The radiation field measurement and surveying play important roles in optimizing and planning operation work in the radioactive area. In this work, a 3-D radiation field reconstruction method is applied to reconstruct a 3-D radiation field of a facility with two radioactive drums and a shielding, not limited to the existing methods under two-dimensional grids. A numerical scheme of the method is introduced. Monte Carlo simulation of the gamma radiation field was performed to be an original field so that the reconstruction results can be verified. Random and regular sampling ways are taken into consideration while the sampling rates are kept at 4.12%. Quantitative evaluation of the results evaluates MSE (mean relative error) and MRE (mean squared error), MRE is kept at less than 4%, which shows a good reconstruction accuracy of the method in almost all 3D space. The work in this paper has good reference value for applications such as radiation field detection, inversion, and reconstruction, operations under radioactive environments like nuclear waste decommissioning, and nuclear power radiation modeling.
{"title":"3-D Radiation Field Reconstruction for a Facility With Multiple Radioactive Sources","authors":"Shangzhen Zhu, Xinwen Dong, Shuhan Zhuang, Shengjiang Fang, Jian-zhu Cao, Wenqian Li","doi":"10.1115/icone29-92587","DOIUrl":"https://doi.org/10.1115/icone29-92587","url":null,"abstract":"\u0000 The radiation field measurement and surveying play important roles in optimizing and planning operation work in the radioactive area. In this work, a 3-D radiation field reconstruction method is applied to reconstruct a 3-D radiation field of a facility with two radioactive drums and a shielding, not limited to the existing methods under two-dimensional grids. A numerical scheme of the method is introduced. Monte Carlo simulation of the gamma radiation field was performed to be an original field so that the reconstruction results can be verified. Random and regular sampling ways are taken into consideration while the sampling rates are kept at 4.12%. Quantitative evaluation of the results evaluates MSE (mean relative error) and MRE (mean squared error), MRE is kept at less than 4%, which shows a good reconstruction accuracy of the method in almost all 3D space. The work in this paper has good reference value for applications such as radiation field detection, inversion, and reconstruction, operations under radioactive environments like nuclear waste decommissioning, and nuclear power radiation modeling.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128760854","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Flow boiling in vertical pipes widely exists in nuclear reactors. The bubble motion and behavior attract more and more attention due to their influence on the heat transfer capacity. However, the most used two-phase flow model in numerical simulation, called Eulerian two-fluid model, is not able to obtain parameters of phase interfaces and provide an observed image of bubble movements. Therefore, the present study aims to simulate the bubble behavior and trajectory in the saturated flow boiling in the vertical pipe using VOF model which can treat the deformation of interfaces. Firstly, the bubble parameters were calculated based on existing bubble dynamic models. Then, the random generation of bubbles was simulated by directly adding mass and momentum sources into the flow region by the User Defined Function (UDF). Finally, the deformation, collapse, and coalescence of bubbles were simulated and predicted in bubbly flow. The feasibility of the newly proposed model is proved by comparing with experimental data and other simulation results. The present study provides a new idea for simulating the movement of a large number of bubbles in the saturated flow boiling in vertical pipes, which adopts bubble parameters as boundary conditions directly, instead of heat and mass transfer models. In comparison, the present model can greatly improve the calculation efficiency.
{"title":"Bubble Behaviors in Saturated Flow Boiling Based on the VOF Model","authors":"Deyang Gao, Wen He, H. Bo","doi":"10.1115/icone29-92489","DOIUrl":"https://doi.org/10.1115/icone29-92489","url":null,"abstract":"\u0000 Flow boiling in vertical pipes widely exists in nuclear reactors. The bubble motion and behavior attract more and more attention due to their influence on the heat transfer capacity. However, the most used two-phase flow model in numerical simulation, called Eulerian two-fluid model, is not able to obtain parameters of phase interfaces and provide an observed image of bubble movements. Therefore, the present study aims to simulate the bubble behavior and trajectory in the saturated flow boiling in the vertical pipe using VOF model which can treat the deformation of interfaces. Firstly, the bubble parameters were calculated based on existing bubble dynamic models. Then, the random generation of bubbles was simulated by directly adding mass and momentum sources into the flow region by the User Defined Function (UDF). Finally, the deformation, collapse, and coalescence of bubbles were simulated and predicted in bubbly flow. The feasibility of the newly proposed model is proved by comparing with experimental data and other simulation results. The present study provides a new idea for simulating the movement of a large number of bubbles in the saturated flow boiling in vertical pipes, which adopts bubble parameters as boundary conditions directly, instead of heat and mass transfer models. In comparison, the present model can greatly improve the calculation efficiency.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129388109","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In the floating nuclear power plant, Core Make-up Tank (CMT) is one of the passive safety facilities, has been widely used in nuclear engineering. Under emergency conditions, natural circulation occurs and consequent thermal stratification appears in CMT system. However, rolling motion will directly affect the natural circulation and destroy the formation of thermal stratification, so there will be uncertain and unknown change of the safety injection capacity of CMT. Based on above analysis, this work carried out experiments under rolling conditions to reveal natural circulation and thermal stratification characteristics in CMT. A temperature detecting matrix formed by thermocouples is deployed in CMT. It is concluded that thermal stratification distribution is affected by the rolling conditions. Rolling conditions increase the thickness of thermal stratification and reduce the temperature gradient. The CMT natural circulation has two stages, the influence of rolling motion on the average flow in the first stage of natural circulation is very weak, but rolling motion will significantly affect the second stage of natural circulation and advance the decline time of direct vessel injection (DVI) flow rate. The transition time of natural circulation stage and the flow change in the second stage of natural circulation are related to the thermal stratification in CMT. The impact of rolling conditions should not be ignored when evaluating the performance of CMT.
{"title":"Experimental Investigation on Natural Circulation Characteristics of Core Make-Up Tank Under Rolling Conditions","authors":"Shuguang Wang, Jinlei Zhang, Ya'ou Shen, Xiaobo Liu, Shuhua Ding, Yusheng Liu, Dongyang Li, Jilin Tang","doi":"10.1115/icone29-92254","DOIUrl":"https://doi.org/10.1115/icone29-92254","url":null,"abstract":"\u0000 In the floating nuclear power plant, Core Make-up Tank (CMT) is one of the passive safety facilities, has been widely used in nuclear engineering. Under emergency conditions, natural circulation occurs and consequent thermal stratification appears in CMT system. However, rolling motion will directly affect the natural circulation and destroy the formation of thermal stratification, so there will be uncertain and unknown change of the safety injection capacity of CMT. Based on above analysis, this work carried out experiments under rolling conditions to reveal natural circulation and thermal stratification characteristics in CMT. A temperature detecting matrix formed by thermocouples is deployed in CMT. It is concluded that thermal stratification distribution is affected by the rolling conditions. Rolling conditions increase the thickness of thermal stratification and reduce the temperature gradient. The CMT natural circulation has two stages, the influence of rolling motion on the average flow in the first stage of natural circulation is very weak, but rolling motion will significantly affect the second stage of natural circulation and advance the decline time of direct vessel injection (DVI) flow rate. The transition time of natural circulation stage and the flow change in the second stage of natural circulation are related to the thermal stratification in CMT. The impact of rolling conditions should not be ignored when evaluating the performance of CMT.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126110446","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Water hammer is common in nuclear engineering pipeline systems, such as feedwater systems, essential service water systems and passive residual heat removal systems, etc. Water hammer can cause a large fluctuation of fluid pressure, posing a threat to the normal operation of pipeline systems, so it is necessary to reduce the pressure fluctuation and energy. The flow resistance components such as the orifice plate are the key parts to mitigate the influence of water hammer. For the reservoir-pipe-valve (RPV) system, a water hammer model that considered the effect of unsteady friction was established and solved by the method of characteristics. The calculation results of theoretical model agreed well with the experimental results. On this basis, the orifice plate was added to analyze its influence on the transient flow process. Based on the traditional analysis method and energy analysis method, the dissipation effects of the frictional resistance and the local resistance of pipeline system were studied. The analysis results show that as the orifice plate moves from upstream to downstream of pipeline system, the fluctuation of internal energy and kinetic energy decreases, the orifice dissipation work increases, and the friction dissipation work decreases. The research results show the advantage of energy analysis method and provide guidance for the design of energy dissipation equipment.
{"title":"Analysis of Energy Dissipation Effects of Resistance Components in the Water Hammer Process","authors":"Linqing Yang, Benke Qin, H. Bo","doi":"10.1115/icone29-90915","DOIUrl":"https://doi.org/10.1115/icone29-90915","url":null,"abstract":"\u0000 Water hammer is common in nuclear engineering pipeline systems, such as feedwater systems, essential service water systems and passive residual heat removal systems, etc. Water hammer can cause a large fluctuation of fluid pressure, posing a threat to the normal operation of pipeline systems, so it is necessary to reduce the pressure fluctuation and energy. The flow resistance components such as the orifice plate are the key parts to mitigate the influence of water hammer. For the reservoir-pipe-valve (RPV) system, a water hammer model that considered the effect of unsteady friction was established and solved by the method of characteristics. The calculation results of theoretical model agreed well with the experimental results. On this basis, the orifice plate was added to analyze its influence on the transient flow process. Based on the traditional analysis method and energy analysis method, the dissipation effects of the frictional resistance and the local resistance of pipeline system were studied. The analysis results show that as the orifice plate moves from upstream to downstream of pipeline system, the fluctuation of internal energy and kinetic energy decreases, the orifice dissipation work increases, and the friction dissipation work decreases. The research results show the advantage of energy analysis method and provide guidance for the design of energy dissipation equipment.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"59 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124182331","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Tianyi Wei, Guanhui Xie, Dongyang Li, S. Tan, Yangyang Du, Zhongyi Li, Yuan Wang
Liquid sloshing will occur in liquid storage tanks such as accumulator of floating nuclear power plant (FNPP) subjected to additional inertial forces under motion conditions. The study carried out measurement experiments based on the 6-DOF platform to study the sloshing characteristics and pressure variation rule of the accumulator. The results show that surging will induce many kinds of nonlinear free surface sloshing forms, it can be seen that the law of pressure variation is mainly dominated by natural frequency and excitation frequency based on time and frequency domain analysis. Then the study combines the automatic encoder and extreme learning machine to build the deep extreme learning machine (DELM) network to predict the pressure in time series. Based on the phase space reconstruction of the time sequence, the pressure results of the next time are output after the last 15 pressure data are input. The prediction results show that the DELM model has fast speed and high precision and the predicted value is in good agreement with the experimental data. So this study can provide a reference for the pressure monitoring and the artificial intelligence application of FNPP.
{"title":"Experimental Study and Intelligent Prediction on Pressure Fluctuation of Accumulator Under Ocean Conditions","authors":"Tianyi Wei, Guanhui Xie, Dongyang Li, S. Tan, Yangyang Du, Zhongyi Li, Yuan Wang","doi":"10.1115/icone29-89212","DOIUrl":"https://doi.org/10.1115/icone29-89212","url":null,"abstract":"\u0000 Liquid sloshing will occur in liquid storage tanks such as accumulator of floating nuclear power plant (FNPP) subjected to additional inertial forces under motion conditions. The study carried out measurement experiments based on the 6-DOF platform to study the sloshing characteristics and pressure variation rule of the accumulator. The results show that surging will induce many kinds of nonlinear free surface sloshing forms, it can be seen that the law of pressure variation is mainly dominated by natural frequency and excitation frequency based on time and frequency domain analysis. Then the study combines the automatic encoder and extreme learning machine to build the deep extreme learning machine (DELM) network to predict the pressure in time series. Based on the phase space reconstruction of the time sequence, the pressure results of the next time are output after the last 15 pressure data are input. The prediction results show that the DELM model has fast speed and high precision and the predicted value is in good agreement with the experimental data. So this study can provide a reference for the pressure monitoring and the artificial intelligence application of FNPP.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129331748","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Due to the small size, the high inherent safety, and easy modularization and expansion, heat-pipe cooled reactor (HPR) has broad application prospects in deep space exploration, deep-sea submarine exploration and other scenarios. However, the HPR conducts thermal energy through the evaporation and condensation of the working fluid inside the heat pipe. This feature makes the HPR a large time-delay system. If the power control system adopts the conventional PID algorithm, there will be a long settling time and large overshoot. Therefore, the model predictive control algorithm is proposed for the power control system to improve the control performance. The HPR linear model, which is developed by linearization of its nonlinear model, is chosen as the predictive model. Based on the predictive model and the reactor power feedback value, the optimal control value is obtained by solving the optimization problem. Due to the discrepancy between the predive model and the actual system response, a steady-state error occurs. To solve this problem, an integral controller is added before the model predictive controller to eliminate the error. The appropriate control system parameters are tuned by trial and error method. Finally, taking the nonlinear HPR model as the controlled object, it is verified by typical transient that the control system has satisfactory control performance. The model predictive control can effectively overcome the influence of the large time-delay characteristics.
{"title":"Power Control System Design for Heat Pipe Cooled Reactor Based on Model Predictive Control","authors":"Jiajun Huang, Peiwei Sun, Songmao Pu, Xinyu Wei","doi":"10.1115/icone29-89495","DOIUrl":"https://doi.org/10.1115/icone29-89495","url":null,"abstract":"\u0000 Due to the small size, the high inherent safety, and easy modularization and expansion, heat-pipe cooled reactor (HPR) has broad application prospects in deep space exploration, deep-sea submarine exploration and other scenarios. However, the HPR conducts thermal energy through the evaporation and condensation of the working fluid inside the heat pipe. This feature makes the HPR a large time-delay system. If the power control system adopts the conventional PID algorithm, there will be a long settling time and large overshoot. Therefore, the model predictive control algorithm is proposed for the power control system to improve the control performance.\u0000 The HPR linear model, which is developed by linearization of its nonlinear model, is chosen as the predictive model. Based on the predictive model and the reactor power feedback value, the optimal control value is obtained by solving the optimization problem. Due to the discrepancy between the predive model and the actual system response, a steady-state error occurs. To solve this problem, an integral controller is added before the model predictive controller to eliminate the error. The appropriate control system parameters are tuned by trial and error method. Finally, taking the nonlinear HPR model as the controlled object, it is verified by typical transient that the control system has satisfactory control performance. The model predictive control can effectively overcome the influence of the large time-delay characteristics.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134212245","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Small pressurized water reactors (PWRs) have become a new trend in the current nuclear energy development due to their many advantages, such as compact equipment layout, high thermal efficiency, and strong cycle capability. Compared with large PWRs, small PWRs are designed to reduce the coolant inventory and increase the core power density, which is not good for nuclear safety. Severe accident studies on large PWRs cannot be directly applied to small PWRs. Loss of coolant accident is one of the main inducements of reactor core melting, which needs to be focused on prevention and treatment. It is of great significance for the safe operation of small PWRs to analyze and study severe accident induced by loss of coolant accident. In this paper, MELCOR is used to establish the severe accident analysis model of the primary loop system of a small PWR, and the loss of coolant accident is introduced to obtain the accident sequence from the shutdown of the reactor until the core degradation. At the same time, the core pressure, core liquid level and other key parameters are analyzed. The results show that in the case of a severe accident, compared with the large PWR, the small PWR takes a faster time for the pressure of the primary circuit and the containment pressure to reach equilibrium after the break accident occurs. The unbalanced radial power distribution causes the cladding of the 3rd ring to fail first. In the later stage of the severe accident, the melt plays a major role in heating the coolant. During the entire core degradation process, the upper fuel assemblies start to melt first, and the core does not completely collapse. The research results can provide reference for the formulation of severe accident management guidelines for small PWRs.
{"title":"Analysis of Severe Accident Induced by Loss of Coolant Accident for a Small Pressurized Water Reactor","authors":"Xiaolong Bi, Peiwei Sun, Xinyu Wei","doi":"10.1115/icone29-92167","DOIUrl":"https://doi.org/10.1115/icone29-92167","url":null,"abstract":"\u0000 Small pressurized water reactors (PWRs) have become a new trend in the current nuclear energy development due to their many advantages, such as compact equipment layout, high thermal efficiency, and strong cycle capability. Compared with large PWRs, small PWRs are designed to reduce the coolant inventory and increase the core power density, which is not good for nuclear safety. Severe accident studies on large PWRs cannot be directly applied to small PWRs. Loss of coolant accident is one of the main inducements of reactor core melting, which needs to be focused on prevention and treatment. It is of great significance for the safe operation of small PWRs to analyze and study severe accident induced by loss of coolant accident.\u0000 In this paper, MELCOR is used to establish the severe accident analysis model of the primary loop system of a small PWR, and the loss of coolant accident is introduced to obtain the accident sequence from the shutdown of the reactor until the core degradation. At the same time, the core pressure, core liquid level and other key parameters are analyzed.\u0000 The results show that in the case of a severe accident, compared with the large PWR, the small PWR takes a faster time for the pressure of the primary circuit and the containment pressure to reach equilibrium after the break accident occurs. The unbalanced radial power distribution causes the cladding of the 3rd ring to fail first. In the later stage of the severe accident, the melt plays a major role in heating the coolant. During the entire core degradation process, the upper fuel assemblies start to melt first, and the core does not completely collapse. The research results can provide reference for the formulation of severe accident management guidelines for small PWRs.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"53 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114703508","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Fei Zhao, D. Lu, Yu Liu, Dong Liu, Jiliang Xu, Jinghui Wu, Fei Xie, Yuchao Wang
There are a large number of equipment densely arranged in liquids in nuclear power plants, such as fuel assemblies, steam generator heat transfer tubes, spent fuel storage and transportation racks, etc. These equipment are complex in shape and compact in arrangement and have strong fluid-structure coupling effects under excitations, so the calculation is computationally intensive. For such complex structures, the use of porous media models is an important means of structure simplification. The parameters of porous media are often calculated by CFD modeling, and the calculation process is complicated and time-consuming. BP neural network has strong nonlinear mapping capability and can be used to calculate the parameters of porous media. For different racks designs, the gap arrangement is different, and the fluid-structure coupling parameters are also different. Therefore, it is necessary to study the fluid-structure coupling parameters of square tube bundles such as racks. Taking porous storage racks as an example, by building different CFD models, 1366 sets of valid data were obtained for training. This paper uses BP neural network to study the porous medium parameters required for fluid-structure interaction of porous racks. Compared with the CFD calculation method of fine modeling, the calculation error of the additional mass of the porous media model established by the porous media parameters predicted by the neural network is controlled at about 10%. The research results provide a reference for the fast calculation of porous media parameters and fluid-structure interaction.
{"title":"Parameter Calculation Method of Porous Media Based on BP Neural Network","authors":"Fei Zhao, D. Lu, Yu Liu, Dong Liu, Jiliang Xu, Jinghui Wu, Fei Xie, Yuchao Wang","doi":"10.1115/icone29-92382","DOIUrl":"https://doi.org/10.1115/icone29-92382","url":null,"abstract":"\u0000 There are a large number of equipment densely arranged in liquids in nuclear power plants, such as fuel assemblies, steam generator heat transfer tubes, spent fuel storage and transportation racks, etc. These equipment are complex in shape and compact in arrangement and have strong fluid-structure coupling effects under excitations, so the calculation is computationally intensive. For such complex structures, the use of porous media models is an important means of structure simplification. The parameters of porous media are often calculated by CFD modeling, and the calculation process is complicated and time-consuming. BP neural network has strong nonlinear mapping capability and can be used to calculate the parameters of porous media. For different racks designs, the gap arrangement is different, and the fluid-structure coupling parameters are also different. Therefore, it is necessary to study the fluid-structure coupling parameters of square tube bundles such as racks. Taking porous storage racks as an example, by building different CFD models, 1366 sets of valid data were obtained for training. This paper uses BP neural network to study the porous medium parameters required for fluid-structure interaction of porous racks. Compared with the CFD calculation method of fine modeling, the calculation error of the additional mass of the porous media model established by the porous media parameters predicted by the neural network is controlled at about 10%. The research results provide a reference for the fast calculation of porous media parameters and fluid-structure interaction.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121085672","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The calculation of three-dimensional dose rate fields plays a key role in radiation dose rate estimation and the service for the nuclear emergency. The recent fast calculation method based on the Fast Fourier Transform (FFT) method can greatly speed up the calculation without losing accuracy, which is promising for operational usage in nuclear emergency response systems. But it can only be used for a uniform grid. Unfortunately, most atmospheric dispersion models use a non-uniform grid, which prevents the direct application of FFT-based calculation. Therefore, interpolation is required beforehand to use the Fourier transform, which may introduce errors and affect computing efficiency. In this paper, an atmospheric dispersion modeling case of a typical nuclear power plant (NPP) is used to investigate the efficiency of different interpolation methods, which are based on a non-uniform grid. These methods are linear interpolation and nearest-neighbor interpolation. The sensitive analysis of grid resolution is investigated in the slices of x, y, and z at typical positions, which confirms the smooth-out and speed-up effects in rough grids. A grid size over 10 m at any slice commonly causes losses of change details of dose rate fields. Given the same resolution of 50 m × 50 m × 50 m, the nearest neighbor performs a 717 times calculation faster than the linear method, which preserves more change details of dose rate fields as well. For complex calculation tasks, e.,g., non-uniform NPP buildings, the nearest neighbor interpolation method is recommended with a resolution of 10 m × 10 m × 10 m to make a good balance between accuracy and speed.
三维剂量率场的计算在辐射剂量率估算和核应急服务中起着至关重要的作用。基于快速傅里叶变换(fast Fourier Transform, FFT)方法的快速计算方法可以在不损失计算精度的前提下大大提高计算速度，在核应急响应系统中具有应用前景。但它只能用于统一的网格。不幸的是，大多数大气色散模型使用非均匀网格，这阻碍了基于fft计算的直接应用。因此，在使用傅里叶变换之前需要进行插值，这可能会引入误差，影响计算效率。本文以典型核电站的大气扩散模型为例，研究了基于非均匀网格的不同插值方法的有效性。这些方法是线性插值和最近邻插值。对典型位置的x、y和z切片进行了网格分辨率的敏感性分析，证实了粗糙网格的平滑和加速效应。在任何切片上超过10米的网格尺寸通常会导致剂量率场变化细节的损失。在相同分辨率为50 m × 50 m × 50 m的情况下，最近邻法的计算速度比线性法快717倍，并保留了更多的剂量率场变化细节。对于复杂的计算任务，例如:对于非均匀核电厂建筑物，建议采用最近邻插值法，分辨率为10m × 10m × 10m，以便在精度和速度之间取得良好的平衡。
{"title":"Interpolation Influence on the Fast Fourier Transform Based Calculation of Three-Dimensional Dose Rate Field","authors":"Xinwen Dong, Shengjiang Fang, Shuhan Zhuang","doi":"10.1115/icone29-89244","DOIUrl":"https://doi.org/10.1115/icone29-89244","url":null,"abstract":"\u0000 The calculation of three-dimensional dose rate fields plays a key role in radiation dose rate estimation and the service for the nuclear emergency. The recent fast calculation method based on the Fast Fourier Transform (FFT) method can greatly speed up the calculation without losing accuracy, which is promising for operational usage in nuclear emergency response systems. But it can only be used for a uniform grid. Unfortunately, most atmospheric dispersion models use a non-uniform grid, which prevents the direct application of FFT-based calculation. Therefore, interpolation is required beforehand to use the Fourier transform, which may introduce errors and affect computing efficiency. In this paper, an atmospheric dispersion modeling case of a typical nuclear power plant (NPP) is used to investigate the efficiency of different interpolation methods, which are based on a non-uniform grid. These methods are linear interpolation and nearest-neighbor interpolation. The sensitive analysis of grid resolution is investigated in the slices of x, y, and z at typical positions, which confirms the smooth-out and speed-up effects in rough grids. A grid size over 10 m at any slice commonly causes losses of change details of dose rate fields. Given the same resolution of 50 m × 50 m × 50 m, the nearest neighbor performs a 717 times calculation faster than the linear method, which preserves more change details of dose rate fields as well. For complex calculation tasks, e.,g., non-uniform NPP buildings, the nearest neighbor interpolation method is recommended with a resolution of 10 m × 10 m × 10 m to make a good balance between accuracy and speed.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"124 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121367287","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Guangliang Yang, Wenpei Feng, Tao Ding, Hongli Chen
Fuel plate performance under irradiation is an important topic for the research reactor. The fuel plate is expected to deform after a long-time operation, which is a threat to the reactor safety. In this research, irradiation behaviors like interaction layer growth, swelling, creep and plasticity of the UMo/Al dispersion fuel plate are carefully reviewed from literatures. Taking these behaviors into consideration, the longtime three-dimensional stress update algorithms for the fuel meat and cladding are developed respectively based on the large-deformation incremental constitutive relation. Since local properties in the fuel meat change with burnup, equivalent material properties models are used to take account of this characteristic. These models and algorithms are implemented into Abaqus user subroutines to simulate the thermo-mechanical performance of the fuel plate under the typical operating condition in research reactors. In order to illustrate the macroscopic deformation of the fuel plate and reduce the computational cost, 1/4 symmetrical geometry of the fuel plate is employed in the simulation. The evaluation and distribution of temperature, stress and deformation of the fuel plate are obtained and analyzed. Results show that the growth of interaction layer affects the temperature profiles heavily. The stress concentration mainly locates at the contact interface between fuel meat and cladding, especially at the side of the interface.
{"title":"Integral Simulation on the Thermo-Mechanical Behavior of the UMo/Al Dispersion Fuel Plate Using Finite Element Method","authors":"Guangliang Yang, Wenpei Feng, Tao Ding, Hongli Chen","doi":"10.1115/icone29-90559","DOIUrl":"https://doi.org/10.1115/icone29-90559","url":null,"abstract":"\u0000 Fuel plate performance under irradiation is an important topic for the research reactor. The fuel plate is expected to deform after a long-time operation, which is a threat to the reactor safety. In this research, irradiation behaviors like interaction layer growth, swelling, creep and plasticity of the UMo/Al dispersion fuel plate are carefully reviewed from literatures. Taking these behaviors into consideration, the longtime three-dimensional stress update algorithms for the fuel meat and cladding are developed respectively based on the large-deformation incremental constitutive relation. Since local properties in the fuel meat change with burnup, equivalent material properties models are used to take account of this characteristic. These models and algorithms are implemented into Abaqus user subroutines to simulate the thermo-mechanical performance of the fuel plate under the typical operating condition in research reactors. In order to illustrate the macroscopic deformation of the fuel plate and reduce the computational cost, 1/4 symmetrical geometry of the fuel plate is employed in the simulation. The evaluation and distribution of temperature, stress and deformation of the fuel plate are obtained and analyzed. Results show that the growth of interaction layer affects the temperature profiles heavily. The stress concentration mainly locates at the contact interface between fuel meat and cladding, especially at the side of the interface.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129368534","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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