Shun Zhang, H. Xue, Yubiao Zhang, K. Zhao, Jiaqing Zhang
The micro-mechanical state at the crack front is one of the key factors affecting the stress corrosion cracking (SCC) growth behavior. The mechanical heterogeneity and residual stress in the dissimilar metal welded joint (DMWJ) induce the micro-mechanical state at the crack front to become more complex. The sandwich model and dual-field model of the DMWJ with inner surface axial crack were established in this study. The stress and strain states at the crack front with different crack locations and lengths under the interaction of the mechanical property and the residual stress were investigated. The results show that a more accurate evaluation of stress and strain states can be obtained when using the dual-field model to describe the material mechanical property and residual stress of the DMWJ. The sandwich model overestimates the crack driving force including the stress and strain at the crack front. The tensile stress in the middle of shallow cracks is smaller than that at both ends, while the tensile stress in the middle of deep crack is larger than that at both ends. The variation trend of the tensile stress and normal strain at the crack apex is basically the same as that of the residual stress with the crack depth. However, there is almost no normal plastic strain in the initial stage of crack propagation due to the small residual stress in the initial stage.
{"title":"Stress and Strain State Analysis of Crack Front in Dissimilar Metal Welded Joints with Dual Field of Mechanical Heterogeneity and Residual Stress","authors":"Shun Zhang, H. Xue, Yubiao Zhang, K. Zhao, Jiaqing Zhang","doi":"10.1155/2022/2858320","DOIUrl":"https://doi.org/10.1155/2022/2858320","url":null,"abstract":"The micro-mechanical state at the crack front is one of the key factors affecting the stress corrosion cracking (SCC) growth behavior. The mechanical heterogeneity and residual stress in the dissimilar metal welded joint (DMWJ) induce the micro-mechanical state at the crack front to become more complex. The sandwich model and dual-field model of the DMWJ with inner surface axial crack were established in this study. The stress and strain states at the crack front with different crack locations and lengths under the interaction of the mechanical property and the residual stress were investigated. The results show that a more accurate evaluation of stress and strain states can be obtained when using the dual-field model to describe the material mechanical property and residual stress of the DMWJ. The sandwich model overestimates the crack driving force including the stress and strain at the crack front. The tensile stress in the middle of shallow cracks is smaller than that at both ends, while the tensile stress in the middle of deep crack is larger than that at both ends. The variation trend of the tensile stress and normal strain at the crack apex is basically the same as that of the residual stress with the crack depth. However, there is almost no normal plastic strain in the initial stage of crack propagation due to the small residual stress in the initial stage.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45690810","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The gravity cooling water tank is a remarkable structural feature of third-generation pressurized water reactor nuclear power plant. To investigate the influence of fluid-structure interaction (FSI) on the seismic response of the structure, this study designed two 1 : 50 simplified models of the AP1000 shield building. A series of shaking table tests were conducted to study the seismic responses with and without FSI effect. The natural frequency, acceleration, strain, and hydrodynamic pressure of the two models were analyzed, and the seismic reduction effect of the water tank was evaluated. Moreover, the test data were compared with the results of numerical analysis using the ABAQUS software. The results show that the presence of water and the sloshing of water reduce the natural frequency and seismic response of the model structure. Thus, the gravity cooling water tank has a certain seismic reduction effect. The simplified model of water sloshing can be used to analyze the seismic response of the shield building.
{"title":"Experimental and Numerical Studies of AP1000 Shield Building considering Fluid-Structure Interaction","authors":"Zhi Zhang, Chenning Song, Zhining Duan, Zhi-yuan Cheng","doi":"10.1155/2022/6458549","DOIUrl":"https://doi.org/10.1155/2022/6458549","url":null,"abstract":"The gravity cooling water tank is a remarkable structural feature of third-generation pressurized water reactor nuclear power plant. To investigate the influence of fluid-structure interaction (FSI) on the seismic response of the structure, this study designed two 1 : 50 simplified models of the AP1000 shield building. A series of shaking table tests were conducted to study the seismic responses with and without FSI effect. The natural frequency, acceleration, strain, and hydrodynamic pressure of the two models were analyzed, and the seismic reduction effect of the water tank was evaluated. Moreover, the test data were compared with the results of numerical analysis using the ABAQUS software. The results show that the presence of water and the sloshing of water reduce the natural frequency and seismic response of the model structure. Thus, the gravity cooling water tank has a certain seismic reduction effect. The simplified model of water sloshing can be used to analyze the seismic response of the shield building.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44146782","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In this study, we evaluate hydrogen production costs using small modular reactors (SMRs). Furthermore, we employ a machine learning-based approach to predict important parameters that affect the hydrogen production cost. Additionally, we use a hydrogen economic evaluation program to calculate the hydrogen production cost when using the two types of SMRs: system-integrated modular advanced reactor (SMART) developed by the Korea Atomic Energy Research Institute (KAERI) and NuScale power module™ (NPM) developed by the NuScale Power, LLC. Different storage and transportation means were selected to find the cheapest option. Using SMART, storing hydrogen in compressed gas and transporting it through pipes (CG-Pipe) is the best option, with an estimated cost of USD 2.77/kg. Other options when using SMART include storing in compressed gas and transporting with a vehicle (CG-Vehicle), with an estimated cost of USD 3.27/kg; storing by liquefaction and transporting with a vehicle (L-Vehicle), with an estimated cost of USD 3.31/kg; and storing in metal hydrides and transporting with a vehicle (MH-Vehicle), with an estimated cost of USD 6.97/kg. Using NPM, CG-Pipe is the cheapest option to generate hydrogen, with an estimated cost of USD 2.95/kg. Other options include CG-Vehicle (USD 3.35/kg), L-Vehicle (USD 3.42/kg), and MH-Vehicle (USD 7.04/kg). Hydrogen production using SMART is cheaper than using NPM. However, the observed difference between the hydrogen production costs using the two reactors was insignificant. We conclude that the optimal hydrogen production cost ranges from USD 3.27/kg (CG-Vehicle) to USD 3.42 (L-Vehicle). This conclusion is because the common hydrogen transportation means is with a vehicle. From a machine learning-based approach, we determine the important parameters that affect hydrogen production costs. The most important parameter is the heat consumption (MWth/unit) at hydrogen generation plants, and other parameters include electricity rating and heat for hydrogen generation plants.
{"title":"Machine Learning-Based Approach for Hydrogen Economic Evaluation of Small Modular Reactors","authors":"Juyoul Kim, M. Rweyemamu, B. Purevsuren","doi":"10.1155/2022/9297122","DOIUrl":"https://doi.org/10.1155/2022/9297122","url":null,"abstract":"In this study, we evaluate hydrogen production costs using small modular reactors (SMRs). Furthermore, we employ a machine learning-based approach to predict important parameters that affect the hydrogen production cost. Additionally, we use a hydrogen economic evaluation program to calculate the hydrogen production cost when using the two types of SMRs: system-integrated modular advanced reactor (SMART) developed by the Korea Atomic Energy Research Institute (KAERI) and NuScale power module™ (NPM) developed by the NuScale Power, LLC. Different storage and transportation means were selected to find the cheapest option. Using SMART, storing hydrogen in compressed gas and transporting it through pipes (CG-Pipe) is the best option, with an estimated cost of USD 2.77/kg. Other options when using SMART include storing in compressed gas and transporting with a vehicle (CG-Vehicle), with an estimated cost of USD 3.27/kg; storing by liquefaction and transporting with a vehicle (L-Vehicle), with an estimated cost of USD 3.31/kg; and storing in metal hydrides and transporting with a vehicle (MH-Vehicle), with an estimated cost of USD 6.97/kg. Using NPM, CG-Pipe is the cheapest option to generate hydrogen, with an estimated cost of USD 2.95/kg. Other options include CG-Vehicle (USD 3.35/kg), L-Vehicle (USD 3.42/kg), and MH-Vehicle (USD 7.04/kg). Hydrogen production using SMART is cheaper than using NPM. However, the observed difference between the hydrogen production costs using the two reactors was insignificant. We conclude that the optimal hydrogen production cost ranges from USD 3.27/kg (CG-Vehicle) to USD 3.42 (L-Vehicle). This conclusion is because the common hydrogen transportation means is with a vehicle. From a machine learning-based approach, we determine the important parameters that affect hydrogen production costs. The most important parameter is the heat consumption (MWth/unit) at hydrogen generation plants, and other parameters include electricity rating and heat for hydrogen generation plants.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43378412","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Intense fluid-dynamic interaction at the impeller outlet strongly affects the unsteady flow and pressure stability within the centrifugal pump. In order to have a better understanding of the pressure fluctuation of centrifugal pumps, a numerical calculation is carried out by using the RNG k-epsilon turbulence model under various flow rates. The numerical calculation results are compared with the experimental results in order to verify the reliability of the calculation model. The amplitude and frequency distribution of pressure fluctuation at the impeller outlet is obtained and analyzed in the time and frequency domain. The research results show that the blade passing frequency is the dominant frequency of the pressure fluctuation. And the pressure fluctuation is a periodic fluctuation. As the flow rate decreases, the periodicity of the pressure fluctuation decreases. Besides, the amplitude and intensity of pressure fluctuation are closely related to flow rate and spatial location. At the low flow rate, the amplitude of pressure fluctuation in the time domain and frequency domain is enlarged greatly, especially near the tongue region. The pressure difference distribution on both sides of the blade surface is extremely uneven, and the pressure changes significantly.
{"title":"Numerical Study of Unsteady Pressure Fluctuation at Impeller Outlet of a Centrifugal Pump","authors":"Xiaojie Ma, Lulu Zheng, Jinglei Qu, Mengmeng Wang","doi":"10.1155/2022/1758382","DOIUrl":"https://doi.org/10.1155/2022/1758382","url":null,"abstract":"Intense fluid-dynamic interaction at the impeller outlet strongly affects the unsteady flow and pressure stability within the centrifugal pump. In order to have a better understanding of the pressure fluctuation of centrifugal pumps, a numerical calculation is carried out by using the RNG k-epsilon turbulence model under various flow rates. The numerical calculation results are compared with the experimental results in order to verify the reliability of the calculation model. The amplitude and frequency distribution of pressure fluctuation at the impeller outlet is obtained and analyzed in the time and frequency domain. The research results show that the blade passing frequency is the dominant frequency of the pressure fluctuation. And the pressure fluctuation is a periodic fluctuation. As the flow rate decreases, the periodicity of the pressure fluctuation decreases. Besides, the amplitude and intensity of pressure fluctuation are closely related to flow rate and spatial location. At the low flow rate, the amplitude of pressure fluctuation in the time domain and frequency domain is enlarged greatly, especially near the tongue region. The pressure difference distribution on both sides of the blade surface is extremely uneven, and the pressure changes significantly.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45022921","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
As one of the rotating components in the reluctance motor type control rod drive mechanism (CRDM), the life of the scroll wheel is closely related to the service life of the CRDM. In addition, the prediction of the remaining useful life of the scroll wheel helps to optimize the maintenance process of the CRDM. This paper proposes a hybrid method to predict its remaining useful life when the available degradation data are rare and the failure threshold cannot be accurately defined. First, the particle filtering algorithm, whose state transfer equation is established on the segmental damage physical model, is used to predict the degradation state of the scroll wheel. Second, the proportional hazard model for the relationship between the scroll wheel damage characteristics and reliability model is established to predict the remaining useful life of it. The proposed method focuses on the establishment of segmental damage physical model and the clustering analysis of damage characteristics extracted from vibration signals, which can be used to predict the remaining useful life of the scroll wheel. In addition, the results provide an opportunity for the condition-based preventive maintenance of the CRDM.
{"title":"A Hybrid Method to Predict the Remaining Useful Life of Scroll Wheel of Control Rod Drive Mechanism","authors":"K. Zhu, Xinwen Zhao, Liming Zhang, Hang Yu","doi":"10.1155/2022/2383789","DOIUrl":"https://doi.org/10.1155/2022/2383789","url":null,"abstract":"As one of the rotating components in the reluctance motor type control rod drive mechanism (CRDM), the life of the scroll wheel is closely related to the service life of the CRDM. In addition, the prediction of the remaining useful life of the scroll wheel helps to optimize the maintenance process of the CRDM. This paper proposes a hybrid method to predict its remaining useful life when the available degradation data are rare and the failure threshold cannot be accurately defined. First, the particle filtering algorithm, whose state transfer equation is established on the segmental damage physical model, is used to predict the degradation state of the scroll wheel. Second, the proportional hazard model for the relationship between the scroll wheel damage characteristics and reliability model is established to predict the remaining useful life of it. The proposed method focuses on the establishment of segmental damage physical model and the clustering analysis of damage characteristics extracted from vibration signals, which can be used to predict the remaining useful life of the scroll wheel. In addition, the results provide an opportunity for the condition-based preventive maintenance of the CRDM.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44668159","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
D. Choi, W. Park, S. Son, Kukhee Lim, Yongjin Cho, Choengryul Choi
The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.
{"title":"Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident","authors":"D. Choi, W. Park, S. Son, Kukhee Lim, Yongjin Cho, Choengryul Choi","doi":"10.1155/2022/4531040","DOIUrl":"https://doi.org/10.1155/2022/4531040","url":null,"abstract":"The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47909214","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Evaluations of the Core Disruptive Accident (CDA) are significantly important for safety analysis of Sodium-cooled Fast Reactor (SFR) despite the very low probability of occurrence for CDA. During the material-relocation phase in CDA of SFR, the molten materials are possibly released from the core region into subcooled sodium, subsequently forming the debris bed on the lower part of the reactor vessel after being quenched and fragmented. The accumulated high-temperature debris with decay heat can cause sodium coolant boiling, leading to the so-called “debris bed self-leveling behavior” during which the shape of the debris bed becomes flattered (leveling). It is important to investigate the debris bed self-leveling behavior due to its potential capacity to induce the transfer of debris and affect the ability of cooling and criticality of the debris bed. Thus, in recent years, valuable knowledge concerning the mechanism and characteristics of this behavior was accumulated through lots of experimental results and modeling developments. Aimed at providing a valuable guideline for future investigations on this issue, in this study, the past experimental and modeling investigations on debris bed self-leveling mechanism and characteristics are systematically summarized and reviewed, and some future remarks are also proposed to promote the progression of further research for SFR severe accident analysis.
{"title":"Debris Bed Self-Leveling Mechanism and Characteristics for Core Disruptive Accident of Sodium-Cooled Fast Reactor: Review of Experimental and Modeling Investigations","authors":"Ruicong Xu, Songbai Cheng","doi":"10.1155/2022/2755471","DOIUrl":"https://doi.org/10.1155/2022/2755471","url":null,"abstract":"Evaluations of the Core Disruptive Accident (CDA) are significantly important for safety analysis of Sodium-cooled Fast Reactor (SFR) despite the very low probability of occurrence for CDA. During the material-relocation phase in CDA of SFR, the molten materials are possibly released from the core region into subcooled sodium, subsequently forming the debris bed on the lower part of the reactor vessel after being quenched and fragmented. The accumulated high-temperature debris with decay heat can cause sodium coolant boiling, leading to the so-called “debris bed self-leveling behavior” during which the shape of the debris bed becomes flattered (leveling). It is important to investigate the debris bed self-leveling behavior due to its potential capacity to induce the transfer of debris and affect the ability of cooling and criticality of the debris bed. Thus, in recent years, valuable knowledge concerning the mechanism and characteristics of this behavior was accumulated through lots of experimental results and modeling developments. Aimed at providing a valuable guideline for future investigations on this issue, in this study, the past experimental and modeling investigations on debris bed self-leveling mechanism and characteristics are systematically summarized and reviewed, and some future remarks are also proposed to promote the progression of further research for SFR severe accident analysis.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44336445","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity and shadow corrosion was examined under two different cathodes. This work also investigates the effect of current exchange density and the cathodic Tafel coefficient on the cathodic current. Using COMSOL Multiphysics 5.2, the Laplace equation was solved to obtain the electrostatic potential and current density distributions in the studied domain. When the distance d between the anode (Zry-2) and cathode (platinum/nickel) is ≤0.5 mm, it was observed that a uniform oxide layer of thickness 20 µm grew on the smooth internal surface of Zry-2 for corrosion lasting 1166 h. When d > 0.5 mm, the oxide thickness falls in a manner dictated by the degree of dissociation α of the electrolyte. At a cladding gap of 0.08 mm, a radiation-enhanced uniform corrosion rate of 2.405 10−1 mmpy was obtained for Zry-2. This value is 142 times greater than that obtained at room temperature in the absence of radiation. It was also observed that the corrosion rate falls at higher cladding gaps, and the rate of change depends on the degree of dissociation. Other phenomena such as the dynamics of shadow corrosion under varying electrode separation and electrolyte conductivities, as well as extensive evaluation of critical fuel cladding parameters, are presented in this work.
{"title":"Investigation of Oxidation and Counter-Oxidation in a One-Quarter Circular Geometry due to Shadow Corrosion","authors":"Doctor Enivweru, Qingyu Wang, A. Ayodeji, Yu Zhou","doi":"10.1155/2022/3676334","DOIUrl":"https://doi.org/10.1155/2022/3676334","url":null,"abstract":"To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity and shadow corrosion was examined under two different cathodes. This work also investigates the effect of current exchange density and the cathodic Tafel coefficient on the cathodic current. Using COMSOL Multiphysics 5.2, the Laplace equation was solved to obtain the electrostatic potential and current density distributions in the studied domain. When the distance d between the anode (Zry-2) and cathode (platinum/nickel) is ≤0.5 mm, it was observed that a uniform oxide layer of thickness 20 µm grew on the smooth internal surface of Zry-2 for corrosion lasting 1166 h. When d > 0.5 mm, the oxide thickness falls in a manner dictated by the degree of dissociation α of the electrolyte. At a cladding gap of 0.08 mm, a radiation-enhanced uniform corrosion rate of 2.405 10−1 mmpy was obtained for Zry-2. This value is 142 times greater than that obtained at room temperature in the absence of radiation. It was also observed that the corrosion rate falls at higher cladding gaps, and the rate of change depends on the degree of dissociation. Other phenomena such as the dynamics of shadow corrosion under varying electrode separation and electrolyte conductivities, as well as extensive evaluation of critical fuel cladding parameters, are presented in this work.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45725241","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Geant4 simulation was applied to correct the coincidence summing (CS) effect in detecting a volumetric γ-ray sources, and this technique was applied to a152Eu standard sources. The radioactive sources were a liquid cylindrical, rectangular, and Marinelli beaker shapes of different volume for each one. Radionuclide track (RT) including coincidence summing and monoenergetic track without coincidence summing. The results obtained from two approaches compared with the experimental data and the modified KORSUM code for cylindrical γ-ray source. The comparison showed that the adopted method in this investigation is useful for coincidence summing corrections for a voluminous γ-ray sources.Moreover, this technique requires far less computation time than the techniques that depend on the calculation of total efficiency.
{"title":"Coincidence Summing Factor Calculation for Volumetric γ-ray Sources Using Geant4 Simulation","authors":"D. Aloraini, M. Elsafi, A. Almuqrin, M. I. Sayyed","doi":"10.1155/2022/5718920","DOIUrl":"https://doi.org/10.1155/2022/5718920","url":null,"abstract":"Geant4 simulation was applied to correct the coincidence summing (CS) effect in detecting a volumetric γ-ray sources, and this technique was applied to a152Eu standard sources. The radioactive sources were a liquid cylindrical, rectangular, and Marinelli beaker shapes of different volume for each one. Radionuclide track (RT) including coincidence summing and monoenergetic track without coincidence summing. The results obtained from two approaches compared with the experimental data and the modified KORSUM code for cylindrical γ-ray source. The comparison showed that the adopted method in this investigation is useful for coincidence summing corrections for a voluminous γ-ray sources.Moreover, this technique requires far less computation time than the techniques that depend on the calculation of total efficiency.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41313786","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Lijun Jian, Pengkun Yu, J. Pei, Xiaoshu Zeng, Yidan Yuan
The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy-based viscosity model to realize a smooth transition of viscosity at the solid-liquid interface, and a surface tension model for better simulation of surface shape. Validation of the developed simulation approach is carried out on a classical water column collapse example. The development of the heat transfer model is validated by the example of a one-dimensional semi-infinite plate. A comprehensive example of the melting of “Wood’s alloy” is carried out to verify the capacity of MPS method in the simulation of melting and expansion. The simulation results are in good agreement with the experimental results, which indicates that MPS method promises well in the field of severe accidents.
{"title":"Development of an MPS Code for Corium Behavior Analysis: 3D Alloy Melting","authors":"Lijun Jian, Pengkun Yu, J. Pei, Xiaoshu Zeng, Yidan Yuan","doi":"10.1155/2022/2140729","DOIUrl":"https://doi.org/10.1155/2022/2140729","url":null,"abstract":"The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy-based viscosity model to realize a smooth transition of viscosity at the solid-liquid interface, and a surface tension model for better simulation of surface shape. Validation of the developed simulation approach is carried out on a classical water column collapse example. The development of the heat transfer model is validated by the example of a one-dimensional semi-infinite plate. A comprehensive example of the melting of “Wood’s alloy” is carried out to verify the capacity of MPS method in the simulation of melting and expansion. The simulation results are in good agreement with the experimental results, which indicates that MPS method promises well in the field of severe accidents.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47458580","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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