In alkaline in-situ leaching uranium, oxygen is the most common oxidizer with bicarbonate as a complexing agent. For those sandstone uranium deposits with strongly reductive capacity or complicated hydrogeological environment, the oxidation by oxygen is low efficiency. An efficient leaching method, therefore, is needed for these uranium deposits. In this study, a typical sandstone uranium deposit which characterizes with high TDS and high chloride content in groundwater and intractable uranium leach is selected to investigate the effects of synergetic oxidation by a strong oxidant with oxygen. Based on the research on batch leach, pressure leach and field trials, the oxidants such as hydrogen peroxide, potassium permanganate and sodium dichloroisocyanurate (NaDCC) are tested. The results of pressure batch leach indicate that synergetic oxidization is achieved by NaDCC in oxygen leaching process. Leaching tests indicate that a minor oxidizer of NaDCC shows good synergetic oxidization with oxygen and leaching effects on uranium minerals. The results also demonstrate that hydrogen peroxide shows no oxidation effects when it is used as a single oxidant. While potassium permanganate shows good oxidation on uranium when it is used as a single oxidant, however, it leads inhibiting effects on oxygen oxidation on uranium minerals. The further field tests are conducted to study the synergetic effects of oxygen with and without sodium dichloroisocyanurate. The preliminary results indicate that a fast leach is observed by the composite oxidants in early stage while no synergetic leach is found after 200 days. Further studies should be conducted in laboratory experiments and pilot scale tests for its potential applications.
{"title":"Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study","authors":"Wensheng Liao, Weimin Que, Limin Wang, Zhiming Du","doi":"10.1115/icone2020-16200","DOIUrl":"https://doi.org/10.1115/icone2020-16200","url":null,"abstract":"\u0000 In alkaline in-situ leaching uranium, oxygen is the most common oxidizer with bicarbonate as a complexing agent. For those sandstone uranium deposits with strongly reductive capacity or complicated hydrogeological environment, the oxidation by oxygen is low efficiency. An efficient leaching method, therefore, is needed for these uranium deposits. In this study, a typical sandstone uranium deposit which characterizes with high TDS and high chloride content in groundwater and intractable uranium leach is selected to investigate the effects of synergetic oxidation by a strong oxidant with oxygen. Based on the research on batch leach, pressure leach and field trials, the oxidants such as hydrogen peroxide, potassium permanganate and sodium dichloroisocyanurate (NaDCC) are tested. The results of pressure batch leach indicate that synergetic oxidization is achieved by NaDCC in oxygen leaching process. Leaching tests indicate that a minor oxidizer of NaDCC shows good synergetic oxidization with oxygen and leaching effects on uranium minerals. The results also demonstrate that hydrogen peroxide shows no oxidation effects when it is used as a single oxidant. While potassium permanganate shows good oxidation on uranium when it is used as a single oxidant, however, it leads inhibiting effects on oxygen oxidation on uranium minerals. The further field tests are conducted to study the synergetic effects of oxygen with and without sodium dichloroisocyanurate. The preliminary results indicate that a fast leach is observed by the composite oxidants in early stage while no synergetic leach is found after 200 days. Further studies should be conducted in laboratory experiments and pilot scale tests for its potential applications.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75333561","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}
Takatsugu Miura, K. Igarashi, Tomoyuki Hosaka, Takumi Kitagawa, Tatsurou Yashiki, Yuki Itabayashi, Hirotsugu Suzuki, T. Nakahara, J. Kitamura, T. Sano
In power plants that becoming more compact, it will expend much time and effort to satisfy the requirement for the differential pressure flow measurement according to ISO’s standards. Therefore, it is difficult for engineers in the design phase to completely remove the potential for large errors in flow measurement. This paper presents the 3D fluid analysis system that is a lower cost than the conventional method to confirm the soundness of such measurement in the phase of piping route design. This system has the function to automatically generate the analysis models from general 3D piping CAD data. The analysis program is written by the open source code to reduce a license fee. Also, this system has the function of calculating the swirl strength along the pipe axis as one of the means for efficiently supporting the design change. In order to verify and validate the analysis system, we analyzed several flow paths, confirmed the response of the swirl strength and flow rate indication value of the differential pressure flowmeter model. The analysis result well simulated the increase or decrease swirl strength in the complex flow path, and fluctuation of the flow rate indication value. Also, the system supports to set the flowmeter in the appropriate position by providing visualization of the swirl strength along the pipe axis. In the flow path analysis in this validation, it took about one month to visualization of the swirl strength along the pipe axis from the generation of the analysis models. The 3D fluid analysis system collaborative with 3D piping CAD design system has been developed. This system enable to confirm the effects of swirl strength on flow measurement and the soundness of the differential pressure flow measurement at a lower cost in comparison with conventional method.
{"title":"Development of Design Support System for Piping Route and Differential Pressure Flowmeter by Three-Dimensional Fluid Analysis","authors":"Takatsugu Miura, K. Igarashi, Tomoyuki Hosaka, Takumi Kitagawa, Tatsurou Yashiki, Yuki Itabayashi, Hirotsugu Suzuki, T. Nakahara, J. Kitamura, T. Sano","doi":"10.1115/icone2020-16745","DOIUrl":"https://doi.org/10.1115/icone2020-16745","url":null,"abstract":"\u0000 In power plants that becoming more compact, it will expend much time and effort to satisfy the requirement for the differential pressure flow measurement according to ISO’s standards. Therefore, it is difficult for engineers in the design phase to completely remove the potential for large errors in flow measurement. This paper presents the 3D fluid analysis system that is a lower cost than the conventional method to confirm the soundness of such measurement in the phase of piping route design.\u0000 This system has the function to automatically generate the analysis models from general 3D piping CAD data. The analysis program is written by the open source code to reduce a license fee. Also, this system has the function of calculating the swirl strength along the pipe axis as one of the means for efficiently supporting the design change. In order to verify and validate the analysis system, we analyzed several flow paths, confirmed the response of the swirl strength and flow rate indication value of the differential pressure flowmeter model. The analysis result well simulated the increase or decrease swirl strength in the complex flow path, and fluctuation of the flow rate indication value. Also, the system supports to set the flowmeter in the appropriate position by providing visualization of the swirl strength along the pipe axis. In the flow path analysis in this validation, it took about one month to visualization of the swirl strength along the pipe axis from the generation of the analysis models.\u0000 The 3D fluid analysis system collaborative with 3D piping CAD design system has been developed. This system enable to confirm the effects of swirl strength on flow measurement and the soundness of the differential pressure flow measurement at a lower cost in comparison with conventional method.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80135525","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}
Kaiyue Shen, Weizhen Zheng, Shengchao Ma, Huaqiang Yin, Xuedong He, T. Ma
A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.
{"title":"Study on Heating Process of Dehumidifying Experiment in HTGR","authors":"Kaiyue Shen, Weizhen Zheng, Shengchao Ma, Huaqiang Yin, Xuedong He, T. Ma","doi":"10.1115/icone2020-16358","DOIUrl":"https://doi.org/10.1115/icone2020-16358","url":null,"abstract":"A large number of carbon materials are used in high temperature gas-cooled reactor (HTGR). As a kind of porous material, the carbon material contains a certain amount of moisture and other impurities. In order to reduce the corrosion of internal material in reactor core of HTGR, the initial core or post-accident core must be strictly heated and dehumidified. The current primary circuit heating mainly relies on the rotation of the primary pump to convert the kinetic energy into thermal energy. Obviously, the current scheme was flawed: (1) Due to the insufficient heat generated by rotation of the primary pump, the temperature rising process of the primary circuit is sluggish; (2) The rotation of the primary pump converts the kinetic energy into thermal energy of the helium, at the meantime, the primary circuit dissipates heat outward. For the above reasons, it is difficult to achieve the desired dehumidification temperature in the heating process. While in this paper, an additional thermal source will be added to the steam generator to heat the primary circuit in a new scheme. A proper flow and heat-transfer model of heating the primary circuit in high-temperature reactor was established based on software COMSOL Multiphysics. The numerical analysis of the primary circuit heating process provides rewarding guidance for the selection of the dehumidification scheme in HTGR.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91100404","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 ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed. And then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (RIA and LOCA) are analyzed and compared. The Th0.923U0.077O2 fuel is found to have lower fuel center-line temperature and the thorium-based fuels are observed to have a delayed pellet-cladding mechanical interaction (PCMI) under steady state. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than conventional UO2 fuel under both normal operating conditions and accident conditions.
{"title":"Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident","authors":"Chenjie Qiu, Rong Liu, Wenzhong Zhou","doi":"10.1115/icone2020-16325","DOIUrl":"https://doi.org/10.1115/icone2020-16325","url":null,"abstract":"\u0000 The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed. And then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (RIA and LOCA) are analyzed and compared. The Th0.923U0.077O2 fuel is found to have lower fuel center-line temperature and the thorium-based fuels are observed to have a delayed pellet-cladding mechanical interaction (PCMI) under steady state. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than conventional UO2 fuel under both normal operating conditions and accident conditions.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78246634","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}
I. Ioka, Y. Kuriki, J. Iwatsuki, Daisuke Kawai, Y. Inagaki, S. Kubo
A thermochemical water-splitting iodine-sulfur process (IS process) is one of candidates for the large-scale production of hydrogen using heat from nuclear energy. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. To achieve an industrialization of massive hydrogen production system, one of the key factors is the development of structural materials for the severe corrosive environment. A hybrid material with the corrosion-resistance and the ductility had been made by a silicon powder plasma spraying and laser treatment. To confirm the applicability of the hybrid material as the structural material, corrosion tests of the hybrid materials had been performed in 95 mass% and 47 mass% boiling sulfuric acid. The corrosion resistance of specimen in the condition of 95 mass% boiling sulfuric acid had been excellent. This was attributed to the formation of SiO2 on the surface. To confirm the production characteristics as a container using the hybrid material, the container which has a welded part, a chamfer, a curved surface had been experimentally made. A configuration of the container had been 150mm inside diameter, 120mm in height and 6mm in thickness. The substrate of the container made of Hastelloy C276® superalloy had included TIG weld part. To improve the corrosion resistance of the container, pre-oxidation was performed at 800°C for 100 hours in air. There was no detachment of the plasma spraying and laser treated layer on the base metal and the welded part. The pre-oxidized container using hybrid technique was prepared for the corrosion test in boiling sulfuric acid to evaluate the characteristics of the container.
{"title":"Corrosion Property of Container Using Hybrid Material for Thermal Decomposition Process of Sulfuric Acid","authors":"I. Ioka, Y. Kuriki, J. Iwatsuki, Daisuke Kawai, Y. Inagaki, S. Kubo","doi":"10.1115/icone2020-16783","DOIUrl":"https://doi.org/10.1115/icone2020-16783","url":null,"abstract":"\u0000 A thermochemical water-splitting iodine-sulfur process (IS process) is one of candidates for the large-scale production of hydrogen using heat from nuclear energy. Severe corrosive environment which is thermal decomposition of sulfuric acid exists in the IS process. To achieve an industrialization of massive hydrogen production system, one of the key factors is the development of structural materials for the severe corrosive environment. A hybrid material with the corrosion-resistance and the ductility had been made by a silicon powder plasma spraying and laser treatment. To confirm the applicability of the hybrid material as the structural material, corrosion tests of the hybrid materials had been performed in 95 mass% and 47 mass% boiling sulfuric acid. The corrosion resistance of specimen in the condition of 95 mass% boiling sulfuric acid had been excellent. This was attributed to the formation of SiO2 on the surface. To confirm the production characteristics as a container using the hybrid material, the container which has a welded part, a chamfer, a curved surface had been experimentally made. A configuration of the container had been 150mm inside diameter, 120mm in height and 6mm in thickness. The substrate of the container made of Hastelloy C276® superalloy had included TIG weld part. To improve the corrosion resistance of the container, pre-oxidation was performed at 800°C for 100 hours in air. There was no detachment of the plasma spraying and laser treated layer on the base metal and the welded part. The pre-oxidized container using hybrid technique was prepared for the corrosion test in boiling sulfuric acid to evaluate the characteristics of the container.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75991710","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 large area of east Japan was contaminated by radiocesium following a nuclear accident at the Fukushima Daiichi Nuclear Power Station. Following decontamination of the soil, external effective dose conversion factors were calculated by changing the volume of decontamination soil, depth of cover soil, and distance of the evaluation point from the decontamination soil. The decrement of the factors with an increase of the distance was larger for the smaller volume of decontamination soil. The factors decrease exponentially with an increase of the depth of cover soil in all cases. When there was no cover soil, annual external exposure doses for residents at 1 m from the repository site and public entry were over 10 μSv/y, even for the smallest size (2m × 2m × 1m) and 50 percentile value of radiation concentration (700 Bq/kg). When the surface was covered by 30 cm of non-contaminated soil, the annual external exposure doses were less than 10 μSv/y for the largest size (200m × 200m × 10m) and 95 percentile concentration (2500 Bq/kg).
{"title":"Sensitivity Analysis of External Exposure Dose for Future Burial Measures of Decontamination Soil Generated Outside Fukushima Prefecture","authors":"A. Shimada, T. Sawaguchi, S. Takeda","doi":"10.1115/icone2020-16826","DOIUrl":"https://doi.org/10.1115/icone2020-16826","url":null,"abstract":"\u0000 A large area of east Japan was contaminated by radiocesium following a nuclear accident at the Fukushima Daiichi Nuclear Power Station. Following decontamination of the soil, external effective dose conversion factors were calculated by changing the volume of decontamination soil, depth of cover soil, and distance of the evaluation point from the decontamination soil. The decrement of the factors with an increase of the distance was larger for the smaller volume of decontamination soil. The factors decrease exponentially with an increase of the depth of cover soil in all cases. When there was no cover soil, annual external exposure doses for residents at 1 m from the repository site and public entry were over 10 μSv/y, even for the smallest size (2m × 2m × 1m) and 50 percentile value of radiation concentration (700 Bq/kg). When the surface was covered by 30 cm of non-contaminated soil, the annual external exposure doses were less than 10 μSv/y for the largest size (200m × 200m × 10m) and 95 percentile concentration (2500 Bq/kg).","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73680283","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}
To better understand irradiation creep of nuclear fuel, NRG has prepared, as part of the H2020 European project INSPYRE, a separate effect irradiation experiment in the High Flux Reactor (HFR) in Petten (the Netherlands) aiming to measure fuel creep in-pile as a function of temperature, flux, burn-up and axial pressure load. This continuous type of measurement will supply a large data set, leading to more detailed knowledge on fuel behaviour during irradiation. To support the experiment and make optimal use of the generated data, a model is created of the experiment to better predict the behaviour of the fuel samples during irradiation. The current paper describes the numerical model, which couples the 1.5D fuel performance code TRANSURANUS (TU) with a Finite Element Analysis (FEA). The thermal analysis of the experiment is carried out using the FEA. Such approach enables to model a rather complex geometry of the experiment, and to include axial heat transport, which is not implemented in TU. TU is modified in order to use the fuel pellet temperatures obtained using the FEA and to include the axial load present in the experiment. The model is validated against several test cases and used to predict the fuel behaviour during a selection of foreseen irradiation scenario’s. Results of the model will be used in the future for optimization of the irradiation parameters used in the experiment and for analysis of the data obtained during the irradiation.
{"title":"Modelling of the H2020 INSPYRE Fuel Creep Experiment","authors":"A. Fedorov, Kevin Zwijsen, S. V. Til","doi":"10.1115/icone2020-16231","DOIUrl":"https://doi.org/10.1115/icone2020-16231","url":null,"abstract":"\u0000 To better understand irradiation creep of nuclear fuel, NRG has prepared, as part of the H2020 European project INSPYRE, a separate effect irradiation experiment in the High Flux Reactor (HFR) in Petten (the Netherlands) aiming to measure fuel creep in-pile as a function of temperature, flux, burn-up and axial pressure load. This continuous type of measurement will supply a large data set, leading to more detailed knowledge on fuel behaviour during irradiation. To support the experiment and make optimal use of the generated data, a model is created of the experiment to better predict the behaviour of the fuel samples during irradiation. The current paper describes the numerical model, which couples the 1.5D fuel performance code TRANSURANUS (TU) with a Finite Element Analysis (FEA). The thermal analysis of the experiment is carried out using the FEA. Such approach enables to model a rather complex geometry of the experiment, and to include axial heat transport, which is not implemented in TU. TU is modified in order to use the fuel pellet temperatures obtained using the FEA and to include the axial load present in the experiment. The model is validated against several test cases and used to predict the fuel behaviour during a selection of foreseen irradiation scenario’s. Results of the model will be used in the future for optimization of the irradiation parameters used in the experiment and for analysis of the data obtained during the irradiation.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77004374","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}
It is of practical significance to analyze the multi-physics process of nuclear system, which includes neutronics, heat transfer and thermoelasticity. Fission reaction is the heat source in system, the heat source will affect the distribution of temperature field, which will lead to the change of strain. Strain in turn will affect the distribution of neutron field. Therefore, it is necessary to analyze the distribution of neutron flux, temperature and strain in system. Three aspects of work have been carried out: 1) Based on Galerkin finite element theory, the governing equations of neutronics, heat transfer and thermoelasticity are established; 2) The multi-physics analysis code is developed; 3) The calculation results are analyzed and discussed.
{"title":"Numerical Simulation of Multi-Physics Processes in Nuclear System Based on Galerkin Finite Element Method","authors":"Bao-Xin Yuan, Wankui Yang, Songbao Zhang, Bin Zhong, Junxia Wei, Yangjun Ying","doi":"10.1115/icone2020-16801","DOIUrl":"https://doi.org/10.1115/icone2020-16801","url":null,"abstract":"\u0000 It is of practical significance to analyze the multi-physics process of nuclear system, which includes neutronics, heat transfer and thermoelasticity. Fission reaction is the heat source in system, the heat source will affect the distribution of temperature field, which will lead to the change of strain. Strain in turn will affect the distribution of neutron field. Therefore, it is necessary to analyze the distribution of neutron flux, temperature and strain in system. Three aspects of work have been carried out: 1) Based on Galerkin finite element theory, the governing equations of neutronics, heat transfer and thermoelasticity are established; 2) The multi-physics analysis code is developed; 3) The calculation results are analyzed and discussed.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87363784","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. Nishida, Zuoyi Kang, Y. Okuda, H. Tsubota, Yinsheng Li
Studies on the local damage to reinforced concrete (RC) panels subjected to projectile impact have mainly focused on collisions that occur at an angle normal to the structure; thus, research on oblique impact is scarce. Due to this, we conducted research focusing on oblique impact to enable more realistic impact assessment of projectile collisions. To date, the validity of the analytical method has been confirmed by comparing the results with those of previous tests, and the local damage to RC panels that have collided with projectiles has been analytically investigated focusing on the impact angle. Therefore, this study aims to confirm the validity of the analysis method by conducting specific impact tests under various conditions, including the impact angle, by obtaining the relevant data. This paper outlines the test for the local damage to RC panels subjected to both normal and oblique impacts.
{"title":"Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact by Projectiles: Outline of Impact Test","authors":"A. Nishida, Zuoyi Kang, Y. Okuda, H. Tsubota, Yinsheng Li","doi":"10.1115/icone2020-16843","DOIUrl":"https://doi.org/10.1115/icone2020-16843","url":null,"abstract":"\u0000 Studies on the local damage to reinforced concrete (RC) panels subjected to projectile impact have mainly focused on collisions that occur at an angle normal to the structure; thus, research on oblique impact is scarce. Due to this, we conducted research focusing on oblique impact to enable more realistic impact assessment of projectile collisions. To date, the validity of the analytical method has been confirmed by comparing the results with those of previous tests, and the local damage to RC panels that have collided with projectiles has been analytically investigated focusing on the impact angle. Therefore, this study aims to confirm the validity of the analysis method by conducting specific impact tests under various conditions, including the impact angle, by obtaining the relevant data. This paper outlines the test for the local damage to RC panels subjected to both normal and oblique impacts.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90062165","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}
After the loss of coolant accident (LOCA), the safety injection system injects water into the reactor coolant system (RCS), and the residual heat rejects from the break. The containment spray system is operating in recirculating cooling mode to ensure that the containment is cooldown. This state must be maintained for several months. After the accident, in order to respond the design extension conditions (DEC) of failure of two containment spray pumps or two low pressure safety injection pumps, the design of the original H4 connections was improved, and the H4 procedure (loss of containment spray pumps or low pressure safety injection pumps) was developed. H4 procedure demands to put into service 2 permanent (one for each train) interconnections of containment spray system and safety injection system, called “H4 connections”. Through the design improvement of the H4 connections, the mutual backup function of safety injection and containment spray can be realized implemented. The manual valves of the H4 connections in the original design were changed to electric valves, which ensured the accessibility of operator and avoided the radiation of high radioactivity level to operator after the accident. In addition, the improved H4 connections enable mutual backup of safety injection and containment spray in the early stage after the accident to be implemented, which fully improves the ability to respond to accidents and safety design level of the nuclear power plant (NPP). This also makes it possible to intervene early after the accident.
{"title":"Impact Analysis of NPP H4 Connections Design Improvement on Emergency Operation","authors":"W. Yuqi, Yi Ke","doi":"10.1115/icone2020-16311","DOIUrl":"https://doi.org/10.1115/icone2020-16311","url":null,"abstract":"\u0000 After the loss of coolant accident (LOCA), the safety injection system injects water into the reactor coolant system (RCS), and the residual heat rejects from the break. The containment spray system is operating in recirculating cooling mode to ensure that the containment is cooldown. This state must be maintained for several months. After the accident, in order to respond the design extension conditions (DEC) of failure of two containment spray pumps or two low pressure safety injection pumps, the design of the original H4 connections was improved, and the H4 procedure (loss of containment spray pumps or low pressure safety injection pumps) was developed. H4 procedure demands to put into service 2 permanent (one for each train) interconnections of containment spray system and safety injection system, called “H4 connections”. Through the design improvement of the H4 connections, the mutual backup function of safety injection and containment spray can be realized implemented. The manual valves of the H4 connections in the original design were changed to electric valves, which ensured the accessibility of operator and avoided the radiation of high radioactivity level to operator after the accident. In addition, the improved H4 connections enable mutual backup of safety injection and containment spray in the early stage after the accident to be implemented, which fully improves the ability to respond to accidents and safety design level of the nuclear power plant (NPP). This also makes it possible to intervene early after the accident.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79035074","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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