Science and Technology of Nuclear Installations最新文献_第6页

Effect of Chloride Ions on Electrochemical Behavior of Canister Materials 氯离子对罐状材料电化学行为的影响

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-08-03 DOI: 10.1155/2022/8577144

Gha-Young Kim, Junhyuk Jang, Minsoo Lee, Jin-Seop Kim

引用次数: 3

Characterization of Waste Generated from Nuclide Management Process in Waste Burden Minimization Technology for Spent Nuclear Fuel 乏燃料减负技术中核素管理过程产生的废物特性

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-07-30 DOI: 10.1155/2022/4764825

Jung-Hoon Choi, Byeonggwan Lee, Ki-rak Lee, H. Kang, H. J. Eom, Seong-Sik Shin, Ga-Yeong Kim, Hwan-seo Park

{"title":"Characterization of Waste Generated from Nuclide Management Process in Waste Burden Minimization Technology for Spent Nuclear Fuel","authors":"Jung-Hoon Choi, Byeonggwan Lee, Ki-rak Lee, H. Kang, H. J. Eom, Seong-Sik Shin, Ga-Yeong Kim, Hwan-seo Park","doi":"10.1155/2022/4764825","DOIUrl":"https://doi.org/10.1155/2022/4764825","url":null,"abstract":"To reduce the environmental burden caused by the disposal of spent nuclear fuel, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. To manufacture a waste form of high durability, the characteristics of the waste generated during the process should be evaluated. In this study, the physical, radiological, and thermal characteristics of the waste and waste forms for major nuclides (Cs, Sr, I, transuranic/rare earth, and Tc/Se) generated in the nuclide management process were analyzed. In the case of Cs nuclides, characterization was conducted according to the capture rate of the adsorbent in the high-temperature heat treatment process; meanwhile, in the case of Sr nuclides, characterization was performed by considering the ratio of similar nuclides in the chlorination process. For I nuclide, analysis was performed based on the available waste form, and for TRU/RE and Tc/Se nuclides, analysis was performed by considering chlorination and mid-temperature heat treatment. The radioactivity and heat generation rate of each waste and waste form were evaluated over a period of 1,000 years. The results of this study could be used to derive the centerline temperature for the thermal stability evaluation of waste forms and for the feasibility evaluation of each disposal system considered in the waste burden minimization technology.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48172368","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}

引用次数: 2

An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes 改进的基于MTR-DYN和EUREKA-2/RR程序的RSG-GAS堆芯RIA工况稳态和瞬态分析

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-07-30 DOI: 10.1155/2022/6030504

S. Pinem, S. Dibyo, W. Luthfi, V. Wardhani, D. Hartanto

{"title":"An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes","authors":"S. Pinem, S. Dibyo, W. Luthfi, V. Wardhani, D. Hartanto","doi":"10.1155/2022/6030504","DOIUrl":"https://doi.org/10.1155/2022/6030504","url":null,"abstract":"Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49574823","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}

引用次数: 0

Remaining Useful Life Prediction of Nuclear Power Machinery Based on an Exponential Degradation Model 基于指数退化模型的核电机械剩余使用寿命预测

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-06-16 DOI: 10.1155/2022/9895907

Gaojun Liu, Weijie Fan, Feng-lei Li, Gaixia Wang, Dongdong You

引用次数: 2

Study on the Dispersion of Radionuclides under Different Hydrological Conditions of Spent Fuel Shipping in Daya Bay 不同水文条件下放射性核素在大亚湾乏燃料运输中的扩散研究

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-06-10 DOI: 10.1155/2022/7265821

Liwei Chen, Wei Chen, Jiazhen Lin, Chunhua Chen, Yalin Luo, Longlong Tao

{"title":"Study on the Dispersion of Radionuclides under Different Hydrological Conditions of Spent Fuel Shipping in Daya Bay","authors":"Liwei Chen, Wei Chen, Jiazhen Lin, Chunhua Chen, Yalin Luo, Longlong Tao","doi":"10.1155/2022/7265821","DOIUrl":"https://doi.org/10.1155/2022/7265821","url":null,"abstract":"The radionuclide dispersion in coastal water is mainly controlled by the water flow and tidal effect. Tracing and analysis of radioactive pollutant dispersion in coastal water can predict distribution of radionuclide under marine transportation accident of spent fuel. In this work, factors such as continuous emission, radioactive decay, and water depth are considered, and a hydrodynamic model of radionuclide dispersion based on shallow water equations is established to simulate the dispersion of the radioactive pollutant in coastal waters under different hydrological conditions. As far as the characteristics of the radionuclide dispersion in coastal water are concerned, the simulation of pollutants by the hydrodynamic model is in good agreement with the work of Bailly du Bois et al., which validated the correctness of this model. The model has been applied to simulate the distribution of radionuclides in coastal water following a marine transport accident of spent fuel near Daya Bay Nuclear Power Plant in China. The simulation reveals that the distribution features are significantly affected by different hydrological conditions. In addition to limiting the diffusion range, the vortex effect can also cause the accumulation of radionuclides near the vortex, which helps to provide more practical information for nuclear emergency decision makers.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-06-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49473921","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}

引用次数: 1

Uranium Recovery from Phosphates for Self-Sufficient Nuclear Power in the Eastern Mediterranean 从磷酸盐中回收铀用于东地中海自给自足的核能

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-06-09 DOI: 10.1155/2022/3985408

N. Haidar

{"title":"Uranium Recovery from Phosphates for Self-Sufficient Nuclear Power in the Eastern Mediterranean","authors":"N. Haidar","doi":"10.1155/2022/3985408","DOIUrl":"https://doi.org/10.1155/2022/3985408","url":null,"abstract":"Production of phosphate fertilizers (PF), without uranium recovery, amounts to dispersing uranium compounds on agricultural fields. These compounds are naturally hidden in phosphate rock deposits prior to processing. Such a dispersion is a cumulative environmental damage, that may become rather catastrophic in few hundred years, under the current rates & impurities of phosphate fertilization of agricultural lands. It is also an avoidable irreversible waste of one of the world’s major energy resources. This study demonstrates quantitatively the low impact of U costs on the nuclear power generation costs, which happens, so far, to be a main reason for nonrecovery of uranium from the present PF industry. It reports on novel procedures for (i) estimating the required U feed to nuclear power plants (NPPs), (ii) pricing U as a function of its cumulative world production, and (iii) for quantifying U accumulation in phosphate fertilized lands. We also demonstrate that countries of the eastern Mediterranean can, in the long run, become collectively U partially self-sufficient, by recovering U from their phosphate resources, to power 13.2% of their entire electric energy generation contemporary needs.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-06-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"43740710","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}

引用次数: 0

Cold Atmospheric Plasma Inhibits the Proliferation of CAL-62 Cells through the ROS-Mediated PI3K/Akt/mTOR Signaling Pathway 低温大气等离子体通过ros介导的PI3K/Akt/mTOR信号通路抑制CAL-62细胞的增殖

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-06-08 DOI: 10.1155/2022/3884695

Fang Liu, Yuanyuan Zhou, Wencheng Song, Hongzhi Wang

引用次数: 2

An Information Granulated Based SVM Approach for Anomaly Detection of Main Transformers in Nuclear Power Plants 基于信息粒化的支持向量机在核电厂主变压器异常检测中的应用

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-06-03 DOI: 10.1155/2022/3931374

Wenmin Yu, Ren Yu, Cheng Li

引用次数: 1

Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis 用故障树分析法研究APR-1400给水损失事故

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-05-26 DOI: 10.1155/2022/4666161

M. Zubair

{"title":"Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis","authors":"M. Zubair","doi":"10.1155/2022/4666161","DOIUrl":"https://doi.org/10.1155/2022/4666161","url":null,"abstract":"Nuclear power plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2022-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"45154876","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}

引用次数: 0

Influence of Weak Compressibility on the Hydrodynamic Performance Evaluation of Pump Turbines in the Pump Mode 弱可压缩性对水泵水轮机泵态水动力性能评价的影响

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2022-05-20 DOI: 10.1155/2022/3544436

Fangfang Zhang, Nana Li, Di Zhu, R. Xiao, Weichao Liu, R. Tao

{"title":"Influence of Weak Compressibility on the Hydrodynamic Performance Evaluation of Pump Turbines in the Pump Mode","authors":"Fangfang Zhang, Nana Li, Di Zhu, R. Xiao, Weichao Liu, R. Tao","doi":"10.1155/2022/3544436","DOIUrl":"https://doi.org/10.1155/2022/3544436","url":null,"abstract":"In general, weak compressibility is one of the properties of liquids. That is, in actual operation of hydraulic machinery, the flow is weakly compressible. However, the influence of weak compressibility is often neglected in usual numerical simulation, which makes the simulation results different from the experimental results. Based on the Computational Fluid Dynamics (CFD) solver and model test rig, by means of mutual verification between numerical simulation and experiment, the fitting degree between numerical results and experimental results before and after considering weak compressibility is compared and analyzed in this paper; it is obtained that the numerical results is closer to the experimental results after considering the weak compressibility. In addition, velocity field of pump turbines, head loss of main components, and the change of entropy yield are analyzed and reasons for numerical value being closer to the experimental value after considering weak compressibility of fluid are summarized and analyzed. It is proved that the consideration of weak compressibility is of great significance to improve the accuracy of results in the numerical simulation of pump turbines.","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":"42854701","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}

引用次数: 4