Because of the characteristics of the small water volume of OTSG, it is hard to control the outlet steam pressure when the load is changed or disturbed. This study is devoted to the control of the once-through steam generator (OTSG). A double-layer controller based on the PPO algorithm is proposed to control the outlet steam pressure of OTSG. The bottom layer is the PID controller; it directly regulates the OTSG feed water valve and then controls the steam pressure. The top layer of the controller is the agent based on the PPO algorithm, which is responsible for optimizing the parameters of the PID in real time to obtain better control performance. The agent chooses PID parameters as actions to the environment, and then, the reward value is obtained through the reward function of the environment which enables online learning of the agent. Compared with the PID controller, the simulation experiment result shows that the method not only has a good control performance but also has a good anti-interference ability.
{"title":"Design of Control System of Once-Through Steam Generator Based on Proximal Policy Optimization Algorithm","authors":"Cheng Li, Ren Yu, Wenmin Yu, Tianshu Wang","doi":"10.1155/2022/2941705","DOIUrl":"https://doi.org/10.1155/2022/2941705","url":null,"abstract":"Because of the characteristics of the small water volume of OTSG, it is hard to control the outlet steam pressure when the load is changed or disturbed. This study is devoted to the control of the once-through steam generator (OTSG). A double-layer controller based on the PPO algorithm is proposed to control the outlet steam pressure of OTSG. The bottom layer is the PID controller; it directly regulates the OTSG feed water valve and then controls the steam pressure. The top layer of the controller is the agent based on the PPO algorithm, which is responsible for optimizing the parameters of the PID in real time to obtain better control performance. The agent chooses PID parameters as actions to the environment, and then, the reward value is obtained through the reward function of the environment which enables online learning of the agent. Compared with the PID controller, the simulation experiment result shows that the method not only has a good control performance but also has a good anti-interference ability.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-05-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44359845","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}
Yu‐Liang Zhang, Jin-Fu Li, Tao Wang, Jun-Jian Xiao, Xiaoqi Jia, Li Zhang
In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.
{"title":"Pressure Distribution on the Inner Wall of the Volute Casing of a Centrifugal Pump","authors":"Yu‐Liang Zhang, Jin-Fu Li, Tao Wang, Jun-Jian Xiao, Xiaoqi Jia, Li Zhang","doi":"10.1155/2022/3563459","DOIUrl":"https://doi.org/10.1155/2022/3563459","url":null,"abstract":"In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-05-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43737775","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}
Q. Tran, K. Nguyen, B. Luong, Ton-Nghiem Huynh, Q. Pham, Doan-Hai-Dang Vo, Nhi-Dien Nguyen
This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown.
{"title":"Determination of the Activity Inventory in the Structural Components of the Dalat Nuclear Research Reactor for Its Decommissioning Planning","authors":"Q. Tran, K. Nguyen, B. Luong, Ton-Nghiem Huynh, Q. Pham, Doan-Hai-Dang Vo, Nhi-Dien Nguyen","doi":"10.1155/2022/5174696","DOIUrl":"https://doi.org/10.1155/2022/5174696","url":null,"abstract":"This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-05-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47303688","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. Guo, Jinkai Wang, Chao Chen, Dongqin Xia, Nuo Yong, Daochuan Ge
Fusion reactors are expected to be safer, more environmentally friendly, and to have a lower nuclear proliferation risk, compared with other nuclear energy systems. However, it is widely recognized that a large amount of radioactive materials will be produced by a fusion reactor. Therefore, it is important to fully understand the overall radiation risk level of fusion radioactive wastes (radwaste) compared with existing nuclear energy systems. Studies on the treatment of the fusion radwaste have been currently focused on three ultimate options: clearance, recycling, and disposal by activation assessment of radioactive materials from the operation and decommissioning of fusion reactors. However, the radiation risk in the management of fusion radwaste, especially in the final disposal, was seldom studied. Based on the comparative analysis of fusion radioactive waste with ITER and fission reactors (e.g., pressurized water reactor, PWR), this paper tries to discuss how to determine the radiation risk in the process of fusion radwaste management on the premise of the current feasible industrial technology. On this basis, a risk assessment framework for repository disposal under normal degradation and external events is proposed.
{"title":"Preliminary Study on Risk Identification and Assessment Framework for Fusion Radioactive Waste Management","authors":"D. Guo, Jinkai Wang, Chao Chen, Dongqin Xia, Nuo Yong, Daochuan Ge","doi":"10.1155/2022/4870208","DOIUrl":"https://doi.org/10.1155/2022/4870208","url":null,"abstract":"Fusion reactors are expected to be safer, more environmentally friendly, and to have a lower nuclear proliferation risk, compared with other nuclear energy systems. However, it is widely recognized that a large amount of radioactive materials will be produced by a fusion reactor. Therefore, it is important to fully understand the overall radiation risk level of fusion radioactive wastes (radwaste) compared with existing nuclear energy systems. Studies on the treatment of the fusion radwaste have been currently focused on three ultimate options: clearance, recycling, and disposal by activation assessment of radioactive materials from the operation and decommissioning of fusion reactors. However, the radiation risk in the management of fusion radwaste, especially in the final disposal, was seldom studied. Based on the comparative analysis of fusion radioactive waste with ITER and fission reactors (e.g., pressurized water reactor, PWR), this paper tries to discuss how to determine the radiation risk in the process of fusion radwaste management on the premise of the current feasible industrial technology. On this basis, a risk assessment framework for repository disposal under normal degradation and external events is proposed.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-05-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48086325","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}
Achieving high neutron fluxes in research pressurized water reactors is directly related to the intensity of the coolant flow through the core and the pressure in it, which provides an increased saturation temperature and a margin to critical heat flux. Therefore, it is practically impossible to provide very high neutron fluxes in pool-type reactors, especially in the case of downward movement of the coolant in the core. At the same time, vessel-type research reactors (for example, SM-3 and HFIR) make it possible to achieve neutron flux densities up to 4 × 1015 n/(cm2 s), but at the same time, the risks of core degradation in case of violations in the heat removal system become quite high. The proposed concept of a heat removal system for a high-flux reactor facility combines the increased reliability of safe heat removal from the core and the convenience of handling irradiation cells, for example, in the production of isotopes. The concept provides for the location of a compact core in a pressurized vessel and the placement of a neutron reflector around the vessel in the reactor pool. Cooling of the reactor core in the housing and the irradiation channels in the neutron reflector is carried out by different systems of forced circulation of the coolant. At the same time, at the shutdown reactor, after opening the natural circulation valves, safe heat removal from the reactor core and the neutron reflector can be carried out by the water of the reactor pool. However, even with a complete failure of all forced circulation circuits, the evaporation of water from the surface of the pool makes it possible to safely remove the residual heat from the fuel assemblies and from the irradiation devices in the cells of the reflector.
{"title":"The Concept of the Heat Removal System of a High-Flux Research Reactor","authors":"Vitaly Uzikov, Ildar Suleimanov, Irina Uzikova","doi":"10.1155/2022/1815342","DOIUrl":"https://doi.org/10.1155/2022/1815342","url":null,"abstract":"Achieving high neutron fluxes in research pressurized water reactors is directly related to the intensity of the coolant flow through the core and the pressure in it, which provides an increased saturation temperature and a margin to critical heat flux. Therefore, it is practically impossible to provide very high neutron fluxes in pool-type reactors, especially in the case of downward movement of the coolant in the core. At the same time, vessel-type research reactors (for example, SM-3 and HFIR) make it possible to achieve neutron flux densities up to 4 × 10<sup>15</sup> n/(cm<sup>2</sup> s), but at the same time, the risks of core degradation in case of violations in the heat removal system become quite high. The proposed concept of a heat removal system for a high-flux reactor facility combines the increased reliability of safe heat removal from the core and the convenience of handling irradiation cells, for example, in the production of isotopes. The concept provides for the location of a compact core in a pressurized vessel and the placement of a neutron reflector around the vessel in the reactor pool. Cooling of the reactor core in the housing and the irradiation channels in the neutron reflector is carried out by different systems of forced circulation of the coolant. At the same time, at the shutdown reactor, after opening the natural circulation valves, safe heat removal from the reactor core and the neutron reflector can be carried out by the water of the reactor pool. However, even with a complete failure of all forced circulation circuits, the evaporation of water from the surface of the pool makes it possible to safely remove the residual heat from the fuel assemblies and from the irradiation devices in the cells of the reflector.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"85 1","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-04-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"138525075","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}
Bin Wu, Jinhua Wang, Yue Li, Haitao Wang, Tengyu Ma
The Chinese high-temperature gas-cooled reactor pebble-bed module, HTR-PM, began fuel loading in August 2021. The reactor refuels continuously, while the spent fuel is discharged from the core. The spent fuel conveying and loading system was designed to convey the spent fuel pebbles to the spent fuel building and load them into dry canisters for on-site interim storage. This study describes the operating principles of the main functions and introduces the experiments and commissioning tests of the system. Functional tests were carried out to indicate the items of mechanical and electrical equipment are functioning in accordance with the designed requirements. Experience learned from commissioning activities was also presented as feedback for future operation and design improvement.
{"title":"Design, Experiment, and Commissioning of the Spent Fuel Conveying and Loading System of HTR-PM","authors":"Bin Wu, Jinhua Wang, Yue Li, Haitao Wang, Tengyu Ma","doi":"10.1155/2022/1817191","DOIUrl":"https://doi.org/10.1155/2022/1817191","url":null,"abstract":"The Chinese high-temperature gas-cooled reactor pebble-bed module, HTR-PM, began fuel loading in August 2021. The reactor refuels continuously, while the spent fuel is discharged from the core. The spent fuel conveying and loading system was designed to convey the spent fuel pebbles to the spent fuel building and load them into dry canisters for on-site interim storage. This study describes the operating principles of the main functions and introduces the experiments and commissioning tests of the system. Functional tests were carried out to indicate the items of mechanical and electrical equipment are functioning in accordance with the designed requirements. Experience learned from commissioning activities was also presented as feedback for future operation and design improvement.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46241012","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. Guo, Jinkai Wang, Daochuan Ge, Chunhua Chen, Liwei Chen
Radionuclide dispersion model, which is of critical importance to the emergency response of severe nuclear accident, is used to estimate the consequences arising from accidental or routine releases and to predict areas of high contamination. It is difficult to evaluate the radioactive consequence accurately and rapidly for the accidental release of radionuclides from marine reactor because of the complex mobility feature in the sea. Based on CFD method, a finite-volume, three-dimensional regional oceanic dispersion model was developed in this paper to simulate the dispersion of radionuclides originating from marine reactor. The simulated dose variation of 137Cs presented good agreement with the monitoring data of marine radioactive pollution caused by Fukushima Dai-ichi nuclear accident, which demonstrated the validity of the method. A severe accident scenario of marine reactor was simulated and analyzed, which indicates that the regional oceanic dispersion model can provide dose assessment for nuclear emergency response.
{"title":"Oceanic Radionuclide Dispersion Method Investigation for Nonfixed Source from Marine Reactor Accident","authors":"D. Guo, Jinkai Wang, Daochuan Ge, Chunhua Chen, Liwei Chen","doi":"10.1155/2022/2822857","DOIUrl":"https://doi.org/10.1155/2022/2822857","url":null,"abstract":"Radionuclide dispersion model, which is of critical importance to the emergency response of severe nuclear accident, is used to estimate the consequences arising from accidental or routine releases and to predict areas of high contamination. It is difficult to evaluate the radioactive consequence accurately and rapidly for the accidental release of radionuclides from marine reactor because of the complex mobility feature in the sea. Based on CFD method, a finite-volume, three-dimensional regional oceanic dispersion model was developed in this paper to simulate the dispersion of radionuclides originating from marine reactor. The simulated dose variation of 137Cs presented good agreement with the monitoring data of marine radioactive pollution caused by Fukushima Dai-ichi nuclear accident, which demonstrated the validity of the method. A severe accident scenario of marine reactor was simulated and analyzed, which indicates that the regional oceanic dispersion model can provide dose assessment for nuclear emergency response.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-04-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47052488","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 use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling.
{"title":"Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU Reactors with Reactor Regulating System Response","authors":"S. Younan, D. Novog","doi":"10.1155/2022/7500629","DOIUrl":"https://doi.org/10.1155/2022/7500629","url":null,"abstract":"The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47776591","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}
A triangle subchannel model was established to study the flow and heat transfer characteristics of lead-bismuth eutectic (LBE) alloy and the erosion rate of the core channel by the particulate in LBE. Under different inlet velocities, particle types, particle diameters, and particle concentrations, the erosion law of the channel wall in LEB was investigated by using a discrete phase model (DPM). The results of this study showed that with the increase of inlet velocity, the outlet temperature of the LEB decreases and the heat transfer capacity was strengthened. The increase of inlet velocity will lead to the increase of erosion rate on the wall, and the change is exponential. The erosion rate of particulate in the low concentration is small but cannot be ignored; with increasing concentration of particulates, the erosion of the wall by particulates becomes serious. The effect of particulate density on the wall erosion rate can be ignored. The effect of changing the particle size on the erosion rate is more significant when the particle size is small, and at the same time, the erosion rate of the particles on the wall increases with the increase of the particle size.
{"title":"Investigation of the Flow and Heat Transfer Characteristics and Erosion Law of Particulate in LBE on the Subchannel","authors":"B. Zhu, Qi Xu, Pengxiang Li","doi":"10.1155/2022/2354978","DOIUrl":"https://doi.org/10.1155/2022/2354978","url":null,"abstract":"A triangle subchannel model was established to study the flow and heat transfer characteristics of lead-bismuth eutectic (LBE) alloy and the erosion rate of the core channel by the particulate in LBE. Under different inlet velocities, particle types, particle diameters, and particle concentrations, the erosion law of the channel wall in LEB was investigated by using a discrete phase model (DPM). The results of this study showed that with the increase of inlet velocity, the outlet temperature of the LEB decreases and the heat transfer capacity was strengthened. The increase of inlet velocity will lead to the increase of erosion rate on the wall, and the change is exponential. The erosion rate of particulate in the low concentration is small but cannot be ignored; with increasing concentration of particulates, the erosion of the wall by particulates becomes serious. The effect of particulate density on the wall erosion rate can be ignored. The effect of changing the particle size on the erosion rate is more significant when the particle size is small, and at the same time, the erosion rate of the particles on the wall increases with the increase of the particle size.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46009037","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}
H. Tekin, Fatema T. Ali, G. Almisned, G. Susoy, S. Issa, A. Ene, W. Elshami, H. Zakaly
In this study, the monotonic effect of Ta2O5 and ZrO2 in some selected borotellurite glasses was investigated in terms of their impact on gamma-ray-shielding competencies. Accordingly, three niobium-reinforced borotellurite glasses (S1 : 75TeO2 + 15B2O3 + 10Nb2O5, S2 : 75TeO2 + 15B2O3 + 9Nb2O5 + 1Ta2O5, and S3 : 75TeO2 + 15B2O3 + 8Nb2O5 + 1Ta2O5 + 1ZrO2) were modelled in the general-purpose MCNPX Monte Carlo code. They have been defined as an attenuator sample between the point isotropic gamma-ray source and the detector in terms of determining their attenuation coefficients. To verify the MC results, attenuation coefficients were then compared with the Phy-X/PSD program data. Our findings clearly demonstrate that although some behavioral changes occurred in the shielding qualities, modest improvements occurred in the attenuation properties depending on the modifier variation and its magnitude. However, the replacement of 2% moles of Nb2O5 with 1% mole of Ta2O5 and 1% mole of ZrO2 provided significant improvements in both glass density and attenuation properties against gamma rays. Finally, the HVL values of the S3 sample were compared with some glass- and concrete-shielding materials and the S3 sample was reported for its outstanding properties. As a consequence of this investigation, it can be concluded that the indicated type of additive to be added to borotellurite glasses will provide some advantages, particularly when used in radiation fields, by increasing the shielding qualities moderately.
{"title":"Multiple Assessments on the Gamma-Ray Protection Properties of Niobium-Doped Borotellurite Glasses: A Wide Range Investigation Using Monte Carlo Simulations","authors":"H. Tekin, Fatema T. Ali, G. Almisned, G. Susoy, S. Issa, A. Ene, W. Elshami, H. Zakaly","doi":"10.1155/2022/5890896","DOIUrl":"https://doi.org/10.1155/2022/5890896","url":null,"abstract":"In this study, the monotonic effect of Ta2O5 and ZrO2 in some selected borotellurite glasses was investigated in terms of their impact on gamma-ray-shielding competencies. Accordingly, three niobium-reinforced borotellurite glasses (S1 : 75TeO2 + 15B2O3 + 10Nb2O5, S2 : 75TeO2 + 15B2O3 + 9Nb2O5 + 1Ta2O5, and S3 : 75TeO2 + 15B2O3 + 8Nb2O5 + 1Ta2O5 + 1ZrO2) were modelled in the general-purpose MCNPX Monte Carlo code. They have been defined as an attenuator sample between the point isotropic gamma-ray source and the detector in terms of determining their attenuation coefficients. To verify the MC results, attenuation coefficients were then compared with the Phy-X/PSD program data. Our findings clearly demonstrate that although some behavioral changes occurred in the shielding qualities, modest improvements occurred in the attenuation properties depending on the modifier variation and its magnitude. However, the replacement of 2% moles of Nb2O5 with 1% mole of Ta2O5 and 1% mole of ZrO2 provided significant improvements in both glass density and attenuation properties against gamma rays. Finally, the HVL values of the S3 sample were compared with some glass- and concrete-shielding materials and the S3 sample was reported for its outstanding properties. As a consequence of this investigation, it can be concluded that the indicated type of additive to be added to borotellurite glasses will provide some advantages, particularly when used in radiation fields, by increasing the shielding qualities moderately.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"44100267","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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