Correction:Nuclear Science and Techniques(2025)36:4 https://doi.org/10.1007/s41365-024-01548-5 In this article,the caption for Fig(s)1,3,and 4 was inadvert-ently truncated.The incorrect and the corrected captions are ...Correction:Nuclear Science and Techniques(2025)36:4 https://doi.org/10.1007/s41365-024-01548-5 In this article,the caption for Fig(s)1,3,and 4 was inadvert-ently truncated.The incorrect and the corrected captions are given below.展开更多
Based on the service characteristics of fuel elements for molten salt reactors,they need to have a high power density,resistance to coolant infiltration,and excellent thermodynamic properties.To solve the problem of t...Based on the service characteristics of fuel elements for molten salt reactors,they need to have a high power density,resistance to coolant infiltration,and excellent thermodynamic properties.To solve the problem of the graphite used in the fuel element for these reactors being susceptible to molten salt infiltration,carbon black(CB)was added to increase the density of the graphite,and a fuel element(TRISO(tri-structural isotropic)fuel particles were randomly distributed in the modified graphite matrix)was prepared by cold isostatic pressing process.An out-of-pile performance study shows that the densification and pore structure of the modified graphite matrix were improved,as was the resistance to molten salt infiltration.The median pore size of the modified graphite was reduced from 673 to 433 nm and the threshold pressure for molten salt(FLiBe,66%(molar fraction)LiF and 34%BeF_(2))infiltration was increased from 0.88 to 1.37 MPa.The isotropic CB made the graphite matrix less anisotropic,while its thermal conductivity and compressive strength were reduced due to the difficult graphitization of CB.Fuel elements containing 20%(volume fraction)TRISO particles were prepared.Numerical simulations show that the power and temperature distribution of the fuel were in line with the design requirements.The modified graphite matrix had a higher density,smaller pores,a lower anisotropy and a greater resistance to FLiBe infiltration.展开更多
Using squeeze-infiltration technique, Mg-9Al-1Zn-0.8Ce composite reinforced by graphite particles and Al2O3 short fibers was fabricated. The reinforcing phases combined closely with the matrix and no agglomeration was...Using squeeze-infiltration technique, Mg-9Al-1Zn-0.8Ce composite reinforced by graphite particles and Al2O3 short fibers was fabricated. The reinforcing phases combined closely with the matrix and no agglomeration was observed. The microstructure, hardness and wear property of the composites with the graphite content of 5%, 10%, 15% and 20% were investigated, respectively. The results reveal that Ce tends to enrich around the boundaries of graphite particles and Al2O3 short fibers, and forms Al3Ce phase. When the graphite content increases to 20%, the grain size becomes small. Moreover, with increasing the graphite content, the microhardness of the composites decreases but the wear resistance increases. The graphite which works as lubricant during dry sliding process decreases the wear loss. At low load, the wear mechanism of the composite is mainly abrasive wear and oxidation wear; at high load, except that the composite with 20% graphite is still with abrasive wear and oxidation wear, the wear mechanism of other composites changes to delamination wear.展开更多
Compared with the long use of carbon materials in human history,the debut of carbon materials in the Chicago Pile-1 nuclear reactor took place only 70 years ago.Since then,carbon materials have played important roles ...Compared with the long use of carbon materials in human history,the debut of carbon materials in the Chicago Pile-1 nuclear reactor took place only 70 years ago.Since then,carbon materials have played important roles in nuclear reactors,especially in high temperature gas-cooled reactors(HTRs)because of their many excellent properties.As the most promising candidate for Generation IV reactors,a demonstration plant for HTRs,an HTR pebble-bed module(HTR-PM)is currently under construction in China.In the HTR-PM,carbon materials act as the core structural material,reflector,fuel matrix,moderator,and thermal and neutron shields.Because the dimensions and properties of the carbon are generally influenced by the high temperature and neutron irradiation in the HTR-PM,there are rigorous requirements for their performance.Since the precursor materials such as cokes and natural graphite,and the subsequent forming method play a critical role in determining the structure,properties and performance of the material under irradiation,a judicious selection of the raw materials and forming method is required to obtain the desired structure and properties.This paper introduces the detailed property requirements of different carbon materials in the HTR-PM and their fabrication processes.In addition,the current status and future commercialization of the HTR-PM in China and abroad are presented.In order to meet the requirement of full local production in a commercial HTR,long-term considerations such as the sustainable and stable supply of the raw materials,optimization of the manufacturing process in the local production of nuclear graphite for structural graphite and graphite pebbles,and the stable production and reduced cost of the precursor materials are discussed.Finally,current progress and future arrangements for the irradiation testing of Chinese nuclear graphite at the Oak Ridge National Laboratory(USA)are presented.This manuscript is intended to act as a reference for carbon material producers who intend to develop nuclear graphite and carbon materials for use in future commercial HTRs.Meanwhile,a great deal of information introduced in the manuscript is also useful for scientific researchers of carbon materials.展开更多
文摘Correction:Nuclear Science and Techniques(2025)36:4 https://doi.org/10.1007/s41365-024-01548-5 In this article,the caption for Fig(s)1,3,and 4 was inadvert-ently truncated.The incorrect and the corrected captions are given below.
文摘Based on the service characteristics of fuel elements for molten salt reactors,they need to have a high power density,resistance to coolant infiltration,and excellent thermodynamic properties.To solve the problem of the graphite used in the fuel element for these reactors being susceptible to molten salt infiltration,carbon black(CB)was added to increase the density of the graphite,and a fuel element(TRISO(tri-structural isotropic)fuel particles were randomly distributed in the modified graphite matrix)was prepared by cold isostatic pressing process.An out-of-pile performance study shows that the densification and pore structure of the modified graphite matrix were improved,as was the resistance to molten salt infiltration.The median pore size of the modified graphite was reduced from 673 to 433 nm and the threshold pressure for molten salt(FLiBe,66%(molar fraction)LiF and 34%BeF_(2))infiltration was increased from 0.88 to 1.37 MPa.The isotropic CB made the graphite matrix less anisotropic,while its thermal conductivity and compressive strength were reduced due to the difficult graphitization of CB.Fuel elements containing 20%(volume fraction)TRISO particles were prepared.Numerical simulations show that the power and temperature distribution of the fuel were in line with the design requirements.The modified graphite matrix had a higher density,smaller pores,a lower anisotropy and a greater resistance to FLiBe infiltration.
基金Project(2006BAE04B04-1) supported by the Special Task Document of National Science and Technology Program of ChinaProject(20060308) supported by Science and Technology Development Program of Jilin Province, ChinaProject supported by "985 Project" of Jilin University, China
文摘Using squeeze-infiltration technique, Mg-9Al-1Zn-0.8Ce composite reinforced by graphite particles and Al2O3 short fibers was fabricated. The reinforcing phases combined closely with the matrix and no agglomeration was observed. The microstructure, hardness and wear property of the composites with the graphite content of 5%, 10%, 15% and 20% were investigated, respectively. The results reveal that Ce tends to enrich around the boundaries of graphite particles and Al2O3 short fibers, and forms Al3Ce phase. When the graphite content increases to 20%, the grain size becomes small. Moreover, with increasing the graphite content, the microhardness of the composites decreases but the wear resistance increases. The graphite which works as lubricant during dry sliding process decreases the wear loss. At low load, the wear mechanism of the composite is mainly abrasive wear and oxidation wear; at high load, except that the composite with 20% graphite is still with abrasive wear and oxidation wear, the wear mechanism of other composites changes to delamination wear.
文摘Compared with the long use of carbon materials in human history,the debut of carbon materials in the Chicago Pile-1 nuclear reactor took place only 70 years ago.Since then,carbon materials have played important roles in nuclear reactors,especially in high temperature gas-cooled reactors(HTRs)because of their many excellent properties.As the most promising candidate for Generation IV reactors,a demonstration plant for HTRs,an HTR pebble-bed module(HTR-PM)is currently under construction in China.In the HTR-PM,carbon materials act as the core structural material,reflector,fuel matrix,moderator,and thermal and neutron shields.Because the dimensions and properties of the carbon are generally influenced by the high temperature and neutron irradiation in the HTR-PM,there are rigorous requirements for their performance.Since the precursor materials such as cokes and natural graphite,and the subsequent forming method play a critical role in determining the structure,properties and performance of the material under irradiation,a judicious selection of the raw materials and forming method is required to obtain the desired structure and properties.This paper introduces the detailed property requirements of different carbon materials in the HTR-PM and their fabrication processes.In addition,the current status and future commercialization of the HTR-PM in China and abroad are presented.In order to meet the requirement of full local production in a commercial HTR,long-term considerations such as the sustainable and stable supply of the raw materials,optimization of the manufacturing process in the local production of nuclear graphite for structural graphite and graphite pebbles,and the stable production and reduced cost of the precursor materials are discussed.Finally,current progress and future arrangements for the irradiation testing of Chinese nuclear graphite at the Oak Ridge National Laboratory(USA)are presented.This manuscript is intended to act as a reference for carbon material producers who intend to develop nuclear graphite and carbon materials for use in future commercial HTRs.Meanwhile,a great deal of information introduced in the manuscript is also useful for scientific researchers of carbon materials.
