Chang Zhang, Ruixiang Wang, Hui Guo, H. Gu, Yao Xiao
The multiple-physics modeling has been demonstrated to be a high-fidelity and effective method for the analysis of reactor core physics and thermal-hydraulics. OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. In this work, OpenMC and CTF are coupled for the analysis of light water reactor fuel assembly. OpenMC provides axial and radial fuel pin normalized power distribution to CTF, and CTF gives the fuel temperatures and coolant properties to the neutronics simulation of OpenMC. The windowed multipole temperature method is used in OpenMC to match the accurate temperature distribution of fuel rods and coolants obtained from CTF. In this study, the OpenMC/CTF is applied and validated using the Virtual Environment for Reactor Applications (VERA) core physics benchmark problem 6, from the Consortium for Advanced Simulation of Light Water Reactors. The problem involves a three-dimensional fuel assembly in Hot Full Power conditions. The Keff eigenvalue, pin power distribution, fuel temperatures, and coolant properties are obtained and compared with VERA’s reference results (MPACT/CTF). Our converged results showed good consistency with the reference solution, eigenvalue differences agreed within 183 pcm, axially-integrated normalized radial fission distribution difference agreed within +1.4%/−1.9%, local volume-averaged fuel pin temperatures agreed within +26.3°C, and local subchannel exit coolant temperatures agreed within +3.6°C/−0.4°C. These preliminary solutions prove that OpenMC coupled to CTF method has shown high-fidelity results in the three-dimensional fuel assembly neutronics and thermal-hydraulics coupled problems.
{"title":"Coupled OpenMC/CTF to VERA Core Physics Benchmark Problem 6","authors":"Chang Zhang, Ruixiang Wang, Hui Guo, H. Gu, Yao Xiao","doi":"10.1115/icone29-92642","DOIUrl":"https://doi.org/10.1115/icone29-92642","url":null,"abstract":"\u0000 The multiple-physics modeling has been demonstrated to be a high-fidelity and effective method for the analysis of reactor core physics and thermal-hydraulics. OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. CTF is a subchannel thermal-hydraulics code designed for Light Water Reactor analysis. In this work, OpenMC and CTF are coupled for the analysis of light water reactor fuel assembly. OpenMC provides axial and radial fuel pin normalized power distribution to CTF, and CTF gives the fuel temperatures and coolant properties to the neutronics simulation of OpenMC. The windowed multipole temperature method is used in OpenMC to match the accurate temperature distribution of fuel rods and coolants obtained from CTF. In this study, the OpenMC/CTF is applied and validated using the Virtual Environment for Reactor Applications (VERA) core physics benchmark problem 6, from the Consortium for Advanced Simulation of Light Water Reactors. The problem involves a three-dimensional fuel assembly in Hot Full Power conditions. The Keff eigenvalue, pin power distribution, fuel temperatures, and coolant properties are obtained and compared with VERA’s reference results (MPACT/CTF). Our converged results showed good consistency with the reference solution, eigenvalue differences agreed within 183 pcm, axially-integrated normalized radial fission distribution difference agreed within +1.4%/−1.9%, local volume-averaged fuel pin temperatures agreed within +26.3°C, and local subchannel exit coolant temperatures agreed within +3.6°C/−0.4°C. These preliminary solutions prove that OpenMC coupled to CTF method has shown high-fidelity results in the three-dimensional fuel assembly neutronics and thermal-hydraulics coupled problems.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"20 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":"132082287","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}
Kang Li, Liangzhi Cao, Jianxin Miao, Haotong Zhang, Tao Dai
The fusion neutronics simulation has been a great challenge to the numerical calculation of the neutron-transport equation. The key is how to deal with the features of fusion devices, such as large-scale, complex geometric models, large vacuum regions, etc. NECP-FISH, a code developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University, is used to address this challenge. NECP-FISH adopts a deterministic numerical method instead of the Monte Carlo method because the deterministic numerical method is of higher computational efficiency and costs less computational time. To deal with large vacuum region, large-scale and complex geometric model, the first order neutron-transport equation is solved, the spherical harmonics function and the finite element method are applied to the expansion of angle and space. NECP-FISH has been validated by benchmark problems such as strong absorption problem, internal void problem, and Kobayashi series of problems. What’s more, NECP-FISH builds the user interface based on the platform SALOME so that users can visually build the necessary models for problems. NECP-FISH has been applied to the neutronics calculation of the breeder unit of Helium Cooling Ceramic Breeder (HCCB) and the blanket of CFETR. The numerical results demonstrate that the NECP-FISH code can efficiently solve the neutron transport problem of the fusion reactor.
{"title":"Neutronics Analysis of Fusion Blanket Based on the Spherical Harmonic Function and Finite Element Method","authors":"Kang Li, Liangzhi Cao, Jianxin Miao, Haotong Zhang, Tao Dai","doi":"10.1115/icone29-92622","DOIUrl":"https://doi.org/10.1115/icone29-92622","url":null,"abstract":"\u0000 The fusion neutronics simulation has been a great challenge to the numerical calculation of the neutron-transport equation. The key is how to deal with the features of fusion devices, such as large-scale, complex geometric models, large vacuum regions, etc. NECP-FISH, a code developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University, is used to address this challenge. NECP-FISH adopts a deterministic numerical method instead of the Monte Carlo method because the deterministic numerical method is of higher computational efficiency and costs less computational time. To deal with large vacuum region, large-scale and complex geometric model, the first order neutron-transport equation is solved, the spherical harmonics function and the finite element method are applied to the expansion of angle and space. NECP-FISH has been validated by benchmark problems such as strong absorption problem, internal void problem, and Kobayashi series of problems. What’s more, NECP-FISH builds the user interface based on the platform SALOME so that users can visually build the necessary models for problems. NECP-FISH has been applied to the neutronics calculation of the breeder unit of Helium Cooling Ceramic Breeder (HCCB) and the blanket of CFETR. The numerical results demonstrate that the NECP-FISH code can efficiently solve the neutron transport problem of the fusion reactor.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"15 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":"123829782","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 term “holdup” refers to nuclear material that has been deposited in the processing facility’s equipment, containers, pipelines, etc. Nuclear material control and accounting require the monitoring of nuclear material holdup in process. Holdup is useful for indicating the state of facility operations, critical radiation safety status, and so on. It is more difficulty to quantitative analyzed nuclear material holdup by destructive analysis method. So non-destructive analysis (NDA) technology for in-situ and on-line measurement satisfied to the actual situation are the most recommended nominator for holdup quantitative analysis. In order to get the detection efficiency of γ-ray spectroscopy for quantitative nuclear material holdup. The intrinsic detection efficiency, geometric detection efficiency, and self-absorption attenuation correction methods are researched in this paper by combining Monte Carlo simulation and experiment. Three-dimensional and the least square fitting methods, and the distribution of the nuclear material holdup in the simulation pipeline system of nuclear facilities’ production line is simplified. Resulted to the experiment, the deviation between declared and the measured value is less than 30%. Compared with the requirements of national nuclear safeguards supervision and the needs of facility operation, the detection efficiency fitting method research by γ-ray spectroscopy for quantitative nuclear material holdup is achievable. It also can be widely used to provide relevant data in radiation dose monitoring, decommissioning source item investigation, and so on.
{"title":"Detection Efficiency Fitting Method Research by Γ-Ray Spectroscopy for Quantitative Nuclear Material Holdup","authors":"Yurong Li, Sijia Wang, Jing Wang, Lixia He","doi":"10.1115/icone29-90210","DOIUrl":"https://doi.org/10.1115/icone29-90210","url":null,"abstract":"\u0000 The term “holdup” refers to nuclear material that has been deposited in the processing facility’s equipment, containers, pipelines, etc. Nuclear material control and accounting require the monitoring of nuclear material holdup in process. Holdup is useful for indicating the state of facility operations, critical radiation safety status, and so on. It is more difficulty to quantitative analyzed nuclear material holdup by destructive analysis method. So non-destructive analysis (NDA) technology for in-situ and on-line measurement satisfied to the actual situation are the most recommended nominator for holdup quantitative analysis. In order to get the detection efficiency of γ-ray spectroscopy for quantitative nuclear material holdup. The intrinsic detection efficiency, geometric detection efficiency, and self-absorption attenuation correction methods are researched in this paper by combining Monte Carlo simulation and experiment. Three-dimensional and the least square fitting methods, and the distribution of the nuclear material holdup in the simulation pipeline system of nuclear facilities’ production line is simplified. Resulted to the experiment, the deviation between declared and the measured value is less than 30%. Compared with the requirements of national nuclear safeguards supervision and the needs of facility operation, the detection efficiency fitting method research by γ-ray spectroscopy for quantitative nuclear material holdup is achievable. It also can be widely used to provide relevant data in radiation dose monitoring, decommissioning source item investigation, and so on.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"50 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":"127764638","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}
A Multiphysics approach to calculate the constituent redistribution in U-Pu-Zr metallic fuel is developed in this work and the fuel performance of U-Pu-Zr metallic fuel under normal operating condition is investigated by CAMPUS code where the model of constituent redistribution is applied. The metallic nuclear fuel is proposed to be used as a candidate of Sodium cooled Fast Neutron Reactor fuel due to its superior thermodynamic properties, such as thermal conductivity and heat capacity. Recently, the constituent redistribution is found to be an important phenomenon in the metallic fuel pellet, which influences the thermodynamic properties of metallic fuel greatly. In this work, the properties of metallic fuel and the calculation method of constituent redistribution are introduced firstly. And the model of constituent redistribution in U-Pu-Zr metallic fuel is developed and implemented into CAMPUS code, and then the metallic fuel performance in the sodium-cooled fast reactor is further studied by CAMPUS code under normal operating condition. Thirdly, the model of constituent redistribution is verified and the fuel performance of U-Pu-Zr metallic fuel is presented and discussed. The Zirconium element is found to migrate to the fuel centerline and fuel surface and the performance of the U-Pu-Zr fuel is found to be influenced by the constituent redistribution dramatically.
{"title":"A Steady-State Analysis of U-Pu-Zr Metallic Fuel Performance on Constituent Redistribution by Multiphysics Modeling","authors":"Rong Liu, S. Liu, H. Han, Liwen Yang","doi":"10.1115/icone29-92447","DOIUrl":"https://doi.org/10.1115/icone29-92447","url":null,"abstract":"\u0000 A Multiphysics approach to calculate the constituent redistribution in U-Pu-Zr metallic fuel is developed in this work and the fuel performance of U-Pu-Zr metallic fuel under normal operating condition is investigated by CAMPUS code where the model of constituent redistribution is applied. The metallic nuclear fuel is proposed to be used as a candidate of Sodium cooled Fast Neutron Reactor fuel due to its superior thermodynamic properties, such as thermal conductivity and heat capacity. Recently, the constituent redistribution is found to be an important phenomenon in the metallic fuel pellet, which influences the thermodynamic properties of metallic fuel greatly. In this work, the properties of metallic fuel and the calculation method of constituent redistribution are introduced firstly. And the model of constituent redistribution in U-Pu-Zr metallic fuel is developed and implemented into CAMPUS code, and then the metallic fuel performance in the sodium-cooled fast reactor is further studied by CAMPUS code under normal operating condition. Thirdly, the model of constituent redistribution is verified and the fuel performance of U-Pu-Zr metallic fuel is presented and discussed. The Zirconium element is found to migrate to the fuel centerline and fuel surface and the performance of the U-Pu-Zr fuel is found to be influenced by the constituent redistribution dramatically.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"62 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":"132847147","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}
Zhenhe Li, Xi Huang, Zixiong Zhan, Qi Zhao, Junxiong Liu, Lihua Wei, Xiaoyan Li
It is well known that the type and concentration of impurity ions in pressurized water reactor environments will significantly alter the corrosion behavior of materials. However, the effect of these water chemistry parameters on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. The type and concentration of impurity ions on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. In this study, the effect of Cr6+ concentrations on the corrosion behavior of 316LN stainless steel in 550 °C/25 MPa supercritical water (SCW) was investigated. The results show that the weight gain first decreased and subsequently increased with the increasing of Cr6+ concentration, which is 0.40±0.01 mg/cm2 for 0 ppm, 0.28±0.03 mg/cm2 for 0.1 ppm, 0.55±0.02 mg/cm2 for 1.0 ppm, and 0.59±0.01 mg/cm2 for 50 ppm, respectively. The Fe-rich magnetite oxides were gradually covered on the surface of samples, and its proportion on the surface followed a similar alteration trend as the weight gain, which is 44.3% for 0 ppm, 37.8% for 0.1 ppm, 67.7% for 1 ppm, and 87.6% for 50 ppm. A duplex structure of the oxide films, including an Fe-rich outer magnetite layer and a Cr-rich inner spinel layer, was developed on the regions covered with Fe-rich magnetite oxides. The alternation in the thickness of oxide film followed a similar trend as the weight gain, which is 7.67 μm for 0 ppm, 5.80 μm for 0.1 ppm, 9.97 μm for 1 ppm, and 11.23 μm for 50 ppm. However, a Cr-rich spinel oxide film was formed on the region without Ferich magnetite oxides. The thickness of this oxide film was unchanged regardless of Cr6+ ion concentration, which is about 2 μm, but the concentration of Cr in this oxide film is significantly higher than reference sample. These results indicate that the increasing of Cr6+ concentration in SCW will result in the corrosion rate to decrease first and then increase, which means that some measurements should be conducted during the operation of SCWR to keep a lower concentration of Cr6+.
{"title":"Effect of Cr6+ Concentrations on the Corrosion Behavior of 316LN Stainless Steel in Supercritical Water","authors":"Zhenhe Li, Xi Huang, Zixiong Zhan, Qi Zhao, Junxiong Liu, Lihua Wei, Xiaoyan Li","doi":"10.1115/icone29-89235","DOIUrl":"https://doi.org/10.1115/icone29-89235","url":null,"abstract":"\u0000 It is well known that the type and concentration of impurity ions in pressurized water reactor environments will significantly alter the corrosion behavior of materials. However, the effect of these water chemistry parameters on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. The type and concentration of impurity ions on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. In this study, the effect of Cr6+ concentrations on the corrosion behavior of 316LN stainless steel in 550 °C/25 MPa supercritical water (SCW) was investigated. The results show that the weight gain first decreased and subsequently increased with the increasing of Cr6+ concentration, which is 0.40±0.01 mg/cm2 for 0 ppm, 0.28±0.03 mg/cm2 for 0.1 ppm, 0.55±0.02 mg/cm2 for 1.0 ppm, and 0.59±0.01 mg/cm2 for 50 ppm, respectively. The Fe-rich magnetite oxides were gradually covered on the surface of samples, and its proportion on the surface followed a similar alteration trend as the weight gain, which is 44.3% for 0 ppm, 37.8% for 0.1 ppm, 67.7% for 1 ppm, and 87.6% for 50 ppm. A duplex structure of the oxide films, including an Fe-rich outer magnetite layer and a Cr-rich inner spinel layer, was developed on the regions covered with Fe-rich magnetite oxides. The alternation in the thickness of oxide film followed a similar trend as the weight gain, which is 7.67 μm for 0 ppm, 5.80 μm for 0.1 ppm, 9.97 μm for 1 ppm, and 11.23 μm for 50 ppm. However, a Cr-rich spinel oxide film was formed on the region without Ferich magnetite oxides. The thickness of this oxide film was unchanged regardless of Cr6+ ion concentration, which is about 2 μm, but the concentration of Cr in this oxide film is significantly higher than reference sample. These results indicate that the increasing of Cr6+ concentration in SCW will result in the corrosion rate to decrease first and then increase, which means that some measurements should be conducted during the operation of SCWR to keep a lower concentration of Cr6+.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"88 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":"133332432","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}
Source localization of accidental radionuclide release is crucial to nuclear emergency and decision making, especially when no reliable information about possible release regions can be previously obtained. To rapidly and accurately determine the source location with a limited number of measurements, this paper proposes a two-step data assimilation method based on forward and backward dispersion simulations in the RIMPUFF model. In the first step, meteorological variables such as wind speed are reversed and the measurements are set as sources for several backward RIMPUFF runs. By identifying the overlap of backward plumes originating from different measurement sites, potential regions discretized into grids can be roughly determined. In the second step, the grids selected by the first step are set as sources for forward RIMPUFF runs with unit release rate and the source location will be refined by minimizing a correlation-based function of simulated and measured data. The method is verified by SCK-CEN 41Ar field experiment. Using gamma dose rate data from all measurement sites, source location is retrieved with errors of 22.36m and 30m respectively on the two days of the field experiment. Compared with the direct correlation-based method, the proposed method achieves satisfactory solutions in an obviously shorter time. In particular, there is no need for predetermining the source release rate throughout the process, indicating that the method can conveniently combine with other source term inversion approaches whose source location is viewed as a known quantity. Therefore, the presented two-step data assimilation method here is potentially a useful tool with high accuracy and efficiency for further integration in nuclear emergency response system.
{"title":"Source Localization of Accidental Radionuclide Release Using a Data Assimilation Method Based on Forward and Backward Dispersion Simulations","authors":"Yuhan Xu, Shengjiang Fang, Xinwen Dong, Shuhan Zhuang","doi":"10.1115/icone29-91645","DOIUrl":"https://doi.org/10.1115/icone29-91645","url":null,"abstract":"\u0000 Source localization of accidental radionuclide release is crucial to nuclear emergency and decision making, especially when no reliable information about possible release regions can be previously obtained. To rapidly and accurately determine the source location with a limited number of measurements, this paper proposes a two-step data assimilation method based on forward and backward dispersion simulations in the RIMPUFF model. In the first step, meteorological variables such as wind speed are reversed and the measurements are set as sources for several backward RIMPUFF runs. By identifying the overlap of backward plumes originating from different measurement sites, potential regions discretized into grids can be roughly determined. In the second step, the grids selected by the first step are set as sources for forward RIMPUFF runs with unit release rate and the source location will be refined by minimizing a correlation-based function of simulated and measured data. The method is verified by SCK-CEN 41Ar field experiment. Using gamma dose rate data from all measurement sites, source location is retrieved with errors of 22.36m and 30m respectively on the two days of the field experiment. Compared with the direct correlation-based method, the proposed method achieves satisfactory solutions in an obviously shorter time. In particular, there is no need for predetermining the source release rate throughout the process, indicating that the method can conveniently combine with other source term inversion approaches whose source location is viewed as a known quantity. Therefore, the presented two-step data assimilation method here is potentially a useful tool with high accuracy and efficiency for further integration in nuclear emergency response system.","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":"133744471","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}
Hang Ni, Xinhe Qu, Gangyong Zhao, Ping Zhang, Wei Peng
Hydrogen is an important clean alternative energy resource for the future, and nuclear hydrogen production can efficiently produce carbon-free hydrogen on a large scale. In this study, a hydrogen and electricity co-production system coupling iodine-sulfur cycle with a very-high-temperature gas-cooled reactor is proposed. The helium on the secondary side of the intermediate heat exchanger provides high-grade heat for the sulfuric acid decomposition reactor and hydroiodic acid decomposition reactor, and the steam extracted from the power generation circuit provides low-grade heat for other hydrogen production components. A supercritical steam generator is used and a reheating section is designed to improve power generation efficiency. The energy analysis reveals that as the hydrogen production rate increases, the power generation efficiency decreases, whereas the overall hydrogen and electricity efficiency increases. The power generation efficiency and overall hydrogen and electricity efficiency of the system are 38.2 % and 45.5 % at a hydrogen production rate of 161.73 mol/s. The parametric study shows that the power generation efficiency and overall hydrogen and electricity efficiency of the system increase with an increase in the main steam temperature or main steam pressure, and decrease with an increase in the reheated steam pressure. Among the three parameters, the main steam temperature markedly affects the performance of the system, followed by the reheated steam pressure and main steam pressure.
{"title":"Energy Analysis and Parametric Study of Hydrogen and Electricity Co-Production System Coupled With a Very-High-Temperature Gas-Cooled Reactor","authors":"Hang Ni, Xinhe Qu, Gangyong Zhao, Ping Zhang, Wei Peng","doi":"10.1115/icone29-91900","DOIUrl":"https://doi.org/10.1115/icone29-91900","url":null,"abstract":"\u0000 Hydrogen is an important clean alternative energy resource for the future, and nuclear hydrogen production can efficiently produce carbon-free hydrogen on a large scale. In this study, a hydrogen and electricity co-production system coupling iodine-sulfur cycle with a very-high-temperature gas-cooled reactor is proposed. The helium on the secondary side of the intermediate heat exchanger provides high-grade heat for the sulfuric acid decomposition reactor and hydroiodic acid decomposition reactor, and the steam extracted from the power generation circuit provides low-grade heat for other hydrogen production components. A supercritical steam generator is used and a reheating section is designed to improve power generation efficiency. The energy analysis reveals that as the hydrogen production rate increases, the power generation efficiency decreases, whereas the overall hydrogen and electricity efficiency increases. The power generation efficiency and overall hydrogen and electricity efficiency of the system are 38.2 % and 45.5 % at a hydrogen production rate of 161.73 mol/s. The parametric study shows that the power generation efficiency and overall hydrogen and electricity efficiency of the system increase with an increase in the main steam temperature or main steam pressure, and decrease with an increase in the reheated steam pressure. Among the three parameters, the main steam temperature markedly affects the performance of the system, followed by the reheated steam pressure and main steam pressure.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"21 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":"125191765","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 heat pipe is the essential component in the heat pipe cooled reactor, it conducts heat through the phase change of the working medium and the circulating flow of gas and liquid. One end of the heat pipe is inserted into the reactor as the evaporation section, while the other end serves as the condensation section connecting to the thermoelectric conversion system. Its heat transfer capacity and temperature distribution under a steady state affect the safety, core temperature and energy conversion efficiency of the heat pipe cooled reactor system. Predicting the heat transfer process and steady-state temperature distribution of the heat pipe is helpful to learn about the heat transfer mechanism inside the heat pipe and key factors affecting its performance. It is also helpful to predict the operating characteristics of the heat pipe cooled reactor system and improve the performance of the heat pipe. The heat pipe is generally considered to be a heat transfer element with good isothermal properties. Past studies, however, proved the significant quantity of its phase-change thermal resistance, resulting in a temperature difference at the gas-liquid interface of the heat pipe that affects the evaporation and condensation. Studies on the relationship between the temperature difference, the working temperature, and heat transfer power of the heat pipe are conductive to building a heat pipe model, to predict the temperature distribution of the heat pipe in a steady state in a much more accurate manner. This paper offered a heat pipe model that applies the temperature difference of the phase change interface to calculate the evaporation and condensation. Meanwhile, a 1-meter-long sodium heat pipe was adopted to carry out steady-state heat transfer experiments. Measured and computed the temperature difference of the phase change interface of the heat pipe under varying power and temperature and conducted the relational equation as the input of the heat pipe model. The model predicted the steady-state temperature distribution of the sodium heat pipe and the transient when the power changes after the steam reach the continuous flow stage and compared the calculated outcomes with the experimental values. The transient error was less than 20K, and the steady-state error was less than 12K. The results show that the temperature difference at the phase change interface has a great influence on the steady-state temperature distribution of the heat pipe, which changes with the operating temperature. However, the physical and geometric factors affecting the temperature difference need to be further studied, to reduce the temperature difference as much as possible and improve the energy conversion efficiency of the heat pipe reactor system in the future.
{"title":"A Transient Model of Sodium Heat Pipe With Phase Change Temperature Difference","authors":"Ruicheng Zhong, Yugao Ma, Qingzhu Zhao, Jian Deng, Yu Liu, Shuhua Ding","doi":"10.1115/icone29-92202","DOIUrl":"https://doi.org/10.1115/icone29-92202","url":null,"abstract":"\u0000 The heat pipe is the essential component in the heat pipe cooled reactor, it conducts heat through the phase change of the working medium and the circulating flow of gas and liquid. One end of the heat pipe is inserted into the reactor as the evaporation section, while the other end serves as the condensation section connecting to the thermoelectric conversion system. Its heat transfer capacity and temperature distribution under a steady state affect the safety, core temperature and energy conversion efficiency of the heat pipe cooled reactor system. Predicting the heat transfer process and steady-state temperature distribution of the heat pipe is helpful to learn about the heat transfer mechanism inside the heat pipe and key factors affecting its performance. It is also helpful to predict the operating characteristics of the heat pipe cooled reactor system and improve the performance of the heat pipe. The heat pipe is generally considered to be a heat transfer element with good isothermal properties. Past studies, however, proved the significant quantity of its phase-change thermal resistance, resulting in a temperature difference at the gas-liquid interface of the heat pipe that affects the evaporation and condensation. Studies on the relationship between the temperature difference, the working temperature, and heat transfer power of the heat pipe are conductive to building a heat pipe model, to predict the temperature distribution of the heat pipe in a steady state in a much more accurate manner. This paper offered a heat pipe model that applies the temperature difference of the phase change interface to calculate the evaporation and condensation. Meanwhile, a 1-meter-long sodium heat pipe was adopted to carry out steady-state heat transfer experiments. Measured and computed the temperature difference of the phase change interface of the heat pipe under varying power and temperature and conducted the relational equation as the input of the heat pipe model. The model predicted the steady-state temperature distribution of the sodium heat pipe and the transient when the power changes after the steam reach the continuous flow stage and compared the calculated outcomes with the experimental values. The transient error was less than 20K, and the steady-state error was less than 12K. The results show that the temperature difference at the phase change interface has a great influence on the steady-state temperature distribution of the heat pipe, which changes with the operating temperature. However, the physical and geometric factors affecting the temperature difference need to be further studied, to reduce the temperature difference as much as possible and improve the energy conversion efficiency of the heat pipe reactor system in the future.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"5 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":"133712724","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}
Yao Zhou, Liangzhi Cao, Changyou Zhao, Zhifeng Li, Yuxiang Zhu
Self-Powered Neutron Detector (SPND) has been widely applied for the on-line monitoring of the key core parameters, including the power distribution, power-peak factor and so on. For the SPND layout design, the vital parameters include the SPND amount and loading position in the radial direction and segment amount in the axial direction. Different SPND layout reflects different ability of power distribution reconstruction, which is directly related to the safety and economy of nuclear power plant. For HPR1000, the neutronics sensitivity of SPND layout design is analyzed based on a 3D on-line power distribution monitoring system NECP-ONION. Through the comparison and analysis of the reconstruction errors of power distribution, the design rationality of SPND is evaluated and the optimization design is recommended. For the recommended SPND layout design, the loading positions in the radial direction are arranged at intervals and segment amount in the axial direction is 15. The recommended SPND layout design shows stable performance in the whole life for HPR1000 reactor, and the RMS errors of the power distribution reconstruction has been in the range of [0.32%, 0.48%], meeting the safety limit and accuracy requirements of power distribution reconstruction. This work verifies the on-line monitoring ability of NECP-ONION system. Additionally, it provides a technical reference for the SPND experiment in HPR1000.
{"title":"Layout Optimization of Self-Powered Neutron Detector for HPR1000","authors":"Yao Zhou, Liangzhi Cao, Changyou Zhao, Zhifeng Li, Yuxiang Zhu","doi":"10.1115/icone29-91784","DOIUrl":"https://doi.org/10.1115/icone29-91784","url":null,"abstract":"\u0000 Self-Powered Neutron Detector (SPND) has been widely applied for the on-line monitoring of the key core parameters, including the power distribution, power-peak factor and so on. For the SPND layout design, the vital parameters include the SPND amount and loading position in the radial direction and segment amount in the axial direction. Different SPND layout reflects different ability of power distribution reconstruction, which is directly related to the safety and economy of nuclear power plant. For HPR1000, the neutronics sensitivity of SPND layout design is analyzed based on a 3D on-line power distribution monitoring system NECP-ONION. Through the comparison and analysis of the reconstruction errors of power distribution, the design rationality of SPND is evaluated and the optimization design is recommended. For the recommended SPND layout design, the loading positions in the radial direction are arranged at intervals and segment amount in the axial direction is 15. The recommended SPND layout design shows stable performance in the whole life for HPR1000 reactor, and the RMS errors of the power distribution reconstruction has been in the range of [0.32%, 0.48%], meeting the safety limit and accuracy requirements of power distribution reconstruction. This work verifies the on-line monitoring ability of NECP-ONION system. Additionally, it provides a technical reference for the SPND experiment in HPR1000.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"41 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":"123503421","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}
Betavoltaics are direct conversion energy devices that are ideal for low, micropower and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radioisotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radioiso-tope. The source density can be increased by the selection of solid-state substrates. While solid-state substrates can be selected from simulations, the viability of the substrate to absorb tritium has to evaluated. The development of a hydrogen loading system was performed to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert method, where the initial pressure and volume is constant and pressure drop in the system is used to determine hydrogen uptake of a film substrate. The procedures of the hydrogen loading system are detailed. To test the procedures of the hydrogen loading system, old, palladium films were loaded. Results show uptake of hydrogen by the thin palladium films, as well as cycles of hydrogen absorption and desorption. Hydrogen loading of palladium was compared to a prior result and was shown to have similar results.
{"title":"Hydrogen Loading System for Thin Films for Betavoltaics","authors":"D. Cheu, T. Adams, S. Revankar","doi":"10.1115/icone29-93910","DOIUrl":"https://doi.org/10.1115/icone29-93910","url":null,"abstract":"\u0000 Betavoltaics are direct conversion energy devices that are ideal for low, micropower and long-lasting, uninterruptable applications. Betavoltaics operate similarly to photovoltaics where a radioisotope irradiates beta particles into a semiconductor p-n junction that converts the kinetic energy into electrical energy. Betavoltaics are limited by their power output from the radioiso-tope. The source density can be increased by the selection of solid-state substrates. While solid-state substrates can be selected from simulations, the viability of the substrate to absorb tritium has to evaluated. The development of a hydrogen loading system was performed to evaluate different film types to understand how they perform during the hydrogen/tritium loading process. The hydrogen loading system utilizes the Sievert method, where the initial pressure and volume is constant and pressure drop in the system is used to determine hydrogen uptake of a film substrate. The procedures of the hydrogen loading system are detailed. To test the procedures of the hydrogen loading system, old, palladium films were loaded. Results show uptake of hydrogen by the thin palladium films, as well as cycles of hydrogen absorption and desorption. Hydrogen loading of palladium was compared to a prior result and was shown to have similar results.","PeriodicalId":302303,"journal":{"name":"Volume 15: Student Paper Competition","volume":"31 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":"129905261","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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