Activation analyses play a vital role in nuclear reactor design.Activation analyses,along with nuclear analyses,provide important information for nuclear safety and maintenance strategies.Activation analyses also help...Activation analyses play a vital role in nuclear reactor design.Activation analyses,along with nuclear analyses,provide important information for nuclear safety and maintenance strategies.Activation analyses also help in the selection of materials for a nuclear reactor,by providing the radioactivity and dose rate levels after irradiation.This information is important to help define maintenance activity for different parts of the reactor,and to plan decommissioning and radioactive waste disposal strategies.The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential,due to the presence of a highenergy neutron environment which makes decisive demands on material selection.This study comprises two parts; in the first part the activation characteristics,in a fusion radiation environment,of several elements which are widely present in structural materials,are studied.It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment.The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions.The structural materials selected for this study,i.e.India-specific Reduced Activation Ferritic-Martensitic steel（IN-RAFMS）,P91-grade steel,stainless steel 316 LN ITER-grade（SS-316 LN-IG）,stainless steel 316 L and stainless steel 304,are candidates for use in ITER either in vessel components or test blanket systems.Tungsten is also included in this study because of its use for ITER plasma-facing components.The study is carried out using the reference parameters of the ITER fusion reactor.The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port.The presence of elements like Nb,Mo,Co and Ta in a structural material enhance the activity level as well as the dose level,which has an impact on design considerations.IN-RAFMS was shown to be a more effective low-activation material than SS-316 LN-IG.展开更多
This article explores the transformative potential of nanotechnology and MMs(memory metals)in enhancing the design and operation of nuclear reactors,encompassing both fission and fusion technologies.Nanotechnology,wit...This article explores the transformative potential of nanotechnology and MMs(memory metals)in enhancing the design and operation of nuclear reactors,encompassing both fission and fusion technologies.Nanotechnology,with its ability to engineer materials at the atomic scale,offers significant improvements in reactor safety,efficiency,and longevity.In fission reactors,nanomaterials enhance fuel rod integrity,optimize thermal management,and improve in-core instrumentation.Fusion reactors benefit from nanostructured materials that bolster containment and heat dissipation,addressing critical challenges in sustaining fusion reactions.The integration of SMAs(shape memory alloys),or MMs,further amplifies these advancements.These materials,characterized by their ability to revert to a pre-defined shape under thermal conditions,provide self-healing capabilities,adaptive structural components,and enhanced magnetic confinement.The synergy between nanotechnology and MMs represents a paradigm shift in nuclear reactor technology,promising a future of cleaner,more efficient,and safer nuclear energy production.This innovative approach positions the nuclear industry to meet the growing global energy demand while addressing environmental and safety concerns.展开更多
Recently, the work function （WF） changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe （KP）, under He ion irradiation in high energy...Recently, the work function （WF） changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe （KP）, under He ion irradiation in high energy （MeV） and / or low energy （500 eV） ranges. The results of polycrystalline Ni samples indicate that the 1 MeV beam only induces decrease in the WF within the experimental fluence range; whereas the irradiation of 500 eV beam results in decrease in the WF firstly, then increase till saturation. A dual layer surface model is employed to explain the observed phenomena, together with computer simulation results by SRIM code. Charges buildup on the surface of lithium ceramics has been found to greatly influence the probe output, which can be explained qualitatively using a model concerning an induction electric field due to external field and free charges on the ceramic surface.展开更多
W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a po...W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.展开更多
Fe-Cr-Mn (W, V) austenite steel was researched in order to supply a theory base for the first wall materials of fusion reactors.Experiments included vacuum melting, forging, annealing, solution treatment, Charpy impac...Fe-Cr-Mn (W, V) austenite steel was researched in order to supply a theory base for the first wall materials of fusion reactors.Experiments included vacuum melting, forging, annealing, solution treatment, Charpy impact tests and microstructure observation. Theresults show that the imped value decreases with the test temperature decreasing. In this system, there is ductile/brittle transition. Themechanism of this decrease of the impact value is considered to be due to γ - ε transformation in sub-stable austenite steel and stoppingoverlapping sacking fault by grain boundaries in stable austenite steel.展开更多
Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic pr...Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.展开更多
The paper is a summary of Russian material studies performed in frames of activi-ties aiming at substantiation of safety of the International Thermonuclear Experimental Reactor(ITER)after 2001.Subthreshold sputtering ...The paper is a summary of Russian material studies performed in frames of activi-ties aiming at substantiation of safety of the International Thermonuclear Experimental Reactor(ITER)after 2001.Subthreshold sputtering of tungsten by 5 eV deuterons was revealed at temper-atures above 1150℃.Mechanism of globular films formation was further studied.Computations of tritium permeation into vacuum vessel coolant confirmed the acceptability of vacuum vessel cooling system for removal of the decay heat.The most dangerous accident with high-current are in toroidal superconducting magnets able to burn out a bore up to 0.6 m in diameter in the cryostat vessel was determined.Radiochemical reprocessing of V-Cr-Ti alloy and its purification from activation products down to a contact dose rate of~10μSv/h was developed.展开更多
Boron doping,combined with neutron capture in fission reactors,has been used to simulate the helium effect on fusion structural materials.However,inhomogeneous helium bubble formation was often observed due to boron s...Boron doping,combined with neutron capture in fission reactors,has been used to simulate the helium effect on fusion structural materials.However,inhomogeneous helium bubble formation was often observed due to boron segregation to grain boundaries.The excess radiation displacements due to^(10)B(n,α)^(7)Li reaction,the high-energy lithium and helium ions,also were not accounted for,which can significantly accelerate the displacements-per-atom(dpa)accumulation alongside helium production(appm).Hereby an isotopically pure^(10)B doping approach is proposed to simulate the extreme envi-ronment inside fusion reactors with a high He appm-to-dpa ratio of about 10,which is about 10^(2)×larger than in fission reactors.Computational modeling showed that~13%of total radiation displacement was induced by^(10)B(n,α)^(7)Li in the case of 1000 appm^(10)B doped Fe samples,which becomes even greater with increasing^(10)B loading.Spatially homogenous radiation damage and helium generation are pre-dicted for grain sizes less than 1 mm,even if the boron partially formed precipitates or segregates on grain boundaries.Feasibility studies with various^(10)B doping(and^(235)U-codoping)levels in research reactors showed the estimated helium generation and radiation damage would significantly mimic fusion conditions and greatly expedite fusion materials testing,from many years down to months.展开更多
The reduced activation ferritic/martenstic steel CLF-1 prepared by the Southwestern Institute of Physics in China was irradiated by helium ions with an energy of 5 keV at room temperature using an electron cyclotron r...The reduced activation ferritic/martenstic steel CLF-1 prepared by the Southwestern Institute of Physics in China was irradiated by helium ions with an energy of 5 keV at room temperature using an electron cyclotron resonance （ECR） ion irradiation apparatus. After the irradiation, the helium retention and desorption were investigated using a technique of thermal desorption spectroscopy （TDS）. The experiment was conducted with both the normal and welded samples. Blisters were observed after the helium ion irradiation, and the surface density of blisters in the welded samples was lower than that in the non-welded samples. Three desorption peaks were observed in both the non-welded and welded samples. These desorption peaks corresponded to those of blister ruptures and the helium release from the inner bubbles and the defects. The amount of helium retained in the welded samples was approximately the same as that in the non- welded samples, which was much less than other reduced activation materials, such as vanadium alloy and SiC/SiC composites.展开更多
Effects of helium implantation on silicon carbide（SiC） and graphite were studied to reveal the possibility of SiC replacing graphite as plasma facing materials. Pressureless sintered SiC and graphite SMF-800 were im...Effects of helium implantation on silicon carbide（SiC） and graphite were studied to reveal the possibility of SiC replacing graphite as plasma facing materials. Pressureless sintered SiC and graphite SMF-800 were implanted with He＋ions of 20 ke V and 100 ke V at different temperatures and different fluences. The He^＋ irradiation induced microstructure changes were studied by field-emission scanning electron microscopy（FESEM）, atomic force microscopy（AFM）, and transmission electron microscopy（TEM）.展开更多
The recent achievement of fusion ignition with laser-driven technologies at the National Ignition Facility sets a historic accomplishment in fusion energy research.This accomplishment paves the way for using laser ine...The recent achievement of fusion ignition with laser-driven technologies at the National Ignition Facility sets a historic accomplishment in fusion energy research.This accomplishment paves the way for using laser inertial fusion as a viable approach for future energy production.Europe has a unique opportunity to empower research in this field internationally,and the scientific community is eager to engage in this journey.We propose establishing a European programme on inertial-fusion energy with the mission to demonstrate laser-driven ignition in the direct-drive scheme and to develop pathway technologies for the commercial fusion reactor.The proposed roadmap is based on four complementary axes:(ⅰ)the physics of laser-plasma interaction and burning plasmas;(ⅱ)high-energy high repetition rate laser technology;(ⅲ)fusion reactor technology and materials;and(ⅳ)reinforcement of the laser fusion community by international education and training programmes.We foresee collaboration with universities,research centres and industry and establishing joint activities with the private sector involved in laser fusion.This project aims to stimulate a broad range of high-profile industrial developments in laser,plasma and radiation technologies along with the expected high-level socio-economic impact.展开更多
文摘Activation analyses play a vital role in nuclear reactor design.Activation analyses,along with nuclear analyses,provide important information for nuclear safety and maintenance strategies.Activation analyses also help in the selection of materials for a nuclear reactor,by providing the radioactivity and dose rate levels after irradiation.This information is important to help define maintenance activity for different parts of the reactor,and to plan decommissioning and radioactive waste disposal strategies.The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential,due to the presence of a highenergy neutron environment which makes decisive demands on material selection.This study comprises two parts; in the first part the activation characteristics,in a fusion radiation environment,of several elements which are widely present in structural materials,are studied.It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment.The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions.The structural materials selected for this study,i.e.India-specific Reduced Activation Ferritic-Martensitic steel（IN-RAFMS）,P91-grade steel,stainless steel 316 LN ITER-grade（SS-316 LN-IG）,stainless steel 316 L and stainless steel 304,are candidates for use in ITER either in vessel components or test blanket systems.Tungsten is also included in this study because of its use for ITER plasma-facing components.The study is carried out using the reference parameters of the ITER fusion reactor.The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port.The presence of elements like Nb,Mo,Co and Ta in a structural material enhance the activity level as well as the dose level,which has an impact on design considerations.IN-RAFMS was shown to be a more effective low-activation material than SS-316 LN-IG.
文摘This article explores the transformative potential of nanotechnology and MMs(memory metals)in enhancing the design and operation of nuclear reactors,encompassing both fission and fusion technologies.Nanotechnology,with its ability to engineer materials at the atomic scale,offers significant improvements in reactor safety,efficiency,and longevity.In fission reactors,nanomaterials enhance fuel rod integrity,optimize thermal management,and improve in-core instrumentation.Fusion reactors benefit from nanostructured materials that bolster containment and heat dissipation,addressing critical challenges in sustaining fusion reactions.The integration of SMAs(shape memory alloys),or MMs,further amplifies these advancements.These materials,characterized by their ability to revert to a pre-defined shape under thermal conditions,provide self-healing capabilities,adaptive structural components,and enhanced magnetic confinement.The synergy between nanotechnology and MMs represents a paradigm shift in nuclear reactor technology,promising a future of cleaner,more efficient,and safer nuclear energy production.This innovative approach positions the nuclear industry to meet the growing global energy demand while addressing environmental and safety concerns.
文摘Recently, the work function （WF） changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe （KP）, under He ion irradiation in high energy （MeV） and / or low energy （500 eV） ranges. The results of polycrystalline Ni samples indicate that the 1 MeV beam only induces decrease in the WF within the experimental fluence range; whereas the irradiation of 500 eV beam results in decrease in the WF firstly, then increase till saturation. A dual layer surface model is employed to explain the observed phenomena, together with computer simulation results by SRIM code. Charges buildup on the surface of lithium ceramics has been found to greatly influence the probe output, which can be explained qualitatively using a model concerning an induction electric field due to external field and free charges on the ceramic surface.
基金supported by the National Science Foundation under Grant No.CMMI-1762190The research was performed in part in the Nebraska Nanoscale Facility:National Nanotechnology Coordinated Infrastructure and the Nebraska Center for Materials and Nanoscience (and/or NERCF),which are supported by the National Science Foundation under Award ECCS:2025298+1 种基金the Nebraska Research Initiativesupported by the U.S.Department of Energy,Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517 as part of a Nuclear Science User Facilities experiment。
文摘W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.
文摘Fe-Cr-Mn (W, V) austenite steel was researched in order to supply a theory base for the first wall materials of fusion reactors.Experiments included vacuum melting, forging, annealing, solution treatment, Charpy impact tests and microstructure observation. Theresults show that the imped value decreases with the test temperature decreasing. In this system, there is ductile/brittle transition. Themechanism of this decrease of the impact value is considered to be due to γ - ε transformation in sub-stable austenite steel and stoppingoverlapping sacking fault by grain boundaries in stable austenite steel.
基金This work was supported by National Natural Science Foundation of China No.19889502.
文摘Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.
文摘The paper is a summary of Russian material studies performed in frames of activi-ties aiming at substantiation of safety of the International Thermonuclear Experimental Reactor(ITER)after 2001.Subthreshold sputtering of tungsten by 5 eV deuterons was revealed at temper-atures above 1150℃.Mechanism of globular films formation was further studied.Computations of tritium permeation into vacuum vessel coolant confirmed the acceptability of vacuum vessel cooling system for removal of the decay heat.The most dangerous accident with high-current are in toroidal superconducting magnets able to burn out a bore up to 0.6 m in diameter in the cryostat vessel was determined.Radiochemical reprocessing of V-Cr-Ti alloy and its purification from activation products down to a contact dose rate of~10μSv/h was developed.
基金supported by Nuclear Global Fellowship Program through the Korea Nuclear International Cooperation Foundation(KONICOF)funded by the Ministry of Science and ICTsupport by DTRA(Award No.HDTRA1-20-2-0002)Interaction of Ionizing Radiation with Matter(IIRM)University Research Alliance(URA).
文摘Boron doping,combined with neutron capture in fission reactors,has been used to simulate the helium effect on fusion structural materials.However,inhomogeneous helium bubble formation was often observed due to boron segregation to grain boundaries.The excess radiation displacements due to^(10)B(n,α)^(7)Li reaction,the high-energy lithium and helium ions,also were not accounted for,which can significantly accelerate the displacements-per-atom(dpa)accumulation alongside helium production(appm).Hereby an isotopically pure^(10)B doping approach is proposed to simulate the extreme envi-ronment inside fusion reactors with a high He appm-to-dpa ratio of about 10,which is about 10^(2)×larger than in fission reactors.Computational modeling showed that~13%of total radiation displacement was induced by^(10)B(n,α)^(7)Li in the case of 1000 appm^(10)B doped Fe samples,which becomes even greater with increasing^(10)B loading.Spatially homogenous radiation damage and helium generation are pre-dicted for grain sizes less than 1 mm,even if the boron partially formed precipitates or segregates on grain boundaries.Feasibility studies with various^(10)B doping(and^(235)U-codoping)levels in research reactors showed the estimated helium generation and radiation damage would significantly mimic fusion conditions and greatly expedite fusion materials testing,from many years down to months.
基金supported by National Natural Science Foundation of China (50701017)Japan-China Core University Program on Plasma and Nuclear Fusion
文摘The reduced activation ferritic/martenstic steel CLF-1 prepared by the Southwestern Institute of Physics in China was irradiated by helium ions with an energy of 5 keV at room temperature using an electron cyclotron resonance （ECR） ion irradiation apparatus. After the irradiation, the helium retention and desorption were investigated using a technique of thermal desorption spectroscopy （TDS）. The experiment was conducted with both the normal and welded samples. Blisters were observed after the helium ion irradiation, and the surface density of blisters in the welded samples was lower than that in the non-welded samples. Three desorption peaks were observed in both the non-welded and welded samples. These desorption peaks corresponded to those of blister ruptures and the helium release from the inner bubbles and the defects. The amount of helium retained in the welded samples was approximately the same as that in the non- welded samples, which was much less than other reduced activation materials, such as vanadium alloy and SiC/SiC composites.
基金supported by the ITER-National Magnetic Confinement Fusion Program,China(Grant Nos.2010GB109000,2011GB108009,and 2014GB123000)the National Natural Science Foundation of China(Grant No.11075119)
文摘Effects of helium implantation on silicon carbide（SiC） and graphite were studied to reveal the possibility of SiC replacing graphite as plasma facing materials. Pressureless sintered SiC and graphite SMF-800 were implanted with He＋ions of 20 ke V and 100 ke V at different temperatures and different fluences. The He^＋ irradiation induced microstructure changes were studied by field-emission scanning electron microscopy（FESEM）, atomic force microscopy（AFM）, and transmission electron microscopy（TEM）.
文摘The recent achievement of fusion ignition with laser-driven technologies at the National Ignition Facility sets a historic accomplishment in fusion energy research.This accomplishment paves the way for using laser inertial fusion as a viable approach for future energy production.Europe has a unique opportunity to empower research in this field internationally,and the scientific community is eager to engage in this journey.We propose establishing a European programme on inertial-fusion energy with the mission to demonstrate laser-driven ignition in the direct-drive scheme and to develop pathway technologies for the commercial fusion reactor.The proposed roadmap is based on four complementary axes:(ⅰ)the physics of laser-plasma interaction and burning plasmas;(ⅱ)high-energy high repetition rate laser technology;(ⅲ)fusion reactor technology and materials;and(ⅳ)reinforcement of the laser fusion community by international education and training programmes.We foresee collaboration with universities,research centres and industry and establishing joint activities with the private sector involved in laser fusion.This project aims to stimulate a broad range of high-profile industrial developments in laser,plasma and radiation technologies along with the expected high-level socio-economic impact.
