Fine-grained nuclear graphite is a key material in high-temperature gas-cooled reactors(HTGRs).During air ingress accidents,core graphite components undergo severe oxidation,threatening structural integrity.Therefore,...Fine-grained nuclear graphite is a key material in high-temperature gas-cooled reactors(HTGRs).During air ingress accidents,core graphite components undergo severe oxidation,threatening structural integrity.Therefore,understanding the oxidation behavior of nuclear graphite is essential for reactor safety.The influence of oxidation involves multiple factors,including temperature,sample size,oxidant,impurities,filler type and size,etc.The size of the filler particles plays a crucial role in this study.Five ultrafine-and superfine-grained nuclear graphite samples(5.9-34.4μm)are manufactured using identical raw materials and manufacturing processes.Isothermal oxidation tests conducted at 650℃-750℃ are used to study the oxidation behavior.Additionally,comprehensive characterization is performed to analyze the crystal structure,surface morphology,and nanoscale to microscale pore structure of the samples.Results indicate that oxidation behavior cannot be predicted solely based on filler grain size.Reactive site concentration,characterized by active surface area,dominates the chemical reaction kinetics,whereas pore tortuosity,quantified by the structural parameterΨ,plays a key role in regulating oxidant diffusion.These findings clarify the dual role of microstructure in oxidation mechanisms and establish a theoretical and experimental basis for the design of high-performance nuclear graphite capable of long-term service in high-temperature gas-cooled reactors.展开更多
The irradiation behavior of graphite is essential for its applications in the nuclear industry.However,the behavioral differences of graphite remain obscure because of the very limited comprehension of its microstruct...The irradiation behavior of graphite is essential for its applications in the nuclear industry.However,the behavioral differences of graphite remain obscure because of the very limited comprehension of its microstructural differences.One typical structure,the quinoline-insoluble(QI)particle,was investigated using IG-110 and NBG-18 graphite.After irradiation,the QI particles on the polished surface were proven to become hillocks,which were easily identifiable via scanning electron microscopy(SEM).Thus,a method that combined ion irradiation and SEM characterization was proposed to study the distribution and concentration of QI particles in graphite.During irradiation,the QI particles were found to evolve into densified spheres,which were weakly bonded with the surrounding graphite structures,thereby indicating that the densification of QI particles did not evidently contribute to graphite dimensional shrinkage.A much higher concentration of QI particles in NBG-18 than IG-110,which was suggested to be responsible for the smaller maximum dimensional shrinkage of former over the latter during irradiation,was characterized.展开更多
Graphite's resilience to high temperatures and neutron damage makes it vital for nuclear reactors,yet irradiation alters its microstructure,degrading key properties.We used small-and wide-angle X-ray scattering to...Graphite's resilience to high temperatures and neutron damage makes it vital for nuclear reactors,yet irradiation alters its microstructure,degrading key properties.We used small-and wide-angle X-ray scattering to study neutron-irradiated fine-grain nuclear graphite(Grade G347A)across varied temperatures and fluences.Results show significant shifts in internal strain and porosity,correlating with radiation-induced volume changes.Notably,porosity volume distribution(fractal dimensions)follows non-monotonic volume changes,suggesting a link to the Weibull distribution of fracture stress.展开更多
The operational lifespan of nuclear graphite is significantly affected by irradiation creep,yet the microstructural mechanism underlying this creep phenomenon remains unclear.Some theories attempt to link microstructu...The operational lifespan of nuclear graphite is significantly affected by irradiation creep,yet the microstructural mechanism underlying this creep phenomenon remains unclear.Some theories attempt to link microstructural evolution with creep behavior,but the rapid migration rate of defects under irradiation and loading makes it difficult to capture the specific evolution process experimentally,resulting in a lack of direct structural evidence.Therefore,in this study,molecular dynamics simulations are employed to investigate the irradiation behavior and microstructural migration under external loading.The aim is to provide microstructural evidence for theories such as the dislocation pinning-unpinning and crystal yielding.The results demonstrate that high tensile loads can increase the potential energy and reduce threshold displacement energy of graphite crystals.Consequently,displacement damage probability and creep rate increase,which is not considered in previous theories.Meanwhile,different creep mechanisms are observed at different damage states and applied loads.In low-dose damage states dominated by interstitials and vacancies,the pinning-unpinning process at basal plane may be caused by a defect diffusion mode.Under high stress levels,direct breaking of pinning structures occurs,leading to rapid migration of basal planes,demonstrating the microstructural evolution process of irradiated crystal yielding and plastic flow.In high-dose damage states characterized significantly by amorphous components,short-range atomic diffusion can become the dominant creep mechanism,and diffusion along the c-axis of graphite crystals is no longer constrained.These findings provide a crucial reference for understanding the irradiation and creep behavior of nuclear graphite in reactors.展开更多
This paper intensively explores the critical issues related to the quantitative and accurate evaluations of FCG behavior in the early stage,macro fatigue fracture toughness,and the critical crack size for damage toler...This paper intensively explores the critical issues related to the quantitative and accurate evaluations of FCG behavior in the early stage,macro fatigue fracture toughness,and the critical crack size for damage tolerance in nuclear graphite.To address these issues,scale-span FCG tests were carried out using two typical specimens,CT and SEM in-situ specimens.These results indicate that the FCG threshold and the effective FCG length have a significant correlation with the modified maximum loop stress theory for a mixed I/II mode.In particular,the effective FCG length(a_(eq))and the applied stress threshold of polycrystalline graphite are important parameters for fatigue damage tolerance design in engineering application.The influencing factors of ΔK_(th,eq) and a_(eq) were discussed in detail using the mixed I/II mode,respectively.In addition,the scattered values of ΔK_(IC) for this graphite can be quantitatively estimated using the Weibull distribution equation.The predicated parameters and experimental results demonstrate a strong correlation.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
基金supported by the National Key Research and Development Program of China(2024YFA1612900)the National Natural Science Foundation of China(Grant No.52103365 and No.12375270)the Guangdong Innovative and Entrepreneurial Research Team Program,China(Grant No.2021ZT09L227).
文摘Fine-grained nuclear graphite is a key material in high-temperature gas-cooled reactors(HTGRs).During air ingress accidents,core graphite components undergo severe oxidation,threatening structural integrity.Therefore,understanding the oxidation behavior of nuclear graphite is essential for reactor safety.The influence of oxidation involves multiple factors,including temperature,sample size,oxidant,impurities,filler type and size,etc.The size of the filler particles plays a crucial role in this study.Five ultrafine-and superfine-grained nuclear graphite samples(5.9-34.4μm)are manufactured using identical raw materials and manufacturing processes.Isothermal oxidation tests conducted at 650℃-750℃ are used to study the oxidation behavior.Additionally,comprehensive characterization is performed to analyze the crystal structure,surface morphology,and nanoscale to microscale pore structure of the samples.Results indicate that oxidation behavior cannot be predicted solely based on filler grain size.Reactive site concentration,characterized by active surface area,dominates the chemical reaction kinetics,whereas pore tortuosity,quantified by the structural parameterΨ,plays a key role in regulating oxidant diffusion.These findings clarify the dual role of microstructure in oxidation mechanisms and establish a theoretical and experimental basis for the design of high-performance nuclear graphite capable of long-term service in high-temperature gas-cooled reactors.
基金This work was supported by Youth Innovation Promotion Association of the Chinese Academy of Sciences(No.2019262)the National Natural Science Foundation of China(Nos.11505265,11805256,11805261).
文摘The irradiation behavior of graphite is essential for its applications in the nuclear industry.However,the behavioral differences of graphite remain obscure because of the very limited comprehension of its microstructural differences.One typical structure,the quinoline-insoluble(QI)particle,was investigated using IG-110 and NBG-18 graphite.After irradiation,the QI particles on the polished surface were proven to become hillocks,which were easily identifiable via scanning electron microscopy(SEM).Thus,a method that combined ion irradiation and SEM characterization was proposed to study the distribution and concentration of QI particles in graphite.During irradiation,the QI particles were found to evolve into densified spheres,which were weakly bonded with the surrounding graphite structures,thereby indicating that the densification of QI particles did not evidently contribute to graphite dimensional shrinkage.A much higher concentration of QI particles in NBG-18 than IG-110,which was suggested to be responsible for the smaller maximum dimensional shrinkage of former over the latter during irradiation,was characterized.
基金supported by the U.S.Department of Energy under Contracts(DE-AC02-06CH11357,DE-AC07-05ID14517,DESC0018322,IRP-22-27674).
文摘Graphite's resilience to high temperatures and neutron damage makes it vital for nuclear reactors,yet irradiation alters its microstructure,degrading key properties.We used small-and wide-angle X-ray scattering to study neutron-irradiated fine-grain nuclear graphite(Grade G347A)across varied temperatures and fluences.Results show significant shifts in internal strain and porosity,correlating with radiation-induced volume changes.Notably,porosity volume distribution(fractal dimensions)follows non-monotonic volume changes,suggesting a link to the Weibull distribution of fracture stress.
基金supported the Science and Technology Commission of Shanghai Municipality(No.21DZ2206900)。
文摘The operational lifespan of nuclear graphite is significantly affected by irradiation creep,yet the microstructural mechanism underlying this creep phenomenon remains unclear.Some theories attempt to link microstructural evolution with creep behavior,but the rapid migration rate of defects under irradiation and loading makes it difficult to capture the specific evolution process experimentally,resulting in a lack of direct structural evidence.Therefore,in this study,molecular dynamics simulations are employed to investigate the irradiation behavior and microstructural migration under external loading.The aim is to provide microstructural evidence for theories such as the dislocation pinning-unpinning and crystal yielding.The results demonstrate that high tensile loads can increase the potential energy and reduce threshold displacement energy of graphite crystals.Consequently,displacement damage probability and creep rate increase,which is not considered in previous theories.Meanwhile,different creep mechanisms are observed at different damage states and applied loads.In low-dose damage states dominated by interstitials and vacancies,the pinning-unpinning process at basal plane may be caused by a defect diffusion mode.Under high stress levels,direct breaking of pinning structures occurs,leading to rapid migration of basal planes,demonstrating the microstructural evolution process of irradiated crystal yielding and plastic flow.In high-dose damage states characterized significantly by amorphous components,short-range atomic diffusion can become the dominant creep mechanism,and diffusion along the c-axis of graphite crystals is no longer constrained.These findings provide a crucial reference for understanding the irradiation and creep behavior of nuclear graphite in reactors.
基金supported by the National S&T Major Project(Grant No.ZX06901)Additional funding was provided by the National Natural Science Foundation of China(Grant Nos.11572170 and 11872225).
文摘This paper intensively explores the critical issues related to the quantitative and accurate evaluations of FCG behavior in the early stage,macro fatigue fracture toughness,and the critical crack size for damage tolerance in nuclear graphite.To address these issues,scale-span FCG tests were carried out using two typical specimens,CT and SEM in-situ specimens.These results indicate that the FCG threshold and the effective FCG length have a significant correlation with the modified maximum loop stress theory for a mixed I/II mode.In particular,the effective FCG length(a_(eq))and the applied stress threshold of polycrystalline graphite are important parameters for fatigue damage tolerance design in engineering application.The influencing factors of ΔK_(th,eq) and a_(eq) were discussed in detail using the mixed I/II mode,respectively.In addition,the scattered values of ΔK_(IC) for this graphite can be quantitatively estimated using the Weibull distribution equation.The predicated parameters and experimental results demonstrate a strong correlation.
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.
