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The effects of the ordered and disordered arrangements of elements on radiation-induced defects production and evolution in NiFe alloys were investigated through atomistic simulations. Results present a sluggish evolution of the overall microstructure in ordered L10 NiFe. Although the disordered phase has fewer Frankel pair accumulation in cascade simulation attributed to the low thermal conductivity reduced by the intrinsic chemical disorder, the difference is negligible when PKA energy increases because of the direct formation of clusters. Interstitial diffusion is restricted in the ordered phase, where Ni and Fe layers are alternately arranged. This condition delays interstitials accumulation and leads to the formation of more Shockley partial loops rather than Frank loops which favor in the disordered phase. The higher stacking fault energy in the ordered phase renders it difficult to form stacking fault tetrahedra

Nickel–graphene nanolayers with high-density interfaces are expected to have excellent resistance to helium (He) embrittlement and proposed as candidate materials for molten salt reactor systems. However, He irradiation effects on nickel–graphene nanolayers remains poorly understood at present. In this work, the influence of a nickel–graphene interface (NGI) on the nucleation and growth of He-related clusters was studied by using atomistic simulations. The NGI reduces formation energies and diffusion energy barriers for He-related clusters. The reduction makes He-related clusters easily be trapped by the interface, thus leading to significant segregation. Consequently, He concentration in the bulk is considerably reduced, and the nucleation and growth rates of He-related clusters in the bulk are delayed. Owing to the high mobility of He-related clusters at the NGI, these clusters easily coalesce to form larger clusters than those in the bulk. A reasonable design of nanolayers may promote He releasing from materials. Results of the current study can provide fundamental support for the service life assessment of nickel–graphene nanolayers in extreme environments.

In this paper, the effects of a 150 keV proton radiation on a GaAs sub-cell of GaInP/GaAs/Ge trip junction solar cells and its radiation damage were studied. The degradation behavior of the solar cells was analyzed by theoretical device modeling combined with electro-optical characterization techniques such as external quantum efficiency, IV measurements, and photoluminescence (PL). The degradation of cell output parameters by protons was plotted as a function of the displacement damage dose. The results show that the short-circuit currents degrades less than open-circuit voltages because the base region of GaAs is severely damaged. The PL results indicated that proton irradiation exerted destructive effect on the photoelectric properties of the material. Such destructiveness was due to the numerous defects introduced by proton irradiation, which destroys the integrity of the lattice space, resulting in a decrease in the diffusion length of the minority and an increase in the surface recombination velocity. By COMSOL simulation analysis, it is found that the reason of radiation degradation of GaAs is the decrease of minority carrier time after irradiation and simulation results show that the PL results are in good agreement with the simulation results which can provide a method for the calculation of internal defects and minority carriers in multi-junction solar cells.

Selective laser melting (SLM) provides a novel path to fabricate austenitic stainless steels used in the nuclear reactors. Meanwhile, obvious differences in the microstructures of materials present between SLM and conventional process, which causes a discrepancy in the helium (He) tolerance. In present work, an austenitic stainless steel (type 316L) manufactured through SLM was irradiated by He ions at 450 °C with the concentration approximately 0.8% and then characterized via multiple methods. Results showed the special microstructure containing cellular sub-grains and nano-oxide inclusions, which formed owing to the SLM process, still distributed in post-irradiated samples. A decrease in the bubbles density, swelling rate, and hardness change has also been observed compared with conventional stainless steel. The interfaces provided by the sub-grain boundaries and nano-oxide inclusions act as effective trap sites for helium bubbles, which contributed to the enhancement of helium tolerance.

Metal–graphene nanocomposites are expected to have excellent radiation tolerance, and may become a candidate structural material for advanced fission reactors. Nevertheless, whether the structural disorder of graphene introduced by preparation or irradiation can strongly affect the radiation tolerance of the composites is still unclear. Here we investigate the radiation tolerance of nickel–graphene nanocomposite by using 300 keV helium-ion irradiation at 823 K. Results showed that the intrinsic crystalline structure of graphene would be continuously disrupted by the elevated temperature and irradiation. However, lesser crystal defects, such as lattice swelling and stacking faults, and smaller helium bubbles are observed in the composite than those in its pure counterpart. The reason may be attributed to graphene's own capability in maintaining two-dimensional structure and inhibiting the formation of large-size defects. Thus, nickel–graphene interfaces can be maintained and their role in healing radiation-induced defects is still able to play. Results of the study highlight the potential of metal–graphene nanocomposites for use as radiation tolerance materials.

The self-healing mechanism of radiation-induced defects in nickel–graphene nanocomposite is investigated by atomistic simulations. Compared with pure nickel, nickel–graphene nanocomposite has less defects remained in the bulk region after collision cascades, illustrating self-healing performance. Nickel–graphene interfaces (NGIs) serve as sinks for radiation-induced defects and preferentially trap interstitials over vacancies. Energetic and kinetic calculations reveal that the defect formation energy and diffusion barrier are reduced in the vicinity of NGIs, and the reduction are pronounced for interstitials. When NGIs are loaded with interstitials, their segregation ability on radiation-induced defects improves significantly, and the radiation-induced defects near the NGIs diffuse more easily. Especially, the vacancies (or interstitials) near the NGIs tend to annihilate (or aggregate) with the interstitials trapped at the NGIs, which only happens at the interstitial-loaded side of NGIs. Therefore, nickel–graphene nanocomposite exhibits excellent radiation tolerance and shows promise as a structural material for advanced nuclear reactors due to its NGIs with the energetic and kinetic driving forces acting on radiation-induced defects.

Nickel–graphene (N–G) composites are potential candidate structural materials for molten salt reactors. A rapid preparation method for these composites by jet electrodeposition was developed, and the microstructure, microhardness, and corrosion properties of these composites were studied to explore the key parameters of the jet electrodeposition. Results indicated that the distribution of graphene in composites mainly depended on the concentration of graphene oxide (GO) in the plating solution. Composites deposited with GO concentration of 0, 0.5, 1 g/L showed surface root-mean-square roughness value (Rq) of 6, 12, and 28 nm, respectively. Meanwhile with the increase of GO concentration, the Hardness value became larger. The corrosion potential Ecorr and current Icorr of composites obtained at 0.5 g/L with the best surface quality were 193 mV and 5.7 × 10−6 A/cm2, respectively, which indicated the best electrochemical corrosion resistance. Hydrogen annealing can help self–repair of graphene microstructure.

This work designed and measured a remote sprayed fast-curing γ-radiation-shielding material that is used to encapsulate and collect radioactive wastes from places far away from high radiation areas in case of radioactive waste leakage accidents. The influence of the chemical composition design on curing time, shielding performance, thermal stability and mechanical performance of the γ-radiation-shielding material is discussed. Results showed that with the addition of catalysts, the cream time of the γ-radiation-shielding material increased remarkably to allow it to be ejected, while the tack-free time of the solidified composite material decreased to guarantee good encapsulation of the radioactive wastes. Due to the denser microstructure, the γ-radiation-shielding material with bis(2-dimethylaminoethyl) ether and phosphorus flame retardant V-490 show better γ-radiation-shielding performance and mechanical properties than other products. When the shielding functional filler content reached 60%, the composite presented excellent comprehensive properties and a good application prospect.

With a highly functional methyl vinyl silicone rubber (VMQ) matrix and filler materials of B4C, PbO, and benzophenone (BP) and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were 323.6°C and 335.3°C, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am–Be neutron source. The transmission of γ-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively. Keywords: Flexible Composite, Neutron Shielding, Radiation Resistance, γ-ray Shielding

This work shows that mechanical properties, thermal conductivity, and secondary gamma-ray shielding ability can be significantly improved when open-cell nickel foams are embedded into the shielding composites. The boracic polyethylene/polyethylene wax blends and open-cell nickel foam composites (PPNM) are designed and prepared by permeating homogeneous mixed melt fillers into open-cell nickel foams. The ratio of polyethylene and polyethylene wax is investigated to achieve higher filling rate. The quasi-static compressive response of PPNMs and polyethylene/polyethylene wax blends is measured, and the crystallization properties are studied by differential scanning calorimetry. The neutron and secondary gamma-ray shielding abilities of PPNMs are also simulated based on Monte Carlo particle transport method. Results show that the compression strength of PPNMs with boron carbide is slightly improved when compared with polyethylene/polyethylene wax blends. The nickel foams in PPNM composites improve the energy-...

Metal–graphene nanocomposites are expected to have excellent radiation resistance. The intrinsic role of the graphene layers (GrLs) in their performance has not been fully understood. Five copper–graphene nanocomposite (CGNC) systems were used to investigate the detailed mechanisms underpinning this behaviour by atomistic simulation. Results showed that GrLs can reduce the formation, growth, and intensity of the thermal spike of CGNC; this effect became more evident with the increasing number of layers of graphene. The role of the GrLs can be explained by three mechanisms: first, the ultra-strength C–C bonds of graphene hindered the penetration of high-energy atoms, second, the number of recoiled atoms decreased with the increasing number of layers of graphene, and third, the energy dissipation along the graphene planes also indirectly weakened the damage caused to the entire system. These mechanisms may provide a pathway to prevent material degradation in extreme radiation environments.

In consideration of the ubiquity of swift particles in space environment, understanding the effects of ion irradiation on the adhesion force and mechanical properties of multi-walled carbon nanotube arrays is significant to apply the arrays as adhesive materials in space in the future. In this study, multi-walled carbon nanotube arrays prepared through chemical vapor deposition were irradiated by He2+ under different fluences and then characterized via multiple methods. The arrays were severely damaged when the fluence reached or exceeded 1 × 1016 cm−2. As the fluence increased, more amorphous carbons were generated from their original position in the carbon nanotubes. The adhesion force quickly decreased when the irradiation fluence rapidly increased, which limited the service life of the arrays as adhesive materials. Finally, the instantaneous modulus of the multi-walled carbon nanotube arrays initially increased and then decreased when the fluence increased because of the different contact modes between the indenter and the samples.

Traditional neutron-shielding materials usually have poor mechanical properties and secondary gamma-shielding capability. The new requirements of modern neutron-shielding materials are difficult to satisfy. A paraffin/nickel foam neutron-shielding composite was prepared and characterized in this study. Open-cell nickel foams were fabricated through electrodeposition. Subsequently, the paraffin/nickel foam composite were prepared by filling the open-cell nickel foams with melted paraffin. The intrinsic parameters of nickel foam and the content of neutron absorber (boron carbide) were controlled to optimize the composite. The mechanical properties of the composite were studied through a static compression test. The compressive strength improved to 0.4 times that of the nickel foams. The Am–Be source transmittance experiment showed that the 8 cm thick PFM presented a neutron transmittance of 56.1%, and the 6 cm thick boron carbide/paraffin/nickel foam (PFM-B) presented a neutron transmittance of 37.6%. The paraffin/nickel foam and PFM-B had approximately the same shielding efficiency as paraffin and boron carbide/paraffin, respectively. However, the second gamma ray shielding efficiency of the paraffin/nickel foam and PFM-B was significantly higher than that of paraffin and boron carbide/paraffin. The mechanical properties and secondary gamma ray-shielding capability of the composite can be improved by increasing the relative density of nickel foams.

High temperature radiation damage in binary bcc Fe alloys containing 1 atomic % Au or Cu due to Fe ion irradiation at 550 °C to a peak dose of 2.8 and 8.3 dpa is studied. The precipitation behavior of gold and copper and its correlation to the irradiation-induced defects is studied by transmission electron microscopy and variable energy positron annihilation spectroscopy (VEPAS). The increase of S parameters from VEPAS indicates the formation of open volume defects upon irradiation. Disc-shaped Au precipitates, grown from the irradiation induced dislocations, are observed in the Fe-Au alloy. In the Fe-Cu alloy, spherical Cu particles are formed but no direct connection between Cu precipitates and radiation damage is detected. For the Fe-Au alloy, the surface hardness dramatically increases for a dose of 2.8 dpa, with a slight decrease as the irradiation dose is enhanced to 8.3 dpa. In the Fe-Cu alloy, radiation hardening increases continuously.

It is of great significance to provide theoretical guidance to the application of carbon nanotubes as adhesive materials via the investigation on their adhesion force. In this paper, molecular dynamics was adopted to investigate the adhesion force between the graphene substrate and the carbon nanotubes with varying deformation degrees and diverse types of irradiation induced defects at different temperatures. No obvious adhesion force was found between the carbon nanotube and graphene until the deformation degree of the former reached > 70%. The adhesion force would be maintained at a high level when the temperature was in the range of 280–320 K, which limited its application. Moreover, the adhesion force between carbon nanotube with vacancies and graphene substrate would decrease with increasing size of vacancies. Finally, compared with monovacancies and divacancies, Stone-Wales defects most remarkably reduced the adhesion force of carbon nanotubes.

The radiation damage and microstructure evolution of different zigzag single-walled carbon nanotubes (SWCNTs) were investigated under incident carbon ion by molecular dynamics (MD) simulations. The radiation damage of SWCNTs under incident carbon ion with energy ranging from 25 eV to 1 keV at 300 K showed many differences at different incident sites, and the defect production increased to the maximum value with the increase in incident ion energy, and slightly decreased but stayed fairly stable within the majority of the energy range. The maximum damage of SWCNTs appeared when the incident ion energy reached 200 eV and the level of damage was directly proportional to incident ion fluence. The radiation damage was also studied at 100 K and 700 K and the defect production decreased distinctly with rising temperature because radiation-induced defects would anneal and recombine by saturating dangling bonds and reconstructing carbon network at the higher temperature. Furthermore, the stability of a large-diameter tube surpassed that of a thin one under the same radiation environments.

High-entropy alloys (HEAs) have become newly emerging candidates as structural materials of advanced fission reactor because of their excellent mechanical properties and irradiation resistance. Recently, carbon doped HEAs exhibited improved mechanical properties, such as yield strength and elongation. However, the effects of carbon doping on the irradiation resistance of HEAs need further investigation. Here, the irradiation-induced defects and irradiation hardening of Fe38Mn40Ni11Al4Cr7 HEA with different carbon contents were investigated by using 5 MeV Xe23+ heavy-ion irradiation at room temperature, and multiple characterization methods were used to provide the essential evidences. Results showed that the carbon doped samples exhibited smaller-sized dislocation loops and significantly lower hardening rate than those of undoped samples. The reason is attributed to two aspects: Firstly, interstitial carbon would significantly increase lattice distortion and migration energy of self-interstitial atoms, thereby inhibiting the formation of defects. Secondly, carbon atoms would act as obstacles that hindered the evolution of defects. Consequently, our study indicated the potential of using carbon doped HEAs as irradiation-resistant materials.

Grain boundary (GB) engineering is crucial in the austenitic stainless steel (ASS) design for nuclear energy applications. In this work, the influence of different GB structures on radiation defect recombination and radiation-induced segregation (RIS) at different temperatures were investigated using molecular dynamics simulation. Four typical GBs in ASSs were selected as model structures. Results showed that GBs remained stable at various temperatures and they all exhibited better self-healing performance than single crystals in terms of radiation defects. However, except Σ3(112) GB, other three GBs cannot inhibit the radiation induced segregation, while promoting the radiation defect recombination. Calculation results showed that the higher Σ value of GBs can lead to a greater lattice mismatch near GBs, which not only results in stronger sink strength for radiation induced defects, but also provides more sites for solute atoms and causes greater segregations eventually. Owing to the intrinsic low Σ and large inclination angle characteristic, Σ3(112) GB achieves an excellent balance between the defect-absorption and RIS. This phenomenon provides a feasible route for the future GB design in ultra-high radiation tolerant materials.

Additive manufacturing technology is a novel path to fabricate complex nuclear components. Considering the hydrogen (H) environment in which nuclear materials work, we proposed that the hydrogen resistance of structure materials should be considered as a deciding factor in the materials selection for nuclear application. In this work, we introduced H atoms into the selective laser melting (SLM) 316L stainless steel (SS) via two different H permeation methods and characterized its H permeation behavior and microstructure change. Results showed that the H diffusion rate in the SLM 316L SS was much higher than that in the cold-rolled (CR) one. The abundant sub-grain boundaries of the SLM 316L SS act as rapid transportation channels for the H atoms. H-induced defects which do not appear in the H charged CR 316L SS were observed in the H charged SLM 316L SS. H softening effect occurs in both SLM and CR 316L SS.

Additive manufacturing (AM) has remarkable application potential in the field of nuclear engineering materials. Although considerable effort has been exerted to improve the mechanical properties of additive-manufactured materials, research on radiation-induced degradation of AM materials must also be conducted to meet the needs of nuclear engineering. This paper describes the radiation-phase stability of 316L stainless steel (SS) fabricated via selective laser melting (SLM) technology and the influence of microstructure change on the swelling and hardening of materials. Results showed that the austenite phase in SLM 316 SS was more stable than that of cold-rolled (CR) 316L SS. However, the swelling of SLM 316L SS occurred earlier than that of CR 316L SS in the room temperature (RT) case, while no distinct difference was observed between the two materials in the 350 °C case. Two 316L SSs behave differently in swelling, due to their difference in misorientation and density of grain boundaries and dislocations. In addition, the hardening of CR 316L SS was stronger than that of SLM 316L SS. Analysis indicates that the inhibiting effect on radiation-induced martensite formation in SLM 316L SS leads to the improved resistance to the radiation hardening.

We investigated the radiation damage resistance of the Fe/Ni multilayer nanocomposites by molecular dynamics. In the paper, two types of interface configuration with different orientation relationship were constructed. Their morphology evolution and number of final surviving defects induced by cascade collisions were discussed respectively. The interfaces of the two types of Fe/Ni multilayers kept distinct during the long-time relaxation before cascade. The comparison of surviving defects number produced by PKA with 5 keV at 100 K showed that the Fe/Ni multilayers have greater radiation tolerance than that of the bulk materials. However, the orientation relationship of the interface influences the defects self-healing capability greatly when the multilayers are irradiated by higher energy PKA or at high temperature. The radiation damage resistance of the Nishiyama – Wassermann type Fe/Ni multilayers which have larger lattice misfit is more stable than that of the Kurdjumov – Sachs type.

Metal–graphene (Gr) nanocomposites are regarded as potential anti-irradiation advanced materials for Gen-IV nuclear reactors. In this work, a phase-field model was proposed to simulate helium (He) bubble evolution of nickel–Gr (Ni/Gr) nanocomposite during constant irradiation at 800 K. Ni/Gr nanocomposite presented an earlier onset time of He bubble nucleation but eventually formed bubbles with smaller size and lower density than those of pure Ni. The early He bubbles nucleated along Ni/Gr interfaces, while a He depletion zone formed nearby consequently, which confirmed that Ni/Gr interfaces inhibit He bubble growth. The effects of vacancies and He atoms on the He bubble growth rate were also quantitatively analyzed. The amount of irradiation-induced vacancies acted as an effective rate-controlling factor that determined the bubble growth rate with respect to the He atoms.

Metal–graphene nanocomposite is a kind of potential radiation tolerant material. Graphene damage of the composite is inevitable within radiation environments. In this paper, two kinds of copper–graphene nanocomposite (CGNC) systems containing perfect graphene and prefabricated damage graphene, respectively, were adopted to expound the influences of graphene damage on the properties (radiation-resistance and configuration) of CGNC under irradiation by atomistic simulations. In the CGNC containing perfect graphene, the increasing graphene damage induced by the increase of the number of cascades, did not obviously impair the role of copper–graphene interface in keeping the properties of CGNC. In the CGNC containing prefabricated damage graphene, the properties of CGNC would significantly deteriorate once the radius of prefabricated damage exceeds 10 Å, and even stacking fault tetrahedral would occur in the CGNC. The results highlighted that prefabricated graphene damage have greater effects on the change of the properties of CGNC. Therefore, it is very necessary to maintain the structural integrity of graphene for improving the radiation-resistance and configuration of CGNC.

Cr/W multilayer nanocomposites were presented in the paper as potential candidate materials for the plasma facing components in fusion reactors. We used neutron reflectometry to measure the depth profile of helium in the multienergy He ions irradiated [Cr/W (50 nm)]3 multilayers. Results showed that He-rich layers with low neutron scattering potential energy form at the Cr/W interfaces, which is in great agreement with previous modeling results of other multilayers. This phenomenon provided a strong evidence for the He trapping effects of Cr/W interfaces and implied the possibility of using the Cr/W multilayer nanocomposites as great He-tolerant plasma facing materials.

Novel, flexible, lead-free X-ray-shielding composites were prepared with a high-functional methyl vinyl silicone rubber (VMQ) matrix with W and Bi2O3 as filler materials. To verify the advanced properties of the lead-free material, composites with the same mass fraction of PbO were compared. With the X-ray energy ranging from 48 to 185 keV, the W/Bi2O3/VMQ composites exhibited higher X-ray-shielding properties. As the filler volume fraction decreased, the tensile strength, elongation, tear strength, and flexibility of the W/Bi2O3/VMQ composites increased. The Shore hardness of the W/Bi2O3/VMQ composites had a maximum value of 46.6 HA and was still very flexible. With decreasing filler volume fraction, the water-vapor transmission performances of the W/Bi2O3/VMQ composites increased, and the W/Bi2O3/VMQ composites also showed better water-vapor permeability. The heat-transfer properties of the W/Bi2O3/VMQ composites increased with increasing W content, and when the W content exceeded 70 wt %, the thermal conductivity of the W/Bi2O3/VMQ material was about 70.45% higher than that of the PbO/VMQ composite. © 2015 Wiley Periodicals, Inc. J. Appl. Polym. Sci. 2015, 132, 43012.