Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis code, the BEAVRS benchmark is adopted as the research object. The BEAVRS problem is a benchmark problem based on real commercial PWRs with detailed material description, geometric information, core operating history, and detector measurement data. In this paper, the CORAL code is used to carry out physical analysis and calculation for the BEAVRS benchmark, and the response rate distribution of radial and axial detectors can be obtained. By comparing with the measurement results provided by the benchmark, it can be found that the calculation results of the CORAL code are in good agreement. This shows that the CORAL code can well simulate the detector response distribution in the core.
{"title":"Calculation of Detector Response Rate Based on BEAVRS Benchmark Using PWR-Core Analysis Code CORAL","authors":"Wen Yang, Bao Liu, Shiwei Yao, Fei Chao, Xing Li, Jinrong Qiu","doi":"10.1115/icone29-93495","DOIUrl":"https://doi.org/10.1115/icone29-93495","url":null,"abstract":"\u0000 Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis code, the BEAVRS benchmark is adopted as the research object. The BEAVRS problem is a benchmark problem based on real commercial PWRs with detailed material description, geometric information, core operating history, and detector measurement data. In this paper, the CORAL code is used to carry out physical analysis and calculation for the BEAVRS benchmark, and the response rate distribution of radial and axial detectors can be obtained. By comparing with the measurement results provided by the benchmark, it can be found that the calculation results of the CORAL code are in good agreement. This shows that the CORAL code can well simulate the detector response distribution in the core.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"62 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84293955","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}
Jiqiang Su, Qun Liu, Honglin Zhang, Lei Shi, Hongjun Liu, Yanrui Li, Jian Hu
As the number of nuclear power plants put into operation in China continues to increase, more and more spent fuel is produced. Before the spent fuel leaves the spent fuel pool to the final disposal of the spent fuel, it needs to go through multiple and various forms of transportation and storage. By establishing an efficient scheduling system for spent fuel, the efficient scheduling and overall coordination of each link is crucial for the safe management of spent fuel. This paper studies the connotation of the efficient scheduling system of spent fuel. Based on the enlightenment of the establishment of the efficient scheduling system for spent fuel in major nuclear power countries, it analyzes the necessity and current situation of establishing an efficient scheduling system for spent fuel in China’s nuclear power plants, and puts forward relevant measures and suggestions.
{"title":"Research on Establishing Efficient Scheduling System of Spent Fuel From Chinese Nuclear Power Plants","authors":"Jiqiang Su, Qun Liu, Honglin Zhang, Lei Shi, Hongjun Liu, Yanrui Li, Jian Hu","doi":"10.1115/icone29-92757","DOIUrl":"https://doi.org/10.1115/icone29-92757","url":null,"abstract":"\u0000 As the number of nuclear power plants put into operation in China continues to increase, more and more spent fuel is produced. Before the spent fuel leaves the spent fuel pool to the final disposal of the spent fuel, it needs to go through multiple and various forms of transportation and storage. By establishing an efficient scheduling system for spent fuel, the efficient scheduling and overall coordination of each link is crucial for the safe management of spent fuel. This paper studies the connotation of the efficient scheduling system of spent fuel. Based on the enlightenment of the establishment of the efficient scheduling system for spent fuel in major nuclear power countries, it analyzes the necessity and current situation of establishing an efficient scheduling system for spent fuel in China’s nuclear power plants, and puts forward relevant measures and suggestions.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"145 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79934323","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}
Because of its excellent heat resistance, nuclear radiation resistance, corrosion resistance and high mechanical properties, PEEK is considered as a promising insulating material in the 21st century, but at present, there is still no research on the mechanical property changes after immersion in boric acid water and aging. This paper has deeply studied the above problems through experiments, the test results show that the PEEK material will not crack or break in the air under high stress levels, however, long-term immersion in boric acid solution may lead to complete separation of the two pieces. Soaking in hot water can increase the IS of the PEEK material, so in order to prevent fracture and improve the toughness of the material, the PEEK material may be soaked in hot water for a certain amount of time, and ensure that the sum of preloading force and acting force when the material is used is lower than the allowable stress value after immersion reduction, then can be used.
{"title":"Study on Properties of Aging and Immersion in Boric Acid Water of PEEK Insulation Material","authors":"J. An, Shan Jin, Li-Qin Zhang, Hao Wu","doi":"10.1115/icone29-93867","DOIUrl":"https://doi.org/10.1115/icone29-93867","url":null,"abstract":"\u0000 Because of its excellent heat resistance, nuclear radiation resistance, corrosion resistance and high mechanical properties, PEEK is considered as a promising insulating material in the 21st century, but at present, there is still no research on the mechanical property changes after immersion in boric acid water and aging. This paper has deeply studied the above problems through experiments, the test results show that the PEEK material will not crack or break in the air under high stress levels, however, long-term immersion in boric acid solution may lead to complete separation of the two pieces. Soaking in hot water can increase the IS of the PEEK material, so in order to prevent fracture and improve the toughness of the material, the PEEK material may be soaked in hot water for a certain amount of time, and ensure that the sum of preloading force and acting force when the material is used is lower than the allowable stress value after immersion reduction, then can be used.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"19 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91057296","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}
Boran Kong, Han Zhang, Kaijie Zhu, C. Hao, Jiong Guo, Fu Li
Optimally diffusive coarse mesh finite difference (odCMFD) method is a recently developed acceleration method for neutron transport equation. Compared with the traditional CMFD, it adds an optimal theta on the diffusion coefficient. This new acceleration method achieves a higher convergence rate than the traditional CMFD and promises the convergence even in large optical thickness region. However, the optimal theta of odCMFD is determined by the Fourier analysis results of the 1D traditional SN. When applying the odCMFD to accelerate DGFEM based SN, this optimal theta is not always suitable. Therefore, it is necessary to find a new optimal theta for this new scheme. In this paper, a Fourier analysis of DGFEM based SN using odCMFD acceleration for k-eigenvalue neutron transport problem is conducted. Fourier analysis results show that when the number of inner iterations is higher than 10, the increment of the inner iterations has little impact on the spectral radius. Meanwhile, the order of the DGFEM based SN and the order of SN quadrature set have little impact on the spectral radius. The scattering ratio has great impact on the spectral radius, the decrement of scattering ratio increases the spectral radius. Set the number of inner iterations as 10, for different scattering ratio, an optimal theta of odCMFD and a polynomial fitting curve are obtained by Fourier analysis. Finally, numerical estimations of spectral radius are obtained by real 1D DGFEM based SN calculation with odCMFD acceleration. The experiment values fit well with the Fourier analysis theoretical results.
{"title":"The Optimal Theta of Optimally Diffusive Coarse Mesh Finite Difference Method to Accelerate DGFEM Based SN","authors":"Boran Kong, Han Zhang, Kaijie Zhu, C. Hao, Jiong Guo, Fu Li","doi":"10.1115/icone29-89221","DOIUrl":"https://doi.org/10.1115/icone29-89221","url":null,"abstract":"\u0000 Optimally diffusive coarse mesh finite difference (odCMFD) method is a recently developed acceleration method for neutron transport equation. Compared with the traditional CMFD, it adds an optimal theta on the diffusion coefficient. This new acceleration method achieves a higher convergence rate than the traditional CMFD and promises the convergence even in large optical thickness region. However, the optimal theta of odCMFD is determined by the Fourier analysis results of the 1D traditional SN. When applying the odCMFD to accelerate DGFEM based SN, this optimal theta is not always suitable. Therefore, it is necessary to find a new optimal theta for this new scheme. In this paper, a Fourier analysis of DGFEM based SN using odCMFD acceleration for k-eigenvalue neutron transport problem is conducted. Fourier analysis results show that when the number of inner iterations is higher than 10, the increment of the inner iterations has little impact on the spectral radius. Meanwhile, the order of the DGFEM based SN and the order of SN quadrature set have little impact on the spectral radius. The scattering ratio has great impact on the spectral radius, the decrement of scattering ratio increases the spectral radius. Set the number of inner iterations as 10, for different scattering ratio, an optimal theta of odCMFD and a polynomial fitting curve are obtained by Fourier analysis. Finally, numerical estimations of spectral radius are obtained by real 1D DGFEM based SN calculation with odCMFD acceleration. The experiment values fit well with the Fourier analysis theoretical results.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91158356","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}
Neutron shielding materials are widely used in aviation, medical treatment, nuclear reactor and other fields. Neutrons is difficult to shield because of their high energy. With the development of neutron shielding materials, different kinds of shielding materials have been developed. Compared with other kinds of materials, composite material is an ideal candidate for neutron shielding material because of their outstanding physical and chemical properties. Therefore, a lot of researchers continue exploring and preparing novel composite neutron shielding materials to meet the complex working conditions. This paper summarized the research status of different neutron shielding materials in recent years, mainly including inorganic non-metallic based neutron shielding materials, polymer based neutron shielding materials, metal based neutron shielding materials, beside for composite shielding materials. Moreover, the existing problems in the research of shielding materials and the possible future development direction are put forward.
{"title":"Research Progress of Neutron Shielding Materials","authors":"B. Chang, Saisai Li, Minghui Li, Ruoyu Chen","doi":"10.1115/icone29-92210","DOIUrl":"https://doi.org/10.1115/icone29-92210","url":null,"abstract":"\u0000 Neutron shielding materials are widely used in aviation, medical treatment, nuclear reactor and other fields. Neutrons is difficult to shield because of their high energy. With the development of neutron shielding materials, different kinds of shielding materials have been developed. Compared with other kinds of materials, composite material is an ideal candidate for neutron shielding material because of their outstanding physical and chemical properties. Therefore, a lot of researchers continue exploring and preparing novel composite neutron shielding materials to meet the complex working conditions. This paper summarized the research status of different neutron shielding materials in recent years, mainly including inorganic non-metallic based neutron shielding materials, polymer based neutron shielding materials, metal based neutron shielding materials, beside for composite shielding materials. Moreover, the existing problems in the research of shielding materials and the possible future development direction are put forward.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87370110","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}
Bao-Xin Yuan, Xiaojuan Luo, Herong Zeng, Huan Huang, Jie Zheng, Wankui Yang, Songbao Zhang, Bin Zhong, Junxia Wei, Yangjun Ying
At present, the technology of reactor neutron noise is developing rapidly, which can not only be used to measure some intrinsic parameters of the reactor core, but also be used for core fault diagnosis. The theoretical solution of reactor neutron noise is of great significance for providing information on the directivity of core faults, and its analytical solution is extremely difficult or even impossible. Therefore, it is necessary to study the numerical simulation method of reactor neutron noise. In this work, the governing equations of the neutron noise problem are firstly established based on the discrete ordinate finite element theory, the equations take the reactor steady-state parameters and the core disturbance cross-section as input parameters. Secondly, the frequency domain solver FrequencyDomain_SNSolver of neutron noise is developed, which is capable of acquiring steady-state parameters, reading disturbance cross-section and calculating neutron noise. Finally, for a calculation case of an experimental reactor core, the neutron noise distribution and spectrum under perturbation are obtained by using the developed FrequencyDomain_SNSolver. This work can provide some reference and technical support for related research.
{"title":"Numerical Simulation of Reactor Neutron Noise Based on Discrete Ordinate Finite Element Method","authors":"Bao-Xin Yuan, Xiaojuan Luo, Herong Zeng, Huan Huang, Jie Zheng, Wankui Yang, Songbao Zhang, Bin Zhong, Junxia Wei, Yangjun Ying","doi":"10.1115/icone29-91726","DOIUrl":"https://doi.org/10.1115/icone29-91726","url":null,"abstract":"\u0000 At present, the technology of reactor neutron noise is developing rapidly, which can not only be used to measure some intrinsic parameters of the reactor core, but also be used for core fault diagnosis. The theoretical solution of reactor neutron noise is of great significance for providing information on the directivity of core faults, and its analytical solution is extremely difficult or even impossible. Therefore, it is necessary to study the numerical simulation method of reactor neutron noise. In this work, the governing equations of the neutron noise problem are firstly established based on the discrete ordinate finite element theory, the equations take the reactor steady-state parameters and the core disturbance cross-section as input parameters. Secondly, the frequency domain solver FrequencyDomain_SNSolver of neutron noise is developed, which is capable of acquiring steady-state parameters, reading disturbance cross-section and calculating neutron noise. Finally, for a calculation case of an experimental reactor core, the neutron noise distribution and spectrum under perturbation are obtained by using the developed FrequencyDomain_SNSolver. This work can provide some reference and technical support for related research.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"30 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82176074","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. Zhuang, Zhongling Han, Xianglong Guo, Le-fu Zhang
Flow-induced vibration is unavoidable in the steam generator, so the wear of heat transfer tubes can’t be ignored. It is essential to evaluate the fretting wear damage of the heat exchanger tube. In this paper, the high frequency fretting wear tests of Alloy 690 tubes/405 stainless steel plate were carried out in the dynamic water loop experimental device, simulated the secondary water (285°C, 10.5 MPa). Changing the contact stress is realized by applying a range of normal forces. As the contact stress increased, the wear volume increased, and the material removal increased the contact area and wear depth. The wear coefficients of Alloy 690 tubes were found to be intensely dependent on the contact stress. The damage mechanisms of Alloy 690 tube were mainly delamination and abrasive wear, and accompanied by material oxidation in the fretting process. The oxide type of wear debris mainly consisted of (Ni, Fe) (Fe, Cr)2O4 and Fe3O4. The profile shape of wear scar was dependent on the action of the wear particles. When the contact stress is minor, the debris was easy to escape from the interface, and the typical profile shapes were U-shaped. As the contact stress increased, the elastic deformation of the contact material increased, followed by adhesion and delamination in the contact area.
{"title":"Effect of Contact Stress on Fretting Corrosion Behavior of Alloy 690 Tube Exposed to Simulated Secondary Water","authors":"W. Zhuang, Zhongling Han, Xianglong Guo, Le-fu Zhang","doi":"10.1115/icone29-92395","DOIUrl":"https://doi.org/10.1115/icone29-92395","url":null,"abstract":"\u0000 Flow-induced vibration is unavoidable in the steam generator, so the wear of heat transfer tubes can’t be ignored. It is essential to evaluate the fretting wear damage of the heat exchanger tube. In this paper, the high frequency fretting wear tests of Alloy 690 tubes/405 stainless steel plate were carried out in the dynamic water loop experimental device, simulated the secondary water (285°C, 10.5 MPa). Changing the contact stress is realized by applying a range of normal forces. As the contact stress increased, the wear volume increased, and the material removal increased the contact area and wear depth. The wear coefficients of Alloy 690 tubes were found to be intensely dependent on the contact stress. The damage mechanisms of Alloy 690 tube were mainly delamination and abrasive wear, and accompanied by material oxidation in the fretting process. The oxide type of wear debris mainly consisted of (Ni, Fe) (Fe, Cr)2O4 and Fe3O4. The profile shape of wear scar was dependent on the action of the wear particles. When the contact stress is minor, the debris was easy to escape from the interface, and the typical profile shapes were U-shaped. As the contact stress increased, the elastic deformation of the contact material increased, followed by adhesion and delamination in the contact area.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"12 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81785351","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}
Kui Hu, Xubo Ma, Xuan Ma, Chen Zhang, Gu Jia, Yixue Chen
It is important for nuclear reactor physics calculation to generate accurate multigroup cross section libraries by using evaluation nuclear data libraries. A new evaluated nuclear data library processing code, named as AXSP (Advanced Cross Section Processing Code) was developed with independent intellectual property rights. The GroupXS module was developed with the ability of generating multi-group cross-sections based on the Bondarenko method which is always a favorable approximation for the fast spectrum reactor, and the method was also used in GROUPR of NJOY. To verify the accuracy of the GroupXS module, the GROUPR module of NJOY was used, and we corrected three numerical verification bugs of GROUPR. The bugs were described in this study. By comparing the calculation results with GROUPR of NJOY2016, the relative error of multigroup flux, multigroup cross sections of various reaction types are less than 0.01%, and the relative errors of transfer matrices of various reaction types are basically less than 0.1%. All the neutron reaction types of all nuclides in ENDF/B-VII.1, ENDF/B-VIII.0 and CENDL-3.2 can be processed by GroupXS, and the multigroup cross sections generated by GroupXS of AXSP have a good agreement with that generated by GROUPR of NJOY2016.
{"title":"Development and Verification of Multigroup Cross Section Generation Module GroupXS in AXSP","authors":"Kui Hu, Xubo Ma, Xuan Ma, Chen Zhang, Gu Jia, Yixue Chen","doi":"10.1115/icone29-91399","DOIUrl":"https://doi.org/10.1115/icone29-91399","url":null,"abstract":"\u0000 It is important for nuclear reactor physics calculation to generate accurate multigroup cross section libraries by using evaluation nuclear data libraries. A new evaluated nuclear data library processing code, named as AXSP (Advanced Cross Section Processing Code) was developed with independent intellectual property rights. The GroupXS module was developed with the ability of generating multi-group cross-sections based on the Bondarenko method which is always a favorable approximation for the fast spectrum reactor, and the method was also used in GROUPR of NJOY. To verify the accuracy of the GroupXS module, the GROUPR module of NJOY was used, and we corrected three numerical verification bugs of GROUPR. The bugs were described in this study. By comparing the calculation results with GROUPR of NJOY2016, the relative error of multigroup flux, multigroup cross sections of various reaction types are less than 0.01%, and the relative errors of transfer matrices of various reaction types are basically less than 0.1%. All the neutron reaction types of all nuclides in ENDF/B-VII.1, ENDF/B-VIII.0 and CENDL-3.2 can be processed by GroupXS, and the multigroup cross sections generated by GroupXS of AXSP have a good agreement with that generated by GROUPR of NJOY2016.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"30 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89423905","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 order to complete the test of trace concentrations in neon carrier gas of solid tritium breeder system for fusion reactor, it is very necessary to establish a high precision analytical system and develop a method of Gas Chromatography (GC). The GC system was composed of three detectors and five separated columns and other auxiliary systems, meanwhile it has established analysis methods of testing trace He, H2 and impurity components in Ne carrier gas. The results showed that the Relative Standard Deviation (RSD) values of the concentration and peak area of each component were less than 5.0%, and the linear correlation coefficients (R2) were greater than 0.99 with the GC analysis system, which could be used to testing trace concentrations in neon with good repeatability. It could support the requirement of tritium efficiency in the Tritium Analysis and Measurement system (TAMS), besides, it could provide technology data and support for the radiation tritium system.
{"title":"Analysis of Trace Components in Neon by Gas Chromatography","authors":"Lili Yang, Qin Zhan","doi":"10.1115/icone29-92413","DOIUrl":"https://doi.org/10.1115/icone29-92413","url":null,"abstract":"In order to complete the test of trace concentrations in neon carrier gas of solid tritium breeder system for fusion reactor, it is very necessary to establish a high precision analytical system and develop a method of Gas Chromatography (GC). The GC system was composed of three detectors and five separated columns and other auxiliary systems, meanwhile it has established analysis methods of testing trace He, H2 and impurity components in Ne carrier gas. The results showed that the Relative Standard Deviation (RSD) values of the concentration and peak area of each component were less than 5.0%, and the linear correlation coefficients (R2) were greater than 0.99 with the GC analysis system, which could be used to testing trace concentrations in neon with good repeatability. It could support the requirement of tritium efficiency in the Tritium Analysis and Measurement system (TAMS), besides, it could provide technology data and support for the radiation tritium system.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"26 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89560955","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}
Solvent extraction has been widely used in spent fuel reprocessing because of its advantages such as high mass transfer rate, short production cycle, easy operation and large extraction capacity. The ligands containing soft S and N atoms usually have a good effect on the separation of trivalent lanthanides actinides. Herein, a novel extractant, 6-carboxylic di(2-ethylhexyl) amide pyridine (DEHAPA, HA), containing carboxyl and amide pyridine, was designed. The extraction of Nd(III), Eu(III), Am(III) and Cm(III), representing trivalent lanthanides and actinides, from nitric solution has been carried out by DEHAPA diluted in toluene as the organic phase. According to the slope analysis, the results show that the extraction of Ln(III) and An(III) with DEHAPA was governed by ion-exchange mechanism and the extraction equilibrium constants of Nd(III), Eu(III), Am(III) and Cm(III) have been calculated. The effect of concentration indicated that the structure of extraction complexes are 1:3/LnA3 and 1:3/AnA3. The temperature has a slight influence to distribution ratio of extraction Nd(III) and Eu(III). The infrared spectrum of DEHAPA and extracted complex analysis showed that -N-C = O and -O-C = O group coordinated with Nd(III). According to 1:3/LnA3 extracted complex structure, the Nd(III) ion in complex was coordinated with three -N-C = O, -O-C = O and pyridine group from three tridentate A− ligands by three tridentate A− ligand in organic solvent. This work reveals the unique extraction and coordination structure and provides a value reference to design more effective extraction ligands for Ln(III)/An(III) separation.
{"title":"Extraction of Nd(III), Eu(III), Am(III) and Cm(III) With 6-Carboxylic Di(2-Ethylhexyl) Amide Pyridine","authors":"Chao Xu, Yu Du, Tingting Liu, Suliang Yang","doi":"10.1115/icone29-90818","DOIUrl":"https://doi.org/10.1115/icone29-90818","url":null,"abstract":"\u0000 Solvent extraction has been widely used in spent fuel reprocessing because of its advantages such as high mass transfer rate, short production cycle, easy operation and large extraction capacity. The ligands containing soft S and N atoms usually have a good effect on the separation of trivalent lanthanides actinides. Herein, a novel extractant, 6-carboxylic di(2-ethylhexyl) amide pyridine (DEHAPA, HA), containing carboxyl and amide pyridine, was designed. The extraction of Nd(III), Eu(III), Am(III) and Cm(III), representing trivalent lanthanides and actinides, from nitric solution has been carried out by DEHAPA diluted in toluene as the organic phase. According to the slope analysis, the results show that the extraction of Ln(III) and An(III) with DEHAPA was governed by ion-exchange mechanism and the extraction equilibrium constants of Nd(III), Eu(III), Am(III) and Cm(III) have been calculated. The effect of concentration indicated that the structure of extraction complexes are 1:3/LnA3 and 1:3/AnA3. The temperature has a slight influence to distribution ratio of extraction Nd(III) and Eu(III). The infrared spectrum of DEHAPA and extracted complex analysis showed that -N-C = O and -O-C = O group coordinated with Nd(III). According to 1:3/LnA3 extracted complex structure, the Nd(III) ion in complex was coordinated with three -N-C = O, -O-C = O and pyridine group from three tridentate A− ligands by three tridentate A− ligand in organic solvent. This work reveals the unique extraction and coordination structure and provides a value reference to design more effective extraction ligands for Ln(III)/An(III) separation.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"69 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84909407","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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