Activated corrosion products are the dominate radiation source during PWR maintenance. Co-58 and Co-60 are two important radionuclides due to the high energy emitting gamma rays and high radio activities in the primary circuit. They have contributed to more than 90% occupational radiation exposure during maintenance according to operating experiences from PWRs. Therefore, it is essential to research the behaviour of the two radionuclides. The Co-58 and Co-60 coolant activity for both short-term tendency and long-term tendency are analyzed in this paper. Operating Experience (OPEX) from CPR1000 PWRs show that the Co-58 and Co-60 coolant activity in a fuel cycle usually result in shape of a bowl and the highest activities are at the beginning of the cycle which is consistent with the shape of the core boiling rate. Namely, fuel crud behavior plays a leading role in the formation of activated corrosion products in the primary circuit. For Co-58 and Co-60 coolant activity in long-term period, it is influenced mainly by the corrosion/release rate and the radionuclides’ half-lives. OPEX combined with simulation by CAMPSIS code show that Co-58 coolant activity reach to maximum value at second to third cycle while Co-60 coolant activity reach to maximum value at fifth to ninth cycle. Research on time-dependent Co-58 and Co-60 coolant activities give a better understanding of the radionuclide behaviour as well as provide a basis for developing corrosion product simulating model.
{"title":"Study on Time-Dependent Co-58 and Co-60 Activities in the Primary Coolant of CPR1000 PWRs","authors":"Xiaoqian Zhang, Pengtao Fu","doi":"10.1115/icone29-92822","DOIUrl":"https://doi.org/10.1115/icone29-92822","url":null,"abstract":"\u0000 Activated corrosion products are the dominate radiation source during PWR maintenance. Co-58 and Co-60 are two important radionuclides due to the high energy emitting gamma rays and high radio activities in the primary circuit. They have contributed to more than 90% occupational radiation exposure during maintenance according to operating experiences from PWRs. Therefore, it is essential to research the behaviour of the two radionuclides.\u0000 The Co-58 and Co-60 coolant activity for both short-term tendency and long-term tendency are analyzed in this paper. Operating Experience (OPEX) from CPR1000 PWRs show that the Co-58 and Co-60 coolant activity in a fuel cycle usually result in shape of a bowl and the highest activities are at the beginning of the cycle which is consistent with the shape of the core boiling rate. Namely, fuel crud behavior plays a leading role in the formation of activated corrosion products in the primary circuit. For Co-58 and Co-60 coolant activity in long-term period, it is influenced mainly by the corrosion/release rate and the radionuclides’ half-lives. OPEX combined with simulation by CAMPSIS code show that Co-58 coolant activity reach to maximum value at second to third cycle while Co-60 coolant activity reach to maximum value at fifth to ninth cycle. Research on time-dependent Co-58 and Co-60 coolant activities give a better understanding of the radionuclide behaviour as well as provide a basis for developing corrosion product simulating model.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"18 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":"114933735","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}
Jiawen Li, Mingyi Chen, L. Li, Feng Liu, Yufei Gao, Jian Zhu, Jie Zhan, Jian Chen, Y. Zeng, Rouxi Chen, Hsing‐Lin Wang
Radioactive aerosols are highly hazardous aerosols containing radionuclides. Inhalation of radioactive aerosols can lead to serious internal exposure hazards to human body. If discharged without proper treatment, it will also harm the working environment and even the ecological environment. Therefore, radioactive aerosol protection is a significant part of environmental protection and personnel protection in the nuclear field. However, majority of existing protective fabrics for radioactive aerosol filtration always meet the trade-off among filtration capacity, mechanical properties and air permeability. In this study, nanofiber layers were prepared by electrospinning technology using TPU, PVDF, PVA polymer materials and electret materials SiO2 as spinning materials. Composite membranes, prepared by coating different nanofiber layers on the PET non-woven fabrics substrate, were investigated. The results show that the 12wt% TPU nanofiber membrane has a three-dimensional spatial hierarchical structure. Its ultra-fine fiber diameter with small pore size greatly enhances the PM capture ability (PM0.3 filtration efficiency 99.99%); and the beaded spatial structure is beneficial to reduce the air resistance to 299 Pa (flow rate 95 L/min). Meanwhile, TPU nanofiber membrane has high extensibility, and it is superior to PVDF and PVA composite membranes in mechanical properties after thermal compounding. Appropriate content of SiO2 can improve filtration performance. The study shed light on developing electrospun nanofiber for radioactive aerosol protection, which can be used in the purification of ambient air in nuclear facilities, or as a high-performance fabric for radioactive aerosol protective clothing.
{"title":"Preparation and Characterization of Radioactive Aerosol Protective Nanofiber Membranes","authors":"Jiawen Li, Mingyi Chen, L. Li, Feng Liu, Yufei Gao, Jian Zhu, Jie Zhan, Jian Chen, Y. Zeng, Rouxi Chen, Hsing‐Lin Wang","doi":"10.1115/icone29-92665","DOIUrl":"https://doi.org/10.1115/icone29-92665","url":null,"abstract":"\u0000 Radioactive aerosols are highly hazardous aerosols containing radionuclides. Inhalation of radioactive aerosols can lead to serious internal exposure hazards to human body. If discharged without proper treatment, it will also harm the working environment and even the ecological environment. Therefore, radioactive aerosol protection is a significant part of environmental protection and personnel protection in the nuclear field. However, majority of existing protective fabrics for radioactive aerosol filtration always meet the trade-off among filtration capacity, mechanical properties and air permeability. In this study, nanofiber layers were prepared by electrospinning technology using TPU, PVDF, PVA polymer materials and electret materials SiO2 as spinning materials. Composite membranes, prepared by coating different nanofiber layers on the PET non-woven fabrics substrate, were investigated. The results show that the 12wt% TPU nanofiber membrane has a three-dimensional spatial hierarchical structure. Its ultra-fine fiber diameter with small pore size greatly enhances the PM capture ability (PM0.3 filtration efficiency 99.99%); and the beaded spatial structure is beneficial to reduce the air resistance to 299 Pa (flow rate 95 L/min). Meanwhile, TPU nanofiber membrane has high extensibility, and it is superior to PVDF and PVA composite membranes in mechanical properties after thermal compounding. Appropriate content of SiO2 can improve filtration performance. The study shed light on developing electrospun nanofiber for radioactive aerosol protection, which can be used in the purification of ambient air in nuclear facilities, or as a high-performance fabric for radioactive aerosol protective clothing.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"15 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":"129731124","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}
For a nuclear power plant, the radioactive gaseous and liquid discharges are the main contributor to radiation exposure to the member of public and non-human biota during normal operation, which hence need to be quantified and to support the environmental impact assessment. When applying the traditional theoretical methodology, due to various and complex mechanisms involved in radioactive gaseous and liquid effluent streams, a number of assumptions need to be made to support the theoretical modeling. The combination of these assumptions can easily lead to overestimate or underestimate of the radioactive discharges and limits and may not represent the actual performance of the plants. As such, to obtain predicted discharges and limits closer to the future actual performance of the plant, it is meaningful and necessary to develop a methodology based on operating experience. This paper has studied and developed a systematic methodology based on operating experience for quantification of radioactive discharges and limits for the 3rd generation pressurized water reactor HPR1000 during normal operation, taking into account the differences on design features and operation management between the HPR1000 and the operating units, the fluctuations due to the variations of plant and system operation parameters and the potential influences from expected events within the normal operation range. This methodology has been successfully applied to HPR1000 and the results have been verified reasonable and appropriate by comparing with the operating experience data from comparable international PWRs. This methodology has been applied to HPR1000 successfully for Generic Design Assessment (GDA) in the UK and the European Utility Requirements for LWR Nuclear Power Plants (EUR) and can also be widely applied for other PWRs with slight adjustment.
{"title":"Study on Methodology for Quantification of Radioactive Discharges and Limits for Pressurized Water Reactor HPR1000 Based on Operating Experience","authors":"Yujia Chen, Weifeng Lv, Zhenyu Jiang, Yongtao Zhou","doi":"10.1115/icone29-92385","DOIUrl":"https://doi.org/10.1115/icone29-92385","url":null,"abstract":"\u0000 For a nuclear power plant, the radioactive gaseous and liquid discharges are the main contributor to radiation exposure to the member of public and non-human biota during normal operation, which hence need to be quantified and to support the environmental impact assessment.\u0000 When applying the traditional theoretical methodology, due to various and complex mechanisms involved in radioactive gaseous and liquid effluent streams, a number of assumptions need to be made to support the theoretical modeling. The combination of these assumptions can easily lead to overestimate or underestimate of the radioactive discharges and limits and may not represent the actual performance of the plants. As such, to obtain predicted discharges and limits closer to the future actual performance of the plant, it is meaningful and necessary to develop a methodology based on operating experience.\u0000 This paper has studied and developed a systematic methodology based on operating experience for quantification of radioactive discharges and limits for the 3rd generation pressurized water reactor HPR1000 during normal operation, taking into account the differences on design features and operation management between the HPR1000 and the operating units, the fluctuations due to the variations of plant and system operation parameters and the potential influences from expected events within the normal operation range. This methodology has been successfully applied to HPR1000 and the results have been verified reasonable and appropriate by comparing with the operating experience data from comparable international PWRs.\u0000 This methodology has been applied to HPR1000 successfully for Generic Design Assessment (GDA) in the UK and the European Utility Requirements for LWR Nuclear Power Plants (EUR) and can also be widely applied for other PWRs with slight adjustment.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"1 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":"116276555","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}
Liu Yitang, S. Rui, Wang Zhou, Wang Zhenchuan, Lan Yunliang, Zhao Wei, T. Xianguo
Segmented Gamma Scanning (SGS) is a kind of nondestructive testing (NDT) techniques, widely used in online detection for nuclear waste drums to record the types and contents of radionuclides in the nuclear wastes. It is convenient to classify and dispose of nuclear waste according to the test results, and it also avoids radiation damage to inspectors from the destructive analysis. A new software of the radioactive analysis and monitoring system for nuclear waste barrels with SGS was developed in this work, which is mainly composed of two parts including control upper computer software and radionuclide analysis software, in which control software contained motion control and Multi-channel Analyzer (MCA) control. The controlling of the mechanical platform realized the rotation of the waste drums and the synchronous lifting and lowering of the transmission source and the detector, so as to facilitate the layered scanning of the drums. The motion control is an indispensable part of the detection system, whose precision of the movement is directly affected the accuracy of the detection results. In the radioactive measurement, a high purity germanium (HPGe) gamma-ray spectrometer was used to obtain the gamma-ray spectrum, in which the MCA was responsible for the control of spectrometer and gamma-ray spectrum data record. Therefore, the MCA control part must adjust the high voltage of the HPGe and the parameters of measurement. The gamma-ray spectrum contained the radioactive information of the nuclear wastes in the drums. The analysis of the radioactive data is the core of the software, including spectrum data resolving and the radioactive reconstructed of the radionuclides in the drums. Finally, the information such as the type and activity of the radionuclides in the barrels was provided to the user. The software was written with the C# programming language, which realized the accurate control and operation of the mechanical device and the orderly performed of motion detection. To establish communication, the software used Ethernet’s TCP/IP as the control network, in which the manual mode and auto mode were alternatives. In conclusion, the software promotes the coordination and integration of motion control, MCA control and gamma-ray spectrum data analysis in the process of automatic detection of barreled nuclear waste with SGS.
{"title":"Software of Radioactivity Measurement and Control System for Barreled Nuclear Waste With Segmented Gamma Scanning","authors":"Liu Yitang, S. Rui, Wang Zhou, Wang Zhenchuan, Lan Yunliang, Zhao Wei, T. Xianguo","doi":"10.1115/icone29-93603","DOIUrl":"https://doi.org/10.1115/icone29-93603","url":null,"abstract":"\u0000 Segmented Gamma Scanning (SGS) is a kind of nondestructive testing (NDT) techniques, widely used in online detection for nuclear waste drums to record the types and contents of radionuclides in the nuclear wastes. It is convenient to classify and dispose of nuclear waste according to the test results, and it also avoids radiation damage to inspectors from the destructive analysis. A new software of the radioactive analysis and monitoring system for nuclear waste barrels with SGS was developed in this work, which is mainly composed of two parts including control upper computer software and radionuclide analysis software, in which control software contained motion control and Multi-channel Analyzer (MCA) control. The controlling of the mechanical platform realized the rotation of the waste drums and the synchronous lifting and lowering of the transmission source and the detector, so as to facilitate the layered scanning of the drums. The motion control is an indispensable part of the detection system, whose precision of the movement is directly affected the accuracy of the detection results. In the radioactive measurement, a high purity germanium (HPGe) gamma-ray spectrometer was used to obtain the gamma-ray spectrum, in which the MCA was responsible for the control of spectrometer and gamma-ray spectrum data record. Therefore, the MCA control part must adjust the high voltage of the HPGe and the parameters of measurement. The gamma-ray spectrum contained the radioactive information of the nuclear wastes in the drums. The analysis of the radioactive data is the core of the software, including spectrum data resolving and the radioactive reconstructed of the radionuclides in the drums. Finally, the information such as the type and activity of the radionuclides in the barrels was provided to the user. The software was written with the C# programming language, which realized the accurate control and operation of the mechanical device and the orderly performed of motion detection. To establish communication, the software used Ethernet’s TCP/IP as the control network, in which the manual mode and auto mode were alternatives. In conclusion, the software promotes the coordination and integration of motion control, MCA control and gamma-ray spectrum data analysis in the process of automatic detection of barreled nuclear waste with SGS.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"11 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":"115977098","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}
Rock cavity disposal is regarded as a preferred option of radioactive waste final disposal solutions with less disturbance to surface and better isolation from people and surface environment which facilitates the long term surveillance and control of facility both during operation and after closure. This paper provides the research progress of key aspects of rock cavity disposal technology, such as site selection criteria, engineering design scheme of disposal vaults, technique process design of the handling of waste packages, research & development of intelligent control system and equipment of waste transportation, documentation of safety case and supporting safety assessment, numerical simulation of radionuclides migration process, and etc. The first practice of rock cavity disposal in China is also discussed. It shows that developing the rock cavity technology is reasonable, feasible and has significant potential in China. Finally, this paper looks forward to the research focus in the next stage. It is suggested to study the compatibility between the complexity level of engineering barrier measures and the waste hazards and to improve the reliability of future evolution and scenario assumption. The nuclides migration behavior study and migration parameter determination involving more key nuclides in waste from other origins are also of importance.
{"title":"Research and Development of Rock Cavity Disposal Technology in China","authors":"Wencheng Yin, Yu-Djai. Pan, Xueling Zhang, Yu Liu, Jian Ma, Tongtong Li","doi":"10.1115/icone29-91843","DOIUrl":"https://doi.org/10.1115/icone29-91843","url":null,"abstract":"\u0000 Rock cavity disposal is regarded as a preferred option of radioactive waste final disposal solutions with less disturbance to surface and better isolation from people and surface environment which facilitates the long term surveillance and control of facility both during operation and after closure. This paper provides the research progress of key aspects of rock cavity disposal technology, such as site selection criteria, engineering design scheme of disposal vaults, technique process design of the handling of waste packages, research & development of intelligent control system and equipment of waste transportation, documentation of safety case and supporting safety assessment, numerical simulation of radionuclides migration process, and etc. The first practice of rock cavity disposal in China is also discussed. It shows that developing the rock cavity technology is reasonable, feasible and has significant potential in China. Finally, this paper looks forward to the research focus in the next stage. It is suggested to study the compatibility between the complexity level of engineering barrier measures and the waste hazards and to improve the reliability of future evolution and scenario assumption. The nuclides migration behavior study and migration parameter determination involving more key nuclides in waste from other origins are also of importance.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"05 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":"127070718","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}
In abroad, the treatment method adopted for the wastes above is mechanical compression and storage [1–2]. In domestic, the wastes management methods could be summarized as follows according to the surface dose of the waste substance: (1) Waste filter cores with a surface dose rate greater than 2 mSv/h are fixed in a 400L steel drum with cement in the cement solidified line of nuclear auxiliary plant (NX), and one waste filter core is installed in a 400L drum [3–5]. After reconditioning, the waste volume is 0.4m3; By storing decay, the surface dose rate of some of the high dose rate level waste filter cores decreased to below 2 mSv/h, and then they were dried, super-compacted and cemented, which greatly reduced the amount of waste generated. (2) The waste filter elements with the surface dose rate below 2 mSv/h are packed into 200L steel drums and sent to the waste treatment auxiliary workshop (QS) for drying, super compaction and cement fixation. About 3∼4 waste filter elements are packed into a 400L drum. After preparation, the average waste volume of each filter core is 0.1m3. However, the treatment time of these methods is very long, the volume of waste after treatment is still relatively large, and the storage and isolation time is long. To sum up, this paper innovatively adopted high-temperature melting method to conduct glass solidification treatment on the simulated contaminated glass fiber, and by this method it forms a stable glass body with stable chemical properties. Moreover, it reduces the waste volume and directly forms a solidified body which is more convenient for treatment and disposal.
{"title":"Study on the High Temperature Melting Treatment of Nuclear Waste Glass Fiber","authors":"Chunyu Liu, Yan Wen","doi":"10.1115/icone29-93802","DOIUrl":"https://doi.org/10.1115/icone29-93802","url":null,"abstract":"\u0000 In abroad, the treatment method adopted for the wastes above is mechanical compression and storage [1–2]. In domestic, the wastes management methods could be summarized as follows according to the surface dose of the waste substance: (1) Waste filter cores with a surface dose rate greater than 2 mSv/h are fixed in a 400L steel drum with cement in the cement solidified line of nuclear auxiliary plant (NX), and one waste filter core is installed in a 400L drum [3–5]. After reconditioning, the waste volume is 0.4m3; By storing decay, the surface dose rate of some of the high dose rate level waste filter cores decreased to below 2 mSv/h, and then they were dried, super-compacted and cemented, which greatly reduced the amount of waste generated. (2) The waste filter elements with the surface dose rate below 2 mSv/h are packed into 200L steel drums and sent to the waste treatment auxiliary workshop (QS) for drying, super compaction and cement fixation. About 3∼4 waste filter elements are packed into a 400L drum. After preparation, the average waste volume of each filter core is 0.1m3. However, the treatment time of these methods is very long, the volume of waste after treatment is still relatively large, and the storage and isolation time is long. To sum up, this paper innovatively adopted high-temperature melting method to conduct glass solidification treatment on the simulated contaminated glass fiber, and by this method it forms a stable glass body with stable chemical properties. Moreover, it reduces the waste volume and directly forms a solidified body which is more convenient for treatment and disposal.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"29 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":"115002573","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}
Hongji Sang, Zhengcheng Gu, Zheng Cui, Ruoxue Zou, Yan Wu
137Cs and 90Sr in high level liquid waste, with high radioactivity, large heat-generating and relatively long half-life. In order to solve the problem of final disposal of 137Cs and 90Sr, natural mineral allophane was chosen as the base materials to synthesize the silicate ceramic solidified products with cold pressing/sintering method. The microstructure, phase composition and surface element distribution of the solidified products were analyzed. The solidification mechanism of the solidified products was also discussed. The surface of the solidified products appeared melting phenomenon after sintering, and the structure was more compact. CsAlSiO4, Sr2Al2SiO7 and SrAl2Si2O8, which can stably solidify Cs and Sr, were formed in the solidified products. The content of allophane in the solidified product has an important influence on the immobilization ratio of Cs and Sr. And at the sintering condition of 1 h duration time at 1200 °C, the immobilization ratios of Cs and Sr can reach 100 %. By increasing the content of cured substrate, the surface characteristics and crystallization properties of the solidified product can be improved, and the volume reduction effect was more obvious.
{"title":"Preparation and Properties of Ceramic Solidified Product Containing Cs and Sr","authors":"Hongji Sang, Zhengcheng Gu, Zheng Cui, Ruoxue Zou, Yan Wu","doi":"10.1115/icone29-92765","DOIUrl":"https://doi.org/10.1115/icone29-92765","url":null,"abstract":"\u0000 137Cs and 90Sr in high level liquid waste, with high radioactivity, large heat-generating and relatively long half-life. In order to solve the problem of final disposal of 137Cs and 90Sr, natural mineral allophane was chosen as the base materials to synthesize the silicate ceramic solidified products with cold pressing/sintering method. The microstructure, phase composition and surface element distribution of the solidified products were analyzed. The solidification mechanism of the solidified products was also discussed. The surface of the solidified products appeared melting phenomenon after sintering, and the structure was more compact. CsAlSiO4, Sr2Al2SiO7 and SrAl2Si2O8, which can stably solidify Cs and Sr, were formed in the solidified products. The content of allophane in the solidified product has an important influence on the immobilization ratio of Cs and Sr. And at the sintering condition of 1 h duration time at 1200 °C, the immobilization ratios of Cs and Sr can reach 100 %. By increasing the content of cured substrate, the surface characteristics and crystallization properties of the solidified product can be improved, and the volume reduction effect was more obvious.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"6 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":"134511044","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}
Jin Lu, Peng Nie, Ren Ren, Yaping Guo, Qi Wang, Xinwang Zhang, Lijun Zhang
Heavy Water Research Reactor was the first reactor in China. It adopted metallic uranium as fuel and aluminum alloy as cladding in the beginning. During the refueling procedure many years ago, several fuel pins dropped at the bottom of the spent fuel pool. The amount and the location of the fuel pins were not recorded. Before decommissioning, it is important to locate and take back the fuel pins. The fuel pin is φ40 mm × 100 mm and the pool is 17.8 m × 5.6 m. It is like dredging for a needle in the sea. What’s worse, it is not clear to find the fuel pins even with a waterproof camera for the bad water quality. Besides, radiation tubes, neutron detectors, and broken control rod were also stored in the pool.After reviewing on the localization method of radioactive source at home and abroad, it is determined to use γ dose rate meter and γ spectrometer in this case. Shields in different width and thickness were calculated with MNCP code. The optimization was 5cm in thickness and 5 cm width lead shield with a hole in the center. The waterproof camera was tied with Gama detector and finally 14 fuel pins were located and taken back safely.
重水研究堆是中国第一座反应堆。它最初采用金属铀作为燃料，铝合金作为包层。在多年前的换料过程中，几个燃料销掉在乏燃料池的底部。燃料销的数量和位置没有记录。在退役之前，重要的是要找到并收回燃料销。燃料销φ40 mm × 100 mm，燃料池17.8 m × 5.6 m。这就像大海捞针。更糟糕的是，由于水质差，即使有防水相机也不清楚找到燃料针。此外，池中还存放了辐射管、中子探测器和破损的控制棒。在对国内外放射源定位方法进行综述后，确定在这种情况下使用γ剂量率计和γ能谱仪。用MNCP程序计算不同宽度和厚度的盾构。优化设计为厚度为5cm，宽度为5cm，中间有孔的铅屏蔽。防水相机与伽马探测器绑在一起，最终找到14个燃料针并安全带回。
{"title":"The Location of Spent Fuel Pins in Heavy Water Research Reactor in the Spent Fuel Pool","authors":"Jin Lu, Peng Nie, Ren Ren, Yaping Guo, Qi Wang, Xinwang Zhang, Lijun Zhang","doi":"10.1115/icone29-91465","DOIUrl":"https://doi.org/10.1115/icone29-91465","url":null,"abstract":"\u0000 Heavy Water Research Reactor was the first reactor in China. It adopted metallic uranium as fuel and aluminum alloy as cladding in the beginning. During the refueling procedure many years ago, several fuel pins dropped at the bottom of the spent fuel pool. The amount and the location of the fuel pins were not recorded. Before decommissioning, it is important to locate and take back the fuel pins. The fuel pin is φ40 mm × 100 mm and the pool is 17.8 m × 5.6 m. It is like dredging for a needle in the sea. What’s worse, it is not clear to find the fuel pins even with a waterproof camera for the bad water quality. Besides, radiation tubes, neutron detectors, and broken control rod were also stored in the pool.After reviewing on the localization method of radioactive source at home and abroad, it is determined to use γ dose rate meter and γ spectrometer in this case. Shields in different width and thickness were calculated with MNCP code. The optimization was 5cm in thickness and 5 cm width lead shield with a hole in the center. The waterproof camera was tied with Gama detector and finally 14 fuel pins were located and taken back safely.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"7 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":"122712306","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}
Resources and environment are two crucial (key) elements for the sustainable development of human society. This paper aims to explore a new glass-ceramic system for the development of high-level radioactive solid-waste vitrification technology, clarify the function of multielemental radionuclides (heterogeneous ion) on the structural and chemical durability characteristics. By changing the crystallization temperature and crystallization time, the heat treatment system of the zirconolite-based glass ceramic waste form was optimized; the K value method was used to obtain the quantitative analysis of the crystal phase; CeO2 was used as a simulated nuclide to explore the effect of CeO2 doping on the phase. The influence of composition, and the valence distribution of cerium in the glass ceramics was analyzed by XPS. The results show that the nucleation temperature is 810 °C for 2 h, and then the crystallization temperature is 950 or 1000 °C for 2 h, the glass ceramics containing only the 2M zirconolite and residual glass. CeO2 doping will lead to the increase of the lattice constant. The ratio of Ce4+ to Ce3+ in the waste form with different CeO2 content is maintained at 6:4. The normalized leaching rate of the sample with the highest doping content is 5.949 × 10−6 g·m−2·d −1.
{"title":"Structure Control of a HLW Immobilized Zirconolite Glass-Ceramic Matrix","authors":"Haiqing Li, Yutong Pan, Chao Gao, Shuming Wang","doi":"10.1115/icone29-93212","DOIUrl":"https://doi.org/10.1115/icone29-93212","url":null,"abstract":"\u0000 Resources and environment are two crucial (key) elements for the sustainable development of human society. This paper aims to explore a new glass-ceramic system for the development of high-level radioactive solid-waste vitrification technology, clarify the function of multielemental radionuclides (heterogeneous ion) on the structural and chemical durability characteristics.\u0000 By changing the crystallization temperature and crystallization time, the heat treatment system of the zirconolite-based glass ceramic waste form was optimized; the K value method was used to obtain the quantitative analysis of the crystal phase; CeO2 was used as a simulated nuclide to explore the effect of CeO2 doping on the phase. The influence of composition, and the valence distribution of cerium in the glass ceramics was analyzed by XPS. The results show that the nucleation temperature is 810 °C for 2 h, and then the crystallization temperature is 950 or 1000 °C for 2 h, the glass ceramics containing only the 2M zirconolite and residual glass. CeO2 doping will lead to the increase of the lattice constant. The ratio of Ce4+ to Ce3+ in the waste form with different CeO2 content is maintained at 6:4. The normalized leaching rate of the sample with the highest doping content is 5.949 × 10−6 g·m−2·d −1.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"1 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":"124408998","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}
Zhuoran Ma, Takaharu Tatsuno, Y. Homma, K. Konashi, Katsuya Suzuki, Tatsuya Suzuki
In order to facilitate the dissolution of these insoluble nuclear debris from the Fukushima accident, it is necessary to investigate methods of pulverizing them to increase their surface area. Although solid uranium dioxide is known to become powder through volume changes caused by oxidation, thermal oxidation can lead to the volatilization of quasi-volatile radioactive materials, so it is desirable to cause reaction under the milder conditions. We therefore developed non-equilibrium plasma oxidation device to verify the powderization of uranium dioxide solids and to compare the results with thermochemical oxidation. For the results of the plasma oxidation experiment, Uranium dioxide solid (42mg) can be partly converted to powder by plasma oxidation (6.5h, O2:0.4 L/min) with a low temperature (less than 200 °C). And Partial oxidation of the powder, uranium dioxide to triuranium oxtoxide was confirmed by XRD. Small uneven bumps of 1μm or less appears on the surface of powders produced by oxidation using non-equilibrium plasma, thus increasing the surface area required for dissolution or chemical conversion to soluble compounds can be expected.
{"title":"Comparison of the Powderization Effect of Non-Equilibrium Plasma Oxidation and Thermochemical Oxidation Powders of Uranium Dioxide Solids for Actinide Analysis","authors":"Zhuoran Ma, Takaharu Tatsuno, Y. Homma, K. Konashi, Katsuya Suzuki, Tatsuya Suzuki","doi":"10.1115/icone29-90894","DOIUrl":"https://doi.org/10.1115/icone29-90894","url":null,"abstract":"\u0000 In order to facilitate the dissolution of these insoluble nuclear debris from the Fukushima accident, it is necessary to investigate methods of pulverizing them to increase their surface area. Although solid uranium dioxide is known to become powder through volume changes caused by oxidation, thermal oxidation can lead to the volatilization of quasi-volatile radioactive materials, so it is desirable to cause reaction under the milder conditions. We therefore developed non-equilibrium plasma oxidation device to verify the powderization of uranium dioxide solids and to compare the results with thermochemical oxidation. For the results of the plasma oxidation experiment, Uranium dioxide solid (42mg) can be partly converted to powder by plasma oxidation (6.5h, O2:0.4 L/min) with a low temperature (less than 200 °C). And Partial oxidation of the powder, uranium dioxide to triuranium oxtoxide was confirmed by XRD. Small uneven bumps of 1μm or less appears on the surface of powders produced by oxidation using non-equilibrium plasma, thus increasing the surface area required for dissolution or chemical conversion to soluble compounds can be expected.","PeriodicalId":249213,"journal":{"name":"Volume 9: Decontamination and Decommissioning, Radiation Protection, and Waste Management","volume":"3 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":"114633373","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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