Carbon-14 is one of the most important radionuclides discharged to the environment from pressurized water reactors due to its long half-life and its important role in the biological chain. Carbon-14 is the largest contributor of the dose rate to the public from all radionuclides discharged to the environment during the normal operation of pressurized water reactors, and thus the production and discharge of Carbon-14 have been focused on in the industry in recent years. Based on the generation mechanism in pressurized water reactor, one theoretical model of Carbon-14 generation has been established and the nitrogen concentration of all chemical species in the primary loops has been determined according to the measured ammonium. It predicts that the annual Carbon-14 generation in the primary loops is proportional to annual electricity output and the typical normalized Carbon-14 is 2.9E+02 GBq/GWe/yr. The theoretical model has been verified by the statistical analysis of annual Carbon-14 discharges from French PWR units. In addition, the quantity of Carbon-14 in radioactive solid waste has been estimated in these PWR units. It shows the generation of Carbon-14 in PWR cannot be effectively minimized because O-17 atoms, the predominant origin of Carbon-14, exist naturally in the primary loops during long-term operation. This approach can be applied to analyze the Carbon-14 production and discharges in operating pressurized water reactors and in the assessment of source term of the new pressurized water reactors.
{"title":"Research on the Expected Carbon-14 Production and Discharge in Pressurized Water Reactors","authors":"Pengtao Fu","doi":"10.1115/icone29-92807","DOIUrl":"https://doi.org/10.1115/icone29-92807","url":null,"abstract":"\u0000 Carbon-14 is one of the most important radionuclides discharged to the environment from pressurized water reactors due to its long half-life and its important role in the biological chain. Carbon-14 is the largest contributor of the dose rate to the public from all radionuclides discharged to the environment during the normal operation of pressurized water reactors, and thus the production and discharge of Carbon-14 have been focused on in the industry in recent years.\u0000 Based on the generation mechanism in pressurized water reactor, one theoretical model of Carbon-14 generation has been established and the nitrogen concentration of all chemical species in the primary loops has been determined according to the measured ammonium. It predicts that the annual Carbon-14 generation in the primary loops is proportional to annual electricity output and the typical normalized Carbon-14 is 2.9E+02 GBq/GWe/yr. The theoretical model has been verified by the statistical analysis of annual Carbon-14 discharges from French PWR units. In addition, the quantity of Carbon-14 in radioactive solid waste has been estimated in these PWR units. It shows the generation of Carbon-14 in PWR cannot be effectively minimized because O-17 atoms, the predominant origin of Carbon-14, exist naturally in the primary loops during long-term operation. This approach can be applied to analyze the Carbon-14 production and discharges in operating pressurized water reactors and in the assessment of source term of the new pressurized water reactors.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"21 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117220216","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}
With the continuous development of science and technology and the iteration of various types of low-voltage protection products, electrical designers are focusing not only on the reliability and security of the power supply system but also its economic benefits. From the calculation of electricity load in the early stage, to the selection of transformers, and then to the determination of the number of low-voltage distribution cabinets, as well as the selection of each feeder circuit protection devices, designers strive to come up with the most reliable and economical choice in each link through rigorous calculation and analysis. In the design and installation of large communities or factories, a small technical improvement can bring substantial economic benefits. Combined with engineering practice, this paper briefly introduces several protective appliances and their functions in low-voltage distribution circuits, the main differences between current limiting circuit breakers and ordinary circuit breakers are also revealed. The selection principle and basis of current limiting circuit breaker in the electrical design of a nuclear chemical engineering project are described in detail. The economic benefits and significance of selecting current limiting circuit breaker are compared and analyzed in this paper. The development prospects of current limiting circuit breaker selection in power supply and distribution system in large nuclear chemical plants and its key role for the whole plant in the future are discussed.
{"title":"Selection and Economic Benefit Analysis of Current Limiting Circuit Breaker in a Project of a Newly-Built Large Nuclear Chemical Plant","authors":"Shuo Gao, Shizhong Tian, Zhi-gang Huang","doi":"10.1115/icone29-92658","DOIUrl":"https://doi.org/10.1115/icone29-92658","url":null,"abstract":"\u0000 With the continuous development of science and technology and the iteration of various types of low-voltage protection products, electrical designers are focusing not only on the reliability and security of the power supply system but also its economic benefits. From the calculation of electricity load in the early stage, to the selection of transformers, and then to the determination of the number of low-voltage distribution cabinets, as well as the selection of each feeder circuit protection devices, designers strive to come up with the most reliable and economical choice in each link through rigorous calculation and analysis. In the design and installation of large communities or factories, a small technical improvement can bring substantial economic benefits. Combined with engineering practice, this paper briefly introduces several protective appliances and their functions in low-voltage distribution circuits, the main differences between current limiting circuit breakers and ordinary circuit breakers are also revealed. The selection principle and basis of current limiting circuit breaker in the electrical design of a nuclear chemical engineering project are described in detail. The economic benefits and significance of selecting current limiting circuit breaker are compared and analyzed in this paper. The development prospects of current limiting circuit breaker selection in power supply and distribution system in large nuclear chemical plants and its key role for the whole plant in the future are discussed.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"144 4","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134323538","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 moisture content of spent resin from NPPs is generally 50% ∼ 60%, which concentrates radionuclides such as 60Co, 90Sr and 137Cs. It belongs to wet solid waste, accounting for about 1/4 of the total solid waste of NPPs. Pent resin is a dispersive material, which must be stabilized and packaged to meet the requirements of waste transportation and final disposal; However, the characteristics of radioactive organic waste also cause the radiolysis of spent resin and the production of H2, CH4, NH3 and other gases; The spent resin contains more sulfur and nitrogen, and the incineration products and degradation products have strong corrosive to the equipment and storage containers. Therefore, how to treat the spent resin safely and effectively is a difficult problem.Microwave process for spent resin using microwave penetration ability and body heat characteristics, remove the water and organic components from spent resin, so as to achieve the inorganic for organic waste. Compared with the original resin, the ashing product capacity and weight reduction ratio greatly. The spent resin dried by microwave does not contain free water, the volume is 1/3 of the resin volume before drying, and the mass is half of the resin mass before drying; After Microwave ashing, the ashing product of the spent resin does not contain organic components, and the ashing product is only 1/20 of the mass of the dried resin and 1/3 of the volume of the dried resin, that is, for the spent resin with a moisture content of about 50%, after microwave drying and ashing process, not only the organic components in the spent resin are removed, but also the volume and quality are greatly reduced. Most of the treated radionuclides exist in the ashing products, 60Co in the ashing products is 1011 Bq/kg and 137Cs is 108 Bq/kg. The technology processing are used to get the product, which can be convenient for subsequent processing. On the study, the technological parameters of microwave drum drying and microwave ashing of spent resin are determined.
{"title":"The Study on Microwave Process for Spent Resin","authors":"Gao Chao, J. Meilan, An Hongxiang","doi":"10.1115/icone29-92401","DOIUrl":"https://doi.org/10.1115/icone29-92401","url":null,"abstract":"\u0000 The moisture content of spent resin from NPPs is generally 50% ∼ 60%, which concentrates radionuclides such as 60Co, 90Sr and 137Cs. It belongs to wet solid waste, accounting for about 1/4 of the total solid waste of NPPs. Pent resin is a dispersive material, which must be stabilized and packaged to meet the requirements of waste transportation and final disposal; However, the characteristics of radioactive organic waste also cause the radiolysis of spent resin and the production of H2, CH4, NH3 and other gases; The spent resin contains more sulfur and nitrogen, and the incineration products and degradation products have strong corrosive to the equipment and storage containers. Therefore, how to treat the spent resin safely and effectively is a difficult problem.Microwave process for spent resin using microwave penetration ability and body heat characteristics, remove the water and organic components from spent resin, so as to achieve the inorganic for organic waste. Compared with the original resin, the ashing product capacity and weight reduction ratio greatly. The spent resin dried by microwave does not contain free water, the volume is 1/3 of the resin volume before drying, and the mass is half of the resin mass before drying; After Microwave ashing, the ashing product of the spent resin does not contain organic components, and the ashing product is only 1/20 of the mass of the dried resin and 1/3 of the volume of the dried resin, that is, for the spent resin with a moisture content of about 50%, after microwave drying and ashing process, not only the organic components in the spent resin are removed, but also the volume and quality are greatly reduced. Most of the treated radionuclides exist in the ashing products, 60Co in the ashing products is 1011 Bq/kg and 137Cs is 108 Bq/kg. The technology processing are used to get the product, which can be convenient for subsequent processing. On the study, the technological parameters of microwave drum drying and microwave ashing of spent resin are determined.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"579 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123038934","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}
According to the conceptual design of high-level radioactive waste (HLW) disposal repositories, the buffer barrier will be constructed by compacted bentonite blocks, which will inevitably produce technological gaps (voids) in the repository. During the consequent long-term operation of the repository, the technological gaps will be gradually filled up on hydration of the bentonite blocks with infiltration of underground water from the host rock. This hydration process will be accompanied by variations of the hydro-mechanical behavior of the buffer system. Therefore, it is necessary to study the changing process of the hydro-mechanical behavior of bentonite blocks with consideration of influences of the technological gaps. In the present work, a series of hydration tests were carried out on compacted GMZ bentonite specimens with initial annular gaps. During the hydration, variations of swelling pressure and hydraulic conductivity were measured. After each test, distributions of dry density and water content in the specimen were determined and the microstructure at different positions was investigated. Results show that, the swelling pressure of the specimen increased rapidly first and then decreased slightly during the hydration process, while the hydraulic conductivity kept decreasing with time. Moreover, closure of the technological gaps induced heterogeneity of the bentonite blocks, which was characterized by the decrease of dry density and the increase of water content with increasing distance to the center of the specimen. With the hydration time elapsed, the bentonite blocks could be divided into compression zone and swelling zone along the radius. In the compression zone, the dry density of the specimen gradually increased, with a decrease in the total pore void ratio and the macro-pores void ratio. In the swelling zone, the dry density decreased with an increase of the total pore void ratio and the undetectable pore void ratio.
{"title":"Effect of Hydration Time on the Hydro-Mechanical Behavior of Compacted GMZ Bentonite With an Artificial Annular Gap","authors":"H. Luo, W. Ye, Qiong Wang, Li-Bo Xu","doi":"10.1115/icone29-92182","DOIUrl":"https://doi.org/10.1115/icone29-92182","url":null,"abstract":"\u0000 According to the conceptual design of high-level radioactive waste (HLW) disposal repositories, the buffer barrier will be constructed by compacted bentonite blocks, which will inevitably produce technological gaps (voids) in the repository. During the consequent long-term operation of the repository, the technological gaps will be gradually filled up on hydration of the bentonite blocks with infiltration of underground water from the host rock. This hydration process will be accompanied by variations of the hydro-mechanical behavior of the buffer system. Therefore, it is necessary to study the changing process of the hydro-mechanical behavior of bentonite blocks with consideration of influences of the technological gaps. In the present work, a series of hydration tests were carried out on compacted GMZ bentonite specimens with initial annular gaps. During the hydration, variations of swelling pressure and hydraulic conductivity were measured. After each test, distributions of dry density and water content in the specimen were determined and the microstructure at different positions was investigated. Results show that, the swelling pressure of the specimen increased rapidly first and then decreased slightly during the hydration process, while the hydraulic conductivity kept decreasing with time. Moreover, closure of the technological gaps induced heterogeneity of the bentonite blocks, which was characterized by the decrease of dry density and the increase of water content with increasing distance to the center of the specimen. With the hydration time elapsed, the bentonite blocks could be divided into compression zone and swelling zone along the radius. In the compression zone, the dry density of the specimen gradually increased, with a decrease in the total pore void ratio and the macro-pores void ratio. In the swelling zone, the dry density decreased with an increase of the total pore void ratio and the undetectable pore void ratio.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"57 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127311872","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}
Tritium is one of the most important radioisotopes discharged into the environment in the reactor and contributes to 99% of the total radioactivity of all radionuclides in the liquid discharges during the normal operation of pressurized water reactors. The discharged tritium can be taken into the human body via drinking water. Therefore the minimization of production and discharge of tritium has been highly focused on in the industry and the public across the world. In the paper, all the sources of tritium generation in HPR1000 have been introduced. Activation of Boron-10 in the primary loops and Beryllium from secondary neutron sources have been recognized as the two main sources of tritium in the primary coolant in HPR1000. The annual tritium production and discharge in HPR1000 have been estimated to be about 4.2E+01 TBq/yr using CGN OPEX data with appropriate corrections to reflect tritium-related differences between HPR1000 and the CGN units. To minimize tritium discharges as far as possible, the feasibility of eliminating the secondary neutron source assemblies from the third cycle in the HPR1000 has been verified and the annual tritium can decrease by up to 48% correspondingly. The method used in the study has been certified by UK Generic Design Assessment and EUR compliance assessment. This approach can also be applied to the minimization of tritium discharges in operating PWR units and in the new PWR units.
氚是反应堆内排放到环境中的最重要的放射性同位素之一，在压水堆正常运行过程中，氚占排液中所有放射性核素总放射性的99%。排出的氚可通过饮用水进入人体。因此，减少氚的生产和排放已成为世界各国工业界和公众高度关注的问题。本文介绍了HPR1000中氚生成的所有来源。原回路中的硼-10活化和次级中子源中的铍活化被认为是HPR1000主冷却剂中氚的两个主要来源。根据中广核运营成本数据，HPR1000的年氚产量和排放量估计约为4.2 2e +01 TBq/年，并进行了适当的修正，以反映HPR1000与中广核机组之间的氚相关差异。为了尽可能减少氚的排放，HPR1000在第三次循环中取消二次中子源组件的可行性已经得到验证，相应的，氚的年排放量最多可以减少48%。研究中使用的方法已通过英国通用设计评估和欧盟合规评估认证。这种方法也可以应用于运行中的压水堆机组和新的压水堆机组中氚排放的最小化。
{"title":"Optimization of the Tritium Production and Discharge in HPR1000","authors":"Pengtao Fu, Mingliang Dai","doi":"10.1115/icone29-92816","DOIUrl":"https://doi.org/10.1115/icone29-92816","url":null,"abstract":"\u0000 Tritium is one of the most important radioisotopes discharged into the environment in the reactor and contributes to 99% of the total radioactivity of all radionuclides in the liquid discharges during the normal operation of pressurized water reactors. The discharged tritium can be taken into the human body via drinking water. Therefore the minimization of production and discharge of tritium has been highly focused on in the industry and the public across the world.\u0000 In the paper, all the sources of tritium generation in HPR1000 have been introduced. Activation of Boron-10 in the primary loops and Beryllium from secondary neutron sources have been recognized as the two main sources of tritium in the primary coolant in HPR1000. The annual tritium production and discharge in HPR1000 have been estimated to be about 4.2E+01 TBq/yr using CGN OPEX data with appropriate corrections to reflect tritium-related differences between HPR1000 and the CGN units. To minimize tritium discharges as far as possible, the feasibility of eliminating the secondary neutron source assemblies from the third cycle in the HPR1000 has been verified and the annual tritium can decrease by up to 48% correspondingly. The method used in the study has been certified by UK Generic Design Assessment and EUR compliance assessment. This approach can also be applied to the minimization of tritium discharges in operating PWR units and in the new PWR units.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"113 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124124156","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}
Feng Liu, Ye He, Jiawen Li, Jie Zhan, Xingdong Su, Xuefei Li, Xiajie Liu
Depleted zinc injection into the reactor coolant system (RCS) of pressurized water reactors (PWRs) has demonstrated benefits for significantly reducing radiation fields. Liquid chromatographic separation of zinc isotopes to produce depleted zinc was attempted. A novel chelating resin PS-g-GMA@IDA containing iminodiacetic acid (IDA resins) for separation of Zn2+ and zinc isotope was synthesized by γ-ray induced co-irradiation grafting of glycidyl methacrylate (GMA) onto polystyrene (PS) spherical resin, followed by ring-opening processes with iminodiacetic acid. The obtained chelating resin PS-g-GMA@IDA before and after adsorption of Zn2+ were characterized by FT-IR spectra. This showed that the adsorption of Zn2+ by chelating resin involved two processes. Firstly, zinc ions displaced the hydrogen of the carboxyl group on chelating resin, then Zn2+ chelated with oxygen atom to form a ring structure through the coordination bonds. Adsorption abilities of PS-g-GMA@IDA were evaluated by batch adsorption experiments. The effect of pH and initial concentration on adsorption capacity of PS-g-GMA@IDA were investigated. The highest adsorption capacity 130.1 mg/g for (CH3COO)2Zn was obtained at pH = 6 within 24h when C0 were 200 ppm. EDTA-(NH4)2 was selected as eluent because of its high desorption rate of zinc. A dynamic column adsorption test was carried out at 25 °C using a 1 m column packed with PS-g-GMA@IDA, and the breakthrough occurred at about 500 ml efflunent volume. Column capacity calculated from the breakthrough curve was 56.83 mg/g. Na type PS-g-GMA@IDA was utilized to carry out column chromatographic experiment for zinc isotope separation. Under the conditions of 278±1 K, migration distance 1+9 m and EDTA-(NH4)2 as eluent, the the maximum enrichment value of 66Zn/64Zn and 68Zn/64Zn in the effluent is 1.0268 and 1.0451 respectively.
{"title":"Novel Syntheses Method of Grafted Chelating Resin and Its Application Study for Depleted Zinc Production","authors":"Feng Liu, Ye He, Jiawen Li, Jie Zhan, Xingdong Su, Xuefei Li, Xiajie Liu","doi":"10.1115/icone29-91904","DOIUrl":"https://doi.org/10.1115/icone29-91904","url":null,"abstract":"\u0000 Depleted zinc injection into the reactor coolant system (RCS) of pressurized water reactors (PWRs) has demonstrated benefits for significantly reducing radiation fields. Liquid chromatographic separation of zinc isotopes to produce depleted zinc was attempted. A novel chelating resin PS-g-GMA@IDA containing iminodiacetic acid (IDA resins) for separation of Zn2+ and zinc isotope was synthesized by γ-ray induced co-irradiation grafting of glycidyl methacrylate (GMA) onto polystyrene (PS) spherical resin, followed by ring-opening processes with iminodiacetic acid. The obtained chelating resin PS-g-GMA@IDA before and after adsorption of Zn2+ were characterized by FT-IR spectra. This showed that the adsorption of Zn2+ by chelating resin involved two processes. Firstly, zinc ions displaced the hydrogen of the carboxyl group on chelating resin, then Zn2+ chelated with oxygen atom to form a ring structure through the coordination bonds. Adsorption abilities of PS-g-GMA@IDA were evaluated by batch adsorption experiments. The effect of pH and initial concentration on adsorption capacity of PS-g-GMA@IDA were investigated. The highest adsorption capacity 130.1 mg/g for (CH3COO)2Zn was obtained at pH = 6 within 24h when C0 were 200 ppm. EDTA-(NH4)2 was selected as eluent because of its high desorption rate of zinc. A dynamic column adsorption test was carried out at 25 °C using a 1 m column packed with PS-g-GMA@IDA, and the breakthrough occurred at about 500 ml efflunent volume. Column capacity calculated from the breakthrough curve was 56.83 mg/g. Na type PS-g-GMA@IDA was utilized to carry out column chromatographic experiment for zinc isotope separation. Under the conditions of 278±1 K, migration distance 1+9 m and EDTA-(NH4)2 as eluent, the the maximum enrichment value of 66Zn/64Zn and 68Zn/64Zn in the effluent is 1.0268 and 1.0451 respectively.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"128 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121467567","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}
W. Xie, Jia Huang, Chenyu Shan, Qin Lei, K. Chang, Hao Zhou, Feng Liu, Peng Lin, Xiajie Liu, L. Li
High Integrity Container (HIC) for nuclear power plant is an advanced waste reconditioning Container specially designed and manufactured. It has the characteristics of High strength, good sealing, strong chemical stability and thermal stability. It can be used to load a variety of Low and Intermediate Level radioactive wastes that have not been cured or fixed, such as dehydrated mud and evaporation residues, Waste resins and Waste water filtration cores. According to relevant national standards and nuclear power plant radioactive waste management regulations, it is necessary to measure the type of radionuclide and radioactive level of waste barrels. The sizes of HIC barrels are DI, DII and DIII, with volumes ranging from 0.42 m3 to 3.44 m3. There are various types of waste in the bucket, and the waste density is not less than 0.25 g·cm−3; The main material of HIC barrel is high density crosslinked polyethylene, which has poor shielding effect on γ-rays, resulting in high dose rate level on the surface of the barrel. Therefore, the existing γ scanning device for 200 L/400 L metal waste drums in nuclear power plants cannot meet the measurement requirements of HIC drums, and it is necessary to develop a special γ scanning device for HIC drums. In this study, Cadmium Zinc Telluride (CZT) semiconductor was used as the main detector to simulate the detection efficiency under different HIC barrel sizes, different waste density, different detection distances and different collimating opening diameters. The simulation results show that the CZT detector can be used for gamma scanning measurement of HIC drums, especially HIC drums with high dose rate, under the condition of proper shielding collimation and detection distance.
{"title":"Simulation of Detection Efficiency of High Integrity Container Gamma Scanning Device for Nuclear Power Plant Based on CZT Detector","authors":"W. Xie, Jia Huang, Chenyu Shan, Qin Lei, K. Chang, Hao Zhou, Feng Liu, Peng Lin, Xiajie Liu, L. Li","doi":"10.1115/icone29-91233","DOIUrl":"https://doi.org/10.1115/icone29-91233","url":null,"abstract":"\u0000 High Integrity Container (HIC) for nuclear power plant is an advanced waste reconditioning Container specially designed and manufactured. It has the characteristics of High strength, good sealing, strong chemical stability and thermal stability. It can be used to load a variety of Low and Intermediate Level radioactive wastes that have not been cured or fixed, such as dehydrated mud and evaporation residues, Waste resins and Waste water filtration cores. According to relevant national standards and nuclear power plant radioactive waste management regulations, it is necessary to measure the type of radionuclide and radioactive level of waste barrels. The sizes of HIC barrels are DI, DII and DIII, with volumes ranging from 0.42 m3 to 3.44 m3. There are various types of waste in the bucket, and the waste density is not less than 0.25 g·cm−3; The main material of HIC barrel is high density crosslinked polyethylene, which has poor shielding effect on γ-rays, resulting in high dose rate level on the surface of the barrel. Therefore, the existing γ scanning device for 200 L/400 L metal waste drums in nuclear power plants cannot meet the measurement requirements of HIC drums, and it is necessary to develop a special γ scanning device for HIC drums. In this study, Cadmium Zinc Telluride (CZT) semiconductor was used as the main detector to simulate the detection efficiency under different HIC barrel sizes, different waste density, different detection distances and different collimating opening diameters. The simulation results show that the CZT detector can be used for gamma scanning measurement of HIC drums, especially HIC drums with high dose rate, under the condition of proper shielding collimation and detection distance.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"108 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133999957","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}
Recently, gallium oxide (Ga2O3) as an ultra-wide bandgap oxide semiconductor has aroused enormous attention in research and development due to its prospects for future power electronic, optoelectronic, and radiation detection applications. Ga2O3-based devices could be subject to fluxes of protons or neutrons if used in aerospace or radiation-hard nuclear systems, which leads to internal defects in Ga2O3 crystals and the degradation of device performance. Therefore, the radiation effects of Ga2O3 under irradiation have become a research focus, and it is of great significance to study the defect behavior and performance influence of Ga2O3 after irradiation. In this paper, the number of displacement defects produced by 1∼20 MeV neutrons in Ga2O3 were calculated using Geant4 simulations, and the factors that impact the displacement damage were studied. The results show that the displacement defects generated by neutrons do not increase monotonously with neutron energy but are closely related to the reaction cross-section and the generation of (PKA). We simulated and calculated the radiation damage by 10∼100keV protons in Ga2O3 using SRIM. It is found that ionization damage is much greater than displacement damage; the number of vacancies generated by proton radiation in Ga2O3 increases with the energy and incident angle of the incident proton. The irradiation resistance of Ga2O3 is between silicon and diamond semiconductor materials.
{"title":"Simulation of Neutron and Proton Displacement Damage in Ultra-Wide Bandgap Semiconductor Ga2O3","authors":"Z. Shao, Ziqi Cai, Qingmin Zhang","doi":"10.1115/icone29-92600","DOIUrl":"https://doi.org/10.1115/icone29-92600","url":null,"abstract":"\u0000 Recently, gallium oxide (Ga2O3) as an ultra-wide bandgap oxide semiconductor has aroused enormous attention in research and development due to its prospects for future power electronic, optoelectronic, and radiation detection applications. Ga2O3-based devices could be subject to fluxes of protons or neutrons if used in aerospace or radiation-hard nuclear systems, which leads to internal defects in Ga2O3 crystals and the degradation of device performance. Therefore, the radiation effects of Ga2O3 under irradiation have become a research focus, and it is of great significance to study the defect behavior and performance influence of Ga2O3 after irradiation. In this paper, the number of displacement defects produced by 1∼20 MeV neutrons in Ga2O3 were calculated using Geant4 simulations, and the factors that impact the displacement damage were studied. The results show that the displacement defects generated by neutrons do not increase monotonously with neutron energy but are closely related to the reaction cross-section and the generation of (PKA). We simulated and calculated the radiation damage by 10∼100keV protons in Ga2O3 using SRIM. It is found that ionization damage is much greater than displacement damage; the number of vacancies generated by proton radiation in Ga2O3 increases with the energy and incident angle of the incident proton. The irradiation resistance of Ga2O3 is between silicon and diamond semiconductor materials.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"27 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133738971","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}
Xiaojun Yan, Xiliang Guo, K. Gao, Xiaobin Guo, Yahui Xi
In original cement curing formula, curing base material adopts the cement from the specific manufacturer. This increases the difficulty and cost of material procurement and storage in different nuclear power plants, and further leads to the waste of materials. In addition, the leaching resistance of the original formula is poor, and the cumulative leaching fraction of 137Cs is close to the values specified by the national standard. In order to make the practical work more convenient and the disposal of radioactive cement waste safer, the formulation should be improved. In this work, the cement plants around nuclear power plant were investigated, and the cement was selected as the raw material based on the requirements of the nuclear power cement solidifying formula. The coagulant calcium chloride was added to the formula to improve the delay of the concentrated solution formula. Molecular sieve was added to improve the leaching resistance of the original formula. After adjusting the proportion and composition of the formula, inclusion rate of concentrate and the waste inclusion rate for the new formula increased from 51% to 56% and from 35% to 45%, separately. Meanwhile, the compressive strength of cement waste was enhanced up to 10%. The formulation cost reduced, as calculated of the wet waste generated by two megawatt units, the solidified waste disposal cost could be saved as 2.2 million RMB per year. The ability of resisting leaching radionuclides was improved. Specifically, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the concentrated solution formula was reduced by 25.77% and 56.91%, respectively. In addition, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the resin formula was reduced by 82.80% and 82.31%, respectively.
{"title":"Study on Improvement of Cement Curing Formula of Wet Radioactive Waste","authors":"Xiaojun Yan, Xiliang Guo, K. Gao, Xiaobin Guo, Yahui Xi","doi":"10.1115/icone29-92404","DOIUrl":"https://doi.org/10.1115/icone29-92404","url":null,"abstract":"\u0000 In original cement curing formula, curing base material adopts the cement from the specific manufacturer. This increases the difficulty and cost of material procurement and storage in different nuclear power plants, and further leads to the waste of materials. In addition, the leaching resistance of the original formula is poor, and the cumulative leaching fraction of 137Cs is close to the values specified by the national standard. In order to make the practical work more convenient and the disposal of radioactive cement waste safer, the formulation should be improved. In this work, the cement plants around nuclear power plant were investigated, and the cement was selected as the raw material based on the requirements of the nuclear power cement solidifying formula. The coagulant calcium chloride was added to the formula to improve the delay of the concentrated solution formula. Molecular sieve was added to improve the leaching resistance of the original formula. After adjusting the proportion and composition of the formula, inclusion rate of concentrate and the waste inclusion rate for the new formula increased from 51% to 56% and from 35% to 45%, separately. Meanwhile, the compressive strength of cement waste was enhanced up to 10%. The formulation cost reduced, as calculated of the wet waste generated by two megawatt units, the solidified waste disposal cost could be saved as 2.2 million RMB per year. The ability of resisting leaching radionuclides was improved. Specifically, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the concentrated solution formula was reduced by 25.77% and 56.91%, respectively. In addition, compared with the original formula, the leaching rate and cumulative leaching fraction of 137Cs in the resin formula was reduced by 82.80% and 82.31%, respectively.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"56 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133006971","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}
Yingzhe Du, Lili Li, Kunfeng Li, Peng Lin, Yugang Zhang, Juan Diwu
Actinides have complex chemical properties, there are a variety of oxidation states and changeable coordination configurations, and their physical properties, electronic structure and molecular orbital theory are much more complex than other elements. On the other hand, actinides in high level liquid waste, especially transuranic elements synthesized by artificial nuclear reaction, are highly radioactive, which is critical for its stable glass solidification. At present, taking elements with similar properties as substitutes to simulate actinides and inferring the properties of actinides by analyzing the compounds formed by substitutes is one of the important works of glass solidification. This paper provides a possible way to use trivalent lanthanides as an alternative for trivalent actinides.
{"title":"Study on the Substitute Nuclides of Actinides in High-Level Radioactive Waste Liquid","authors":"Yingzhe Du, Lili Li, Kunfeng Li, Peng Lin, Yugang Zhang, Juan Diwu","doi":"10.1115/icone29-93771","DOIUrl":"https://doi.org/10.1115/icone29-93771","url":null,"abstract":"\u0000 Actinides have complex chemical properties, there are a variety of oxidation states and changeable coordination configurations, and their physical properties, electronic structure and molecular orbital theory are much more complex than other elements. On the other hand, actinides in high level liquid waste, especially transuranic elements synthesized by artificial nuclear reaction, are highly radioactive, which is critical for its stable glass solidification. At present, taking elements with similar properties as substitutes to simulate actinides and inferring the properties of actinides by analyzing the compounds formed by substitutes is one of the important works of glass solidification. This paper provides a possible way to use trivalent lanthanides as an alternative for trivalent actinides.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126874108","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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