The martensite steels are accepted material used in nuclear power plant.In this study,self-ion irradiation was used to simulate the damage caused by fast neutrons in two kinds of martensite steels,SIMP and T91,under t...The martensite steels are accepted material used in nuclear power plant.In this study,self-ion irradiation was used to simulate the damage caused by fast neutrons in two kinds of martensite steels,SIMP and T91,under the temperature of 300℃.The contrast experiment on the steel samples was carried out with 352.8 MeV Fe-ions.S parameter is a statistical conclusion about vacancy damage caused by irradiation,and it is positively related to the density of vacancy defects Figs.1 and 2 show the change of the S parameter with the irradiation dose.Whereas Figs.3 and 4 show the different S parameters for two kind of steels.展开更多
Because of the low cross-section for neutron capture and its excellent structural, chemical and mechanicalstability, silicon carbide (SiC) is an important material with application in the development of nuclear energy...Because of the low cross-section for neutron capture and its excellent structural, chemical and mechanicalstability, silicon carbide (SiC) is an important material with application in the development of nuclear energyand waste technologies. The (n,) nuclear reaction inevitably introduces numerous He in SiC. Because of the lowsolubility of He atoms in SiC, a certain concentration of He atoms that are trapped in the matrix in the formof helium-vacancy clusters would form bubbles upon annealing. He bubbles would perhaps lead to degradation ofmaterial properties. Especially, He bubbles along Grain boundaries (GBs) can cause embrittlement by intergranularfracture, as usually observed in metals. Therefore, it is important to investigate the nucleation and growth of Hebubbles along GBs in He-implanted SiC.展开更多
Due to its unique properties, such as favorable neutronic characteristic, low melting point(～125 ?C) and highboiling point (> 1 600 ?C), good natural circulation and chemical inertness with water and air (unlike s...Due to its unique properties, such as favorable neutronic characteristic, low melting point(～125 ?C) and highboiling point (> 1 600 ?C), good natural circulation and chemical inertness with water and air (unlike sodium), etc,lead bismuth eutectic (LBE) has been considered as perspective coolant and spallation target for accelerator drivensystems (ADS)[1;2]. The spallation structural materials of ADS will be long-term irradiated with fast neutrons whilesimultaneous being contact with LBE coolant. In the past few decades, irradiation or LBE corrosion behaviors invarious of materials have been massively investigated for the purpose of developing ADS. However, there are onlya few experiment data on the combined effect of irradiation and LBE corrosion on structural materials due to lackof related experimental setups[3;4]. Such tests are not only essential for ensuring the safety and reliability of ADS,but also necessary for building a database for licensing any materials for use in LBE cooled nuclear systems.展开更多
The RAFM (Reduced Activation Ferritic/Martensitic) steel is considered as one of the promising candidatestructural materials for LFRs (Lead alloy-cooled Fast Reactors) and ADS (Accelerator Driven Sub-critical system),...The RAFM (Reduced Activation Ferritic/Martensitic) steel is considered as one of the promising candidatestructural materials for LFRs (Lead alloy-cooled Fast Reactors) and ADS (Accelerator Driven Sub-critical system),and its compatibility with liquid metal and radiation-resistant properties have been extensively studied because ofthe requirements of reliability and safety of the blanket[1]. A number of corrosion experiments of RAFMs (Eurofer97, T91 and 316L, etc.) in liquid LiPb alloy have been investigated, and the corrosion results show that these Febasedsteels suffered more serious corrosion attack from 480 to 550 ?C, and the corrosion layer is made of the oxidelayer (Fe3O4 and CrxFe3?xO4) at steels' surface. Generally speaking, during the stage degeneration of material, theformation of corrosion layer is one of the important features of the process[2]. Cracking, blistering, embrittlementand other changes in materials may be induced by corrosion layers, and the corrosion layers have independentcompositions, structures and radiation-resistant properties with distinguished from the alloy matrix. In a word, inorder to further clarify the applicability of Fe-based structural materials in nuclear facilities, we should study notonly the RAFM steel itself but also its corrosion layer (Fe3O4, mainly). So we report on modifications of mechanicalproperties of Fe3O4 corrosion layer irradiated with high-energy ion.展开更多
When swift heavy ion (SHI) passes through metallic multilayers, the kinetic energy of the ion is mainly depositedto target electron subsystem (electronic energy loss, Se) by the inelastic collisions involving excitati...When swift heavy ion (SHI) passes through metallic multilayers, the kinetic energy of the ion is mainly depositedto target electron subsystem (electronic energy loss, Se) by the inelastic collisions involving excitation and ionizationof the target atoms, which could induce atomic displacements and modify the interfacial structure [1?6]. Therefore,through the study of the process of the interfacial atoms diffusion induced by SHI irradiation, we could explore thepossible mechanism of atomic displacement induced by swift heavy ion irradiation.展开更多
文摘The martensite steels are accepted material used in nuclear power plant.In this study,self-ion irradiation was used to simulate the damage caused by fast neutrons in two kinds of martensite steels,SIMP and T91,under the temperature of 300℃.The contrast experiment on the steel samples was carried out with 352.8 MeV Fe-ions.S parameter is a statistical conclusion about vacancy damage caused by irradiation,and it is positively related to the density of vacancy defects Figs.1 and 2 show the change of the S parameter with the irradiation dose.Whereas Figs.3 and 4 show the different S parameters for two kind of steels.
文摘Because of the low cross-section for neutron capture and its excellent structural, chemical and mechanicalstability, silicon carbide (SiC) is an important material with application in the development of nuclear energyand waste technologies. The (n,) nuclear reaction inevitably introduces numerous He in SiC. Because of the lowsolubility of He atoms in SiC, a certain concentration of He atoms that are trapped in the matrix in the formof helium-vacancy clusters would form bubbles upon annealing. He bubbles would perhaps lead to degradation ofmaterial properties. Especially, He bubbles along Grain boundaries (GBs) can cause embrittlement by intergranularfracture, as usually observed in metals. Therefore, it is important to investigate the nucleation and growth of Hebubbles along GBs in He-implanted SiC.
文摘Due to its unique properties, such as favorable neutronic characteristic, low melting point(～125 ?C) and highboiling point (> 1 600 ?C), good natural circulation and chemical inertness with water and air (unlike sodium), etc,lead bismuth eutectic (LBE) has been considered as perspective coolant and spallation target for accelerator drivensystems (ADS)[1;2]. The spallation structural materials of ADS will be long-term irradiated with fast neutrons whilesimultaneous being contact with LBE coolant. In the past few decades, irradiation or LBE corrosion behaviors invarious of materials have been massively investigated for the purpose of developing ADS. However, there are onlya few experiment data on the combined effect of irradiation and LBE corrosion on structural materials due to lackof related experimental setups[3;4]. Such tests are not only essential for ensuring the safety and reliability of ADS,but also necessary for building a database for licensing any materials for use in LBE cooled nuclear systems.
文摘The RAFM (Reduced Activation Ferritic/Martensitic) steel is considered as one of the promising candidatestructural materials for LFRs (Lead alloy-cooled Fast Reactors) and ADS (Accelerator Driven Sub-critical system),and its compatibility with liquid metal and radiation-resistant properties have been extensively studied because ofthe requirements of reliability and safety of the blanket[1]. A number of corrosion experiments of RAFMs (Eurofer97, T91 and 316L, etc.) in liquid LiPb alloy have been investigated, and the corrosion results show that these Febasedsteels suffered more serious corrosion attack from 480 to 550 ?C, and the corrosion layer is made of the oxidelayer (Fe3O4 and CrxFe3?xO4) at steels' surface. Generally speaking, during the stage degeneration of material, theformation of corrosion layer is one of the important features of the process[2]. Cracking, blistering, embrittlementand other changes in materials may be induced by corrosion layers, and the corrosion layers have independentcompositions, structures and radiation-resistant properties with distinguished from the alloy matrix. In a word, inorder to further clarify the applicability of Fe-based structural materials in nuclear facilities, we should study notonly the RAFM steel itself but also its corrosion layer (Fe3O4, mainly). So we report on modifications of mechanicalproperties of Fe3O4 corrosion layer irradiated with high-energy ion.
文摘When swift heavy ion (SHI) passes through metallic multilayers, the kinetic energy of the ion is mainly depositedto target electron subsystem (electronic energy loss, Se) by the inelastic collisions involving excitation and ionizationof the target atoms, which could induce atomic displacements and modify the interfacial structure [1?6]. Therefore,through the study of the process of the interfacial atoms diffusion induced by SHI irradiation, we could explore thepossible mechanism of atomic displacement induced by swift heavy ion irradiation.
