodicalIF":3.3,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142592586","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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