Journal of Nuclear Fuel Cycle and Waste Technology最新文献_第8页

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-90818

Chao Xu, Yu Du, Tingting Liu, Suliang Yang

{"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}

引用次数: 0

Development and Preliminary Verification of Multi-Group Cross Section Processing Code MAGIC for Fast Reactor Based on Continuous Point Cross Section 基于连续点截面的快堆多组截面处理代码MAGIC的开发与初步验证

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-91396

Bin Zhang, Xubo Ma, Kui Hu, Teng Zhang, Yuqin Huang, Xuan Ma

{"title":"Development and Preliminary Verification of Multi-Group Cross Section Processing Code MAGIC for Fast Reactor Based on Continuous Point Cross Section","authors":"Bin Zhang, Xubo Ma, Kui Hu, Teng Zhang, Yuqin Huang, Xuan Ma","doi":"10.1115/icone29-91396","DOIUrl":"https://doi.org/10.1115/icone29-91396","url":null,"abstract":"\u0000 To meet the needs of fast reactor high precision nuclear data processing, a high precision multi-group cross section processing program MAGIC based on continuous point cross section is developed, which can provide accurate multi-group cross section for fast reactor deterministic program. The MAGIC program uses 1/120 lethargy width to divide the energy interval, and a total of 2082 energy groups are divided in 0.414eV∼14.2MeV for fast reactor energy region. The main functions of the program are as follows: (1) by using the problem dependent neutron spectrum, the effect of resonance interference can be well considered. (2) The resonance self-shielding factor iteration method can be used to better consider the interaction of the resonance self-shielding in the resolved and unresolved resonance energy regions of multiple nuclides. (3) When calculating the elastic scattering matrix, the resonance phenomenon is considered in the one-dimensional elastic scattering cross section, and the prefabricated scattering transfer probability table can greatly improve the calculation efficiency of the scattering matrix under the ultra-fine group energy framework on the premise of maintaining the accuracy. In terms of benchmark verification, the effective increment coefficient was verified by fast energy spectrum critical benchmark based on the multi-group cross section generated by MAGIC. The difference between the calculated results and those of the Monte Carlo program and NJOY was small, which preliminatively verified the effectiveness of the multi-group cross section generated by MAGIC program in the calculation of deterministic method.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"15 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75277899","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}

引用次数: 0

Analysis Method to Predict Grid-to-Rod Fretting Wear of Fuel Rod in Pressurized Water Reactors 压水堆燃料棒网格-棒微动磨损预测分析方法

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-93076

Libing Zhu, Shufeng Zhao, J. Sun, Shiyin Xu, F. Gan, Sirui Li

{"title":"Analysis Method to Predict Grid-to-Rod Fretting Wear of Fuel Rod in Pressurized Water Reactors","authors":"Libing Zhu, Shufeng Zhao, J. Sun, Shiyin Xu, F. Gan, Sirui Li","doi":"10.1115/icone29-93076","DOIUrl":"https://doi.org/10.1115/icone29-93076","url":null,"abstract":"\u0000 Grid-to-rod fretting (GTRF) caused by flow induced vibration is one of the most important mechanisms of fuel rod failure in PWRs, which has a strong impact on the economy and safety of nuclear power plant. To avoid rod failure due to GTRF, several verification tests which are high costs and long term are always carried out for new design fuel assemblies. So as to deal with the disadvantages of tests and speed up progress in fuel design, a progressive analysis method to predict GTRF performance of fuel rod is developed, and systemic tests are carried out to verify the analysis method. The method consists of three main parts, including flow field analysis, structure vibration analysis and wear analysis. In the first part, the flow field excitation force is obtained by the large eddy simulation method, and this CFD method is verified by a particle image velocimetry (PIV) test with a 5 × 5 fuel bundle. In the second part, to get the vibration response of fuel rod, a nonlinear vibration model which can consider end-of-life gaps is established. Full scale fuel rod vibration in air and water conditions are tested to validate the predicted fuel rod response from analysis. In the third part, the fuel rod wear is estimated by tribological theory with fuel rod vibration response, the wear coefficient and the typical feature of wear scar. The GTRF method is finally verified by a long-term wear test of full-scale fuel assembly. It turns out that this method can effectively predict the wear characteristics of fuel rod cladding at the early stage of design. It provides a feasible solution for rapid design iteration of fuel assembly, and could reduce the development costs and risks.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"48 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74079906","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}

引用次数: 0

Design and Investigation of Liquid Lead-Bismuth Test Irradiation Loop in Research Reactor 研究堆液态铅铋试验辐照回路的设计与研究

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-91405

Yuheng Xu, J. Si, Liang Zhang, Sheng-xia Sun, Hongwei Wu, Mengkang Lu

{"title":"Design and Investigation of Liquid Lead-Bismuth Test Irradiation Loop in Research Reactor","authors":"Yuheng Xu, J. Si, Liang Zhang, Sheng-xia Sun, Hongwei Wu, Mengkang Lu","doi":"10.1115/icone29-91405","DOIUrl":"https://doi.org/10.1115/icone29-91405","url":null,"abstract":"\u0000 The Lead-bismuth eutectic (LBE) test irradiation loop is an indispensable testing vehicle for the investigation of LBE-cooled reactor fuel. To make a water-cooled research reactor have the irradiation test capacity of LBE-cooled reactor fuel, a test irradiation loop is investigated and preliminarily designed by absorbing the design and operation experience of the domestic and international heavy liquid metal loop. Structural material corrosion, impurity blockage, polonium diffusion, and metal solidification in the loop are considered in the design process. Corresponding measures have been taken to ensure the safe operation and emergency capacity of the loop. The loop is a 1500KW facility consisting of a coolant system and auxiliary systems the other, operating in a wide temperature range of 250–480 °C, oxygen concentration 0.01–0.05 ppm. The coolant system is to remove the heat from the fuel surface due to irradiation. Auxiliary systems are for coolant chemistry control, emergency fuel cooling, polonium control, and parameter measurement. In this paper, the technological process, main design parameters, and system composition of the loop are described, and the parameters and selections of main equipment, instrument parameters are proposed. The investigation supplies a useful reference for further design of test irradiation loop systems and equipment.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"21 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79194439","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}

引用次数: 0

Oxidation Behaviour of CZ Alloys Under High Temperature Steam CZ合金在高温蒸汽下的氧化行为

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-93795

Minli Chen, Changyuan Gao, Guoliang Zhang, Lin Shi, Liu-tao Chen, J. Tan

{"title":"Oxidation Behaviour of CZ Alloys Under High Temperature Steam","authors":"Minli Chen, Changyuan Gao, Guoliang Zhang, Lin Shi, Liu-tao Chen, J. Tan","doi":"10.1115/icone29-93795","DOIUrl":"https://doi.org/10.1115/icone29-93795","url":null,"abstract":"\u0000 CZ Alloys, including CZ1 and CZ2, are new advanced zirconium alloys developed by China General Nuclear Power Group. This paper presents the oxidation behaviour of CZ Alloys and Zircaloy-4 in flowing steam. A SETSYS Evolution TGA (ThermoGravimetric Analyzer) is used for the high temperature testing. The weight gain of specimens was measured real time in the temperature range of 800∼1200 °C for 3000∼5000s. The hydrogen content after oxidation at 1000 °C was measured. CZ1 (Zr-Sn-Nb alloy), CZ2 (Zr-Nb alloy) and Zircaloy-4 (Zr-Sn alloy) claddings which contain different types of alloying elements have similar oxidation behaviour at above 1100°C but show significant difference at lower temperatures. The weight gain of CZ Alloys are compared with high-temperature oxidation parabolic rate laws of fuel cladding materials which have been used in many LOCA analyses, the Baker-Just equation and the Carthcart-Pawel equation. The results show that the Baker-Just equation is conservative for CZ Alloys and the Carthcart-Pawel equation is more accurate at above 1100°C. The oxidation rate constant and the rate exponent are calculated at each temperature by non-linear fittings.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"41 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88785211","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}

引用次数: 0

The Research of Corrosion Resistance of CZ2 Zirconium Alloy Welding Specimens CZ2锆合金焊接试样的耐腐蚀性能研究

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-93838

Yang Xu, Liu-tao Chen, Chang-Yuan Gao, J. Tan, Lin Shi

{"title":"The Research of Corrosion Resistance of CZ2 Zirconium Alloy Welding Specimens","authors":"Yang Xu, Liu-tao Chen, Chang-Yuan Gao, J. Tan, Lin Shi","doi":"10.1115/icone29-93838","DOIUrl":"https://doi.org/10.1115/icone29-93838","url":null,"abstract":"\u0000 Zirconium alloys are one of the main cladding materials in pressurized water reactors. Corrosion resistance of zirconium alloy cladding weld is one of the most important factor affecting fuel integrity. CZ zirconium alloys have been recently developed for advanced PWR fuel assemblies which are designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 and CZ2 respectively. CZ1 is Zr-Sn-Nb alloy, and CZ2 is Zr-Nb alloy. In this paper, the autoclave corrosion test of CZ2 zirconium alloy welded specimen was executed. Then, the corrosion resistance of CZ2 zirconium alloy welding specimens was evaluated by analyzing the the appearance, corrosion weight gain and oxide film thickness. The result shows that corrosion resistance trend of combination of CZ2 and CZ2, CZ2 and Zr-4 welding specimens is basically the same under the conditions of 360 °C /18.6 MPa deionized water and 400 °C /10.3 MPa superheated steam. There was no obvious difference due to different hydro-chemistry. The corrosion resistance of CZ2 and Zr-4 welding specimens is better than that of CZ2 and CZ2 welding specimens. Therefore, the use of Zr-4 end plug material can reduce the sensitivity of CZ2 alloy to heat effect and improve the corrosion resistance of CZ2 alloy cladding.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"9 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87413780","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}

引用次数: 0

A Method to Evaluate Defect Size of Failed Fuel Rods in CGN 1000MWe PWRs 中广核1000MWe压水堆失效燃料棒缺陷尺寸评估方法

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-92788

Pengtao Fu, Zhijun Li

{"title":"A Method to Evaluate Defect Size of Failed Fuel Rods in CGN 1000MWe PWRs","authors":"Pengtao Fu, Zhijun Li","doi":"10.1115/icone29-92788","DOIUrl":"https://doi.org/10.1115/icone29-92788","url":null,"abstract":"\u0000 Fuel cladding is the first barrier to confine the release of fission products to the environment during normal operation. Once the fuel rods defect, the fission product accumulated in the pellet-cladding gap of failed fuel rods will release into the primary loop and lead to high activity levels in a pressurized water reactor. It is necessary to determine the status of fuel failure to avoid its degradation in the operation.\u0000 This paper introduces the mechanism of the generation and release of fission products in fuel rods and the primary loops in a pressurized water reactor, and provides the theoretical method of the regression between release-to-birth R/B and decay constant λ of radio-iodine. It has been verified by one typical case of fuel failure in CGN 1000 MWe PWR, but the post-irradiation examinations in hot cell reveal that six defects exist in one failed rod and the actinides have an obvious release from the failed rods due to secondary hydriding. The actual status of defect size of failed rods is much more complicated than expected in most analysis codes and modelings and it is suggested to introduce some parameters, e.g. the equivalent defect size, to describe the status of fuel failure in future research. The related operation experience in CGN has also been presented to help to better understand the status of fuel failure in PWR units.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"377 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80585603","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}

引用次数: 0

Study and Development of Irradiation Test Device for Fuel Rods In Research Reactor 研究堆燃料棒辐照试验装置的研制

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Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-90997

W. Zhao, Junping Si, Liang Zhang, Sheng-xia Sun, Wenhua Yang, Shuai Jin, Jin Lei, Hao-ran Liu

{"title":"Study and Development of Irradiation Test Device for Fuel Rods In Research Reactor","authors":"W. Zhao, Junping Si, Liang Zhang, Sheng-xia Sun, Wenhua Yang, Shuai Jin, Jin Lei, Hao-ran Liu","doi":"10.1115/icone29-90997","DOIUrl":"https://doi.org/10.1115/icone29-90997","url":null,"abstract":"\u0000 For the purpose of performing high-temperature irradiation of the fuel rod in the low-temperature and low-pressure water-cooled research reactor and obtaining the post-irradiation performance parameters of the fuel rod, a fuel rod irradiation test device was designed based on the High Flux Engineering Test Reactor (HFETR). The key parameters such as thickness of the lead-bismuth eutectic (LBE) alloy layer, structure of the cooling water flow channel, the structure of throttle plug, and flow rate of the cooling water were determined. The flow field distribution in the test section was obtained by numerical simulation method. The stress values of the device under design and accident conditions were obtained by finite element analysis. The test of hydraulic characteristic out-reactor were performed by using the simulation device. The results show that, the average velocity of the cooling water outside the test assembly was 1.9 m/s under the design flow rate, which met the cooling requirement when the linear power of the fuel rod was 280 W/cm. The structure of the test device met the strength requirement with sufficient safety margin. In the experiment out-reactor, the test device was able to cover the design value of 4.03 t/h on the differential pressure of 200 kPa and 300 kPa with coolant flow rate adjustment ranges of 0.73 t/h ∼ 4.28 t/h and 1.05 t/h ∼ 5.25 t/h, which met the test requirements. The device designed in this paper can be used to perform the irradiation test in research reactor, which provides guarantee for the research and development of new type of the fuel.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"88 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90670810","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}

引用次数: 0

Improvements on Fuel Failures Diagnosis Method for Pressurized Water Reactor Nuclear Power Plant Based on Radioactive Activities Inside the Waste Gas Treatment System 基于废气处理系统内放射性活动的压水堆核电站燃料故障诊断方法改进

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-92996

Weifeng Lyu, Runchun Guo, Wenwang Ran, Yaoyi Gao, Jun Xiong

{"title":"Improvements on Fuel Failures Diagnosis Method for Pressurized Water Reactor Nuclear Power Plant Based on Radioactive Activities Inside the Waste Gas Treatment System","authors":"Weifeng Lyu, Runchun Guo, Wenwang Ran, Yaoyi Gao, Jun Xiong","doi":"10.1115/icone29-92996","DOIUrl":"https://doi.org/10.1115/icone29-92996","url":null,"abstract":"\u0000 The cladding of fuel elements in the core of a pressurized water reactor nuclear power plant (PWR) is the first barrier to the radioactive material in the reactor. Once the fuel cladding fails, the radioactive fission products or even fissile material in the fuel elements will be released into the primary coolant, thereby challenging the safe operation of the plant. Fuel cladding failures diagnosis is indispensable for PWR as fuel failure is not inevitable and the impact on nuclear safety is significant. Based on the continuous research on fission product generation and release mechanisms for decades, several fuel failure diagnosis methods for PWR have been established and specific nuclides have been selected as indicators of fuel failures. Recent research shows Kr-85 is a good indicator for the diagnosis of the number of failed fuel rods for its insensitivity to the fuel rod defect size and failed fuel burn-up, but it is hard to measure the activity of Kr-85 in the primary coolant due to the interference of other nuclides. However, the activity of Kr-85 inside the waste gas treatment system can be measured where the activity of other nuclides has been significantly reduced by decay. Hence, the fuel failure diagnosis method is improved based on Kr-85 activities inside the waste gas treatment system. The operating data of Chinese in-service nuclear power plants are used to validate the improved fuel failure diagnosis method, and the validation results show that the improved diagnosis method of fuel failures can accurately diagnose the fuel failure and has wider applicability.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"71 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86364189","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}

引用次数: 0