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}
Pub Date : 2024-11-06DOI: 10.1016/j.pnucene.2024.105523
Tianxiang Wang , Changyou Zhao , Jun Lin , Shengli Chen , Mingtao He , Hao Chen , Hao Yang , Zhuo Li
In current neutron transport equation calculations, a common strategy to enhance computational efficiency is approximating the fuel temperature distribution as uniform. This approach defines a uniform temperature, known as the effective temperature (), which preserves the reactivity of the model corresponding to the actual temperature gradient. Based on the actual temperature distributions within the UO2 pellet of the AFA-3G fuel, the present study introduces a weighting coefficient for the volume-averaged temperature to extend the Chabert-Santamarina model. This extended model, as a generalized formulation of both the Rowlands and Chabert-Santamarina models, demonstrates superior performance to the current five effective temperature models in terms of deviation fluctuation and central values. This conclusion is verified through independent simulations, with OpenMC employing the Windowed Multi-Pole (WMP) database and RMC utilizing point-wise ACE library or data derived from Gaussian-Hermite quadrature for online Doppler broadening. Therefore, the present refined effective model enhances the accuracy of effective temperature and the resulting reactivity. Furthermore, the deviation of using a uniform effective temperature across the assembly remains within the acceptable uncertainty range (3 ).
{"title":"Enhancing the effective temperature model for typical UO2 fuel in criticality calculations","authors":"Tianxiang Wang , Changyou Zhao , Jun Lin , Shengli Chen , Mingtao He , Hao Chen , Hao Yang , Zhuo Li","doi":"10.1016/j.pnucene.2024.105523","DOIUrl":"10.1016/j.pnucene.2024.105523","url":null,"abstract":"<div><div>In current neutron transport equation calculations, a common strategy to enhance computational efficiency is approximating the fuel temperature distribution as uniform. This approach defines a uniform temperature, known as the effective temperature (<span><math><mrow><msub><mi>T</mi><mtext>eff</mtext></msub></mrow></math></span>), which preserves the reactivity of the model corresponding to the actual temperature gradient. Based on the actual temperature distributions within the UO<sub>2</sub> pellet of the AFA-3G fuel, the present study introduces a weighting coefficient for the volume-averaged temperature to extend the Chabert-Santamarina model. This extended model, as a generalized formulation of both the Rowlands and Chabert-Santamarina models, demonstrates superior performance to the current five effective temperature models in terms of deviation fluctuation and central values. This conclusion is verified through independent simulations, with OpenMC employing the Windowed Multi-Pole (WMP) database and RMC utilizing point-wise ACE library or data derived from Gaussian-Hermite quadrature for online Doppler broadening. Therefore, the present refined effective model enhances the accuracy of effective temperature and the resulting reactivity. Furthermore, the deviation of using a uniform effective temperature across the assembly remains within the acceptable uncertainty range (3 <span><math><mrow><mi>σ</mi></mrow></math></span>).</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105523"},"PeriodicalIF":3.3,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142592587","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.2024.105525
Kepiao Li , Zhiyuan Wu , Kui Zhang , Wenxi Tian , Suizheng Qiu
Upper plenum entrainment phenomenon occurs in the automatic depressurization process during the Small Break Loss of Coolant Accident (SBLOCA) in reactor pressure vessel, which may result in reactor disaster. The upper plenum entrainment experiments with and without reactor internals were carried out with air-water and steam-water as working mediums on the Automatic Depressurization and Entrainment Test Loop for Upper plenum entrainment (ADETEL-U) which scaled after AP1000 nuclear reactor. The experimental phenomena were observed by visualization method and the reliable data were collected and analyzed. The results indicate that the entrainment rate will increase with the increase of gas flow rate under the same , and the entrainment rate will decrease significantly with the decrease of the mixed liquid level when the range of is low. The results confirm that a large number of liquid droplets will be deposited on the surface of the reactor internals, which greatly reduces the entrainment rate. Under the same conditions, the entrainment rate with the reactor internals is about 10% of that without the reactor internals. There is a huge discrepancy between the existing pool entrainment rate models and the experimental data, with the maximum deviation of 200 times. Based on the experimental results, new upper plenum entrainment models for near surface region and high gas flux region of momentum controlled region are proposed. The error decreases by orders of magnitude compared to existing models, which suggested that the new model can accurately predict upper plenum entrainment phenomenon in the pressure vessel.
{"title":"Experimental and theoretical research on upper plenum entrainment with air-water and steam-water","authors":"Kepiao Li , Zhiyuan Wu , Kui Zhang , Wenxi Tian , Suizheng Qiu","doi":"10.1016/j.pnucene.2024.105525","DOIUrl":"10.1016/j.pnucene.2024.105525","url":null,"abstract":"<div><div>Upper plenum entrainment phenomenon occurs in the automatic depressurization process during the Small Break Loss of Coolant Accident (SBLOCA) in reactor pressure vessel, which may result in reactor disaster. The upper plenum entrainment experiments with and without reactor internals were carried out with air-water and steam-water as working mediums on the Automatic Depressurization and Entrainment Test Loop for Upper plenum entrainment (ADETEL-U) which scaled after AP1000 nuclear reactor. The experimental phenomena were observed by visualization method and the reliable data were collected and analyzed. The results indicate that the entrainment rate will increase with the increase of gas flow rate under the same <span><math><mrow><msubsup><mi>h</mi><mi>g</mi><mo>∗</mo></msubsup></mrow></math></span>, and the entrainment rate will decrease significantly with the decrease of the mixed liquid level when the range of <span><math><mrow><msubsup><mi>h</mi><mi>g</mi><mo>∗</mo></msubsup></mrow></math></span> is low. The results confirm that a large number of liquid droplets will be deposited on the surface of the reactor internals, which greatly reduces the entrainment rate. Under the same conditions, the entrainment rate with the reactor internals is about 10% of that without the reactor internals. There is a huge discrepancy between the existing pool entrainment rate models and the experimental data, with the maximum deviation of 200 times. Based on the experimental results, new upper plenum entrainment models for near surface region and high gas flux region of momentum controlled region are proposed. The error decreases by orders of magnitude compared to existing models, which suggested that the new model can accurately predict upper plenum entrainment phenomenon in the pressure vessel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105525"},"PeriodicalIF":3.3,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142592585","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 : 2024-11-05DOI: 10.1016/j.pnucene.2024.105524
Cheng Peng , Chengfa Cao , Jiang Wu , Jian Deng
Liquid sodium can be treated as a prominent medium in many industrial fields, such as photovoltaic technology, chemical synthesis, nuclear industry, etc. However, it poses significant threats to the normal operation of related systems and facilities, and human life as well, due to its potential combustion risk, particularly when multi-leakages take place. Sodium spray combustion is the most severe one, in which spray dynamic process may intensify the heat transfer and subsequent combustion process. In this work, the applicability of the droplet break-up model is firstly confirmed using numerical simulations of liquid sodium spray by Fluent code, and the impact of spray interference on combustion kinetics is examined. The Euler-Lagrange approach, which accounts for droplet break-up, collision, and agglomeration during the spray combustion process, is used in the simulation. Three-dimensional simulations of liquid sodium spray fire are then conducted, in the light of two classical experiments all around the world. The simulated volume-mean air temperature shows an error margin of less than 4%. The thermodynamic characteristics of sodium spray fire in the situation of dual-jets is further investigated. The findings indicate that the spray interference has a greater impact on the sodium content threshold and the corresponding time at which the threshold can be achieved than temperature. When the nozzle spacing varies, the consequences of the spray interference on the droplets’ combustion change. The break-up impact outweighs the agglomeration effect when the nozzle spacing is larger, while the agglomeration effect is relatively stronger when the nozzle spacing is short. This conclusion can be appropriate under both low and high flow rate of liquid sodium. The present work can provide detailed information and mechanism on spray combustion under both single jet and dual-jets conditions, which is beneficial for the evaluation of the risk of real sodium spray fire in any closed environment.
{"title":"Comparison and analysis of combustion characteristics and interference effect between single burning sodium jet and the dual-jets","authors":"Cheng Peng , Chengfa Cao , Jiang Wu , Jian Deng","doi":"10.1016/j.pnucene.2024.105524","DOIUrl":"10.1016/j.pnucene.2024.105524","url":null,"abstract":"<div><div>Liquid sodium can be treated as a prominent medium in many industrial fields, such as photovoltaic technology, chemical synthesis, nuclear industry, etc. However, it poses significant threats to the normal operation of related systems and facilities, and human life as well, due to its potential combustion risk, particularly when multi-leakages take place. Sodium spray combustion is the most severe one, in which spray dynamic process may intensify the heat transfer and subsequent combustion process. In this work, the applicability of the droplet break-up model is firstly confirmed using numerical simulations of liquid sodium spray by Fluent code, and the impact of spray interference on combustion kinetics is examined. The Euler-Lagrange approach, which accounts for droplet break-up, collision, and agglomeration during the spray combustion process, is used in the simulation. Three-dimensional simulations of liquid sodium spray fire are then conducted, in the light of two classical experiments all around the world. The simulated volume-mean air temperature shows an error margin of less than 4%. The thermodynamic characteristics of sodium spray fire in the situation of dual-jets is further investigated. The findings indicate that the spray interference has a greater impact on the sodium content threshold and the corresponding time at which the threshold can be achieved than temperature. When the nozzle spacing varies, the consequences of the spray interference on the droplets’ combustion change. The break-up impact outweighs the agglomeration effect when the nozzle spacing is larger, while the agglomeration effect is relatively stronger when the nozzle spacing is short. This conclusion can be appropriate under both low and high flow rate of liquid sodium. The present work can provide detailed information and mechanism on spray combustion under both single jet and dual-jets conditions, which is beneficial for the evaluation of the risk of real sodium spray fire in any closed environment.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105524"},"PeriodicalIF":3.3,"publicationDate":"2024-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142586262","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 : 2024-11-04DOI: 10.1016/j.pnucene.2024.105522
Xiuli Wang , Shenpeng Yang , YiFan Zhi , Wei Xu
Condition monitoring and identification are effective ways to ensure the safe and reliable operation of nuclear power pumps. However, the condition monitoring of the cutting impeller is blank. In order to effectively monitor and identify the operating status of nuclear power pump impellers corresponding to different cutting amounts. The paper collects the measured stator current signals of nuclear power pumps with 6 cutting quantities under 13 operating conditions. Variational Mode Decomposition (VMD) and Empirical Mode Decomposition (EMD) methods are utilized to analyze the state characteristics of the collected signals. The effect of different blade cutting amount on the signal characteristics of nuclear electric pump is obtained. The research results indicate as follow: when a single method is used to identify large impeller flow, the diagnostic accuracy of EMD and VMD can reach more than 90%, while Total Harmonic Distortion (THD) is less than 70%, and even less than 20% in some areas. However, for different impeller diameters and different flow rates, the identification accuracy of EMD and VMD is relatively low, only 60%. Under special working conditions, it can even be lower, with only about 50% at low flow rates between 0.2Q0-0.3Q0. EMD-VMD can accurately identify impellers of different diameters and different flow rates, and the accuracy of fault identification can be improved to over 90%, even higher than 95% in the range of 0.7Q0-1.2Q0. At the same time, the minimum flow rates of 0.2Q0 can also achieve 80% accuracy, which can effectively achieve fault diagnosis. The research results can provide data support for monitoring the operating status of self-cutting centrifugal pumps, which is of great significance for safe and stable operation.
{"title":"Research on impeller cutting of the nuclear pump based on MCSA","authors":"Xiuli Wang , Shenpeng Yang , YiFan Zhi , Wei Xu","doi":"10.1016/j.pnucene.2024.105522","DOIUrl":"10.1016/j.pnucene.2024.105522","url":null,"abstract":"<div><div>Condition monitoring and identification are effective ways to ensure the safe and reliable operation of nuclear power pumps. However, the condition monitoring of the cutting impeller is blank. In order to effectively monitor and identify the operating status of nuclear power pump impellers corresponding to different cutting amounts. The paper collects the measured stator current signals of nuclear power pumps with 6 cutting quantities under 13 operating conditions. Variational Mode Decomposition (VMD) and Empirical Mode Decomposition (EMD) methods are utilized to analyze the state characteristics of the collected signals. The effect of different blade cutting amount on the signal characteristics of nuclear electric pump is obtained. The research results indicate as follow: when a single method is used to identify large impeller flow, the diagnostic accuracy of EMD and VMD can reach more than 90%, while Total Harmonic Distortion (THD) is less than 70%, and even less than 20% in some areas. However, for different impeller diameters and different flow rates, the identification accuracy of EMD and VMD is relatively low, only 60%. Under special working conditions, it can even be lower, with only about 50% at low flow rates between 0.2Q<sub>0</sub>-0.3Q<sub>0</sub>. EMD-VMD can accurately identify impellers of different diameters and different flow rates, and the accuracy of fault identification can be improved to over 90%, even higher than 95% in the range of 0.7Q<sub>0</sub>-1.2Q<sub>0</sub>. At the same time, the minimum flow rates of 0.2Q<sub>0</sub> can also achieve 80% accuracy, which can effectively achieve fault diagnosis. The research results can provide data support for monitoring the operating status of self-cutting centrifugal pumps, which is of great significance for safe and stable operation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105522"},"PeriodicalIF":3.3,"publicationDate":"2024-11-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142578737","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 : 2024-11-02DOI: 10.1016/j.pnucene.2024.105520
Timothy G. Lane, Shripad T. Revankar
A group of small nuclear reactors that are less than 20 MWe are often referred to as microreactors. This review provides recent advances in the nuclear reactor fuel and core design technology leading to compact microreactor designs, design features, types of microreactors currently considered in the industry and studied by researcher, regulatory design criteria, and deployment potentials for these new microreactors. This review indicates that there are a wide variety of microreactor designs being developed, some of which use coolant other than water such as liquid metal (e.g., sodium), helium gas or molten salt in order to achieve their operational objectives. Some of these designs utilize passive heat pipes in order to transfer heat from the reactor cores. Others make use of helium gas due to its compatibility at high temperature and inert nature. Currently there are no operating reactors which utilize either of these technologies for power generation. To aid the technology commercialization the nuclear regulatory bodies like US NRC are developing new design criteria and the licensing process to assess the microreactors for design certification, construction, and operation. The review indicates that there are potential design criteria challenges for the microreactors. For example, helium reactors need to show that the heat can be dispersed efficiently and passively, and the heat pipe reactors need to demonstrate that the coolant in their heat pipes will not escape the primary boundary. The US NRC has developed design criteria for microreactors are highlighted in the review.
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