In current work, irradiation-induced microstructural evolution of the Cansas 3303 SiC fiber was investigated. This fiber was irradiated by 1.8 MeV argon ions at room temperature in the Shanghai Institute of Applied Physics, Chinese Academy of Sciences. The ion fluences were 5 x 1014, 1 x 1015, 2.5 x 1015 and 5 x 1015 ions/cm2 corresponding to the maximum doses of 0.27, 0.54, 1.35 and 2.70 displacements per atom (dpa), respectively. Irradiation effects were characterized by using micro-Raman spectroscopy and transmission electron microscopy. The results showed that the Cansas 3303 fiber consists of nanocrystalline SiC grains, carbon packets and carbon ribbons. Irradiation-induced amorphization accompanied with the decrease of SiC grain size was observed in the fiber with the increase of ion fluences. However, SiC grains still remained in this fiber up to the dose of 2.70 dpa. Besides, the reduction of the quantity and size in carbon packets and the annihilation of carbon ribbons in this fiber were also detected.

BackgroundWithin GEN-IV reactors, nuclear graphite plays a crucial role as both a moderator and reflector in an environment characterized by high temperatures and intense fast neutron irradiation. The exposure to fast neutron irradiation induces the formation of numerous Frankel defects in the nuclear graphite. These defects undergo processes of annihilation and diffusion, ultimately giving rise to larger defect clusters. This transformation in the microstructure of nuclear graphite directly impacts its macroscopic properties, necessitating a thorough investigation.PurposeThis study aims to comprehensively explore the evolution of defects in nuclear graphite under conditions of high-temperature irradiation which is essential for advancing reactor safety.MethodsFirstly, the 30 MeV 107Ag5+ ion source was employe to irradiate IG-110 nuclear graphite at 420 ℃, simulating the defect evolution behavior during fast neutron irradiation of nuclear graphite. Then, the energy loss, defect distribution, and ion implantation profiles of 30 MeV 58Ni5+ and 107Ag5+ ion beams bombarding standard nuclear graphite ICRU-906 (density of 2.26 g∙cm-3, displacement energy of 28 eV) were calculated using the full cascade damage model in the SRIM (Stopping and Range of Ions in Matter) software. The cross-sectional structure of IG-110 nuclear graphite was characterized using micro-Raman spectroscopy. Finally, the relationship between the Raman spectroscopic features at various depths of IG-110 nuclear graphite and the irradiation damage dose was compared to investigate the evolution of IG-110 nuclear graphite microstructure with increasing irradiation damage dose (Displacements Per Atom, DPA).ResultsWith the increase in particle fluence, the characteristic parameters of the Raman spectra of nuclear graphite, including the ID/IG ratio (the ratio of the D peak height to the G peak height), the Full Width at Half Maximum of the G peak (FWHM(G)), and the shift of the G peak, all show significant increments. Compared to samples irradiated with 58Ni5+ at the same irradiation damage dose, the graphite Raman spectra irradiated with 107Ag5+ demonstrate higher ID/IG ratios and FWHM(G). At the same FWHM(G) level, the ID/IG ratio of the graphite Raman spectra irradiated with 107Ag5+ is greater than that of the samples irradiated with 58Ni5+.ConclusionsThe results of this study suggest that irradiation with heavier ions induces a higher rate of defect accumulation in nuclear graphite, leading to a more rapid reduction in graphite grain size and promoting the progression towards nanocrystallization.

Carbon/carbon (C/C) composites are promising structural materials for molten salt reactors (MSRs) because of their exceptional high-temperature performance, resistance to fluoride salt corrosion, and low-neutron- absorption cross-section. However, fluoride salts can infiltrate the pores of C/C composites, potentially causing local hotspots and accelerating material degradation. In this study, a novel C/C composite was fabricated from mesophase-pitch-based carbon fibers and a pitch-based carbon matrix via hot isostatic pressing (HIP). To evaluate its performance under MSR conditions, the composite was exposed to molten FLiBe salt at 700 degrees C and 2-9 atm for a duration of 20 h. Weight gain, morphological characteristics, and crystallinity were characterized after the salt infiltration test. Compared with C/C composites fabricated via conventional chemical vapor infiltration (CVI), the HIP-C/C composite exhibited much better molten salt barrier properties owing to its compact structure and small pore diameter. The weight gain of the HIP-C/C composite was only 0.17 wt.% under 5 atm, significantly lower than the critical index proposed for carbon materials used in MSR. Morphological characterization revealed that FLiBe salt particles were rarely trapped in the small pores of the HIP-C/C composite. While the crystallinity of the salt-impregnated CVI-C/C composite increased, the HIP-C/C composite exhibited a slightly decreased degree of graphitization after salt infiltration. These findings provide insights into the design of high-performance C/C composites for applications in MSR systems.

在第四代反应堆中，核石墨作为慢化体和反射体材料服役于高温和高通量的快中子辐照环境中。快中子辐照会在核石墨中产生大量的弗伦克尔缺陷对。这些缺陷经过湮灭、扩散、最终形成更大的缺陷团簇，从而改变核石墨的微观结构，进而改变核石墨的宏观性能。因此，研究核石墨在高温辐照条件下的缺陷演化行为和机理对提高反应堆安全性具有重要意义。本研究采用30 MeV的107Ag5+离子在420 ℃下辐照IG-110核石墨来模拟核石墨在快中子辐照过程中的缺陷演化行为。通过微区拉曼光谱对IG-110核石墨截面结构进行表征，并对比IG-110核石墨不同深度处的拉曼光谱特征参数和辐照损伤剂量之间的关系，研究IG-110微观结构随辐照损伤剂量(DPA, Displacements Per Atom)的演化行为。研究结果表明，随着注量的增加，核石墨拉曼光谱的特征参数D峰高度与G峰高度比值(ID/IG)、G峰半高宽(FWHM(G), Full Width at Half Maximum of the G peak)以及G峰的偏移量都显著增加。与58Ni5+辐照样品相比，相同辐照损伤剂量下， 107Ag5+辐照的石墨拉曼光谱的ID/IG和FWHM(G)更大。相同的FWHM(G)下，107Ag5+辐照的石墨拉曼光谱的ID/IG比58Ni5+辐照样品大。这些结果说明更重的重离子辐照会在核石墨中引起更高速率的缺陷积累，从而更快地导致石墨晶粒尺寸变小，并促进纳米晶化进程。

Background3C-SiC (β-SiC) exhibits outstanding electrochemical properties and radiation resistance, surpassing hexagonal-phase silicon carbide in irradiation resistance. As a promising candidate for the next generation of structural materials in nuclear applications and high-performance precision electronic devices for challenging reactor environments, the material has been garnering significant attention in recent decades. Within this realm, the exploration of one-dimensional silicon carbide nanomaterials has become a focal point in silicon carbide materials research. However, their practical applications have been hindered by challenges such as the absence of effective nanomaterial processing methods and processing complexities. Notably, ultrasonic processing technology has demonstrated effectiveness in addressing these challenges.PurposeThis study aims to synthesize and study 3C-SiC nanowires (NWs), investigating their ultrasonic fracture behavior for comprehensive understanding of the ultrasonic fracture characteristics of 3C-SiC NWs, laying the groundwork for basic research in the processing of one-dimensional SiC nanomaterials.MethodsFirstly, silicon carbide nanowires were prepared by chemical vapor deposition. Then the silicon carbide nanowires were characterized by microstructure observed by scanning electron microscope (SEM), transmission electron microscope (TEM), X-Ray diffraction (XRD) and Raman spectrum. Subsequently, the 3C-SiC nanowires were subjected to ultrasonic treatment, and the average length-to-diameter ratios of the ultrasonically treated nanowires were statistically analyzed to elucidate the effect of ultrasonic treatment on the nanowires. Finally, the strength of the silicon carbide nanowires was estimated by combining the bubble jet model and statistical data.ResultsThe findings reveal that the synthesized 3C-SiC NWs are predominantly of the 3C-SiC phase, exhibiting a notable presence of stacking faults. Ultrasonic treatment significantly influences the SiC NWs, leading to a noticeable reduction in the average Length-Diameter ratio, stabilizing at 18 post-treatment.ConclusionsThe observed results align with the effects of bubble jetting and are corroborated by the ultrasonic fragmentation behavior of 3C-SiC NWs. These findings offer valuable insights for the manipulation of nanomaterial size and morphology. This study provides a new perspective for the ultrasonic cutting of silicon carbide nanowires and the strength research of nanowires, and is of great significance for the future application of silicon carbide nanowires in the field of nuclear energy.

Two promising reactor Generation IV designs employ nuclear graphite as their moderator and reflector. To this end, new graphite grades are being developed, which will require their suitability for reactor design to be established. Understanding graphite irradiation behavior, especially dimensional change, is critical when selecting a particular nuclear graphite grade. This is usually established by fast neutron irradiation experiments in a material test reactor (MTR). However, due to the limited accessibility to the MTR irradiation facility, cost, and the time required for such experiments to be carried out, the development of new nuclear graphite grades is restrictive. Here, a novel approach is proposed to enable suitable grades to be selected from a number of candidates before an MTR program is implemented by estimating the irradiation-induced dimensional change behaviors of several nuclear graphite grades based on the bending of graphite foils under ion irradiation. The method is capable of efficiently distinguishing the most dimensionally stable graphite grades in the laboratory without the need to irradiate many samples in an MTR program.

In most molten salt reactor (MSR) designs, the nuclear graphite core directly contacts the molten fuel salt. The infiltration of molten fuel salt has become a primary concern of nuclear graphite for MSR. As the existing nuclear graphite grades can hardly fulfill the demand, various graphite manufacturing and post-treatment technologies were attempted to meet the molten salt reactor design target. Here, a novel approach to achieving the molten salt impermeable nuclear graphite by pore-filling with carbon foam was introduced, which employs vacuum-pressure infiltrating nuclear graphite with low-viscosity diluted resorcinol-furfural solution, followed by a series treatment to convert the solution in graphite pore to carbon foam. The results show that the pore diameter of the pore-filled graphite can be tunable, and the threshold infiltration pressure of FLiBe salt can be significantly increased, as shown by mercury intrusion and the molten salt infiltration experiment. Moreover, treated with low-modulus carbon foam, the graphite sample’s stress-strain curve and thermal expansion behavior were almost unchanged. Finally, The low modulus of the filled carbon foam may be capable of accommodating the dimensional change of the graphite under fast neutron irradiation, which makes the method promising in treating nuclear for MSR.

A sample of highly orientated pyrolytic graphite (HOPG) was implanted with Cs+ ions in order to study the effect on the graphite matrix and the chemical form of Cs once embedded. The mean implantation depth was calculated to be-22 nm, allowing for investigation with X-ray Photoelectron Spectroscopy, (XPS), where both a traditional Al K alpha X-ray source and a higher energy Ag L alpha source was used for an increased sampling depth deeper into the surface. Analysis found that there was a reduction of sp2 C-C bonding with increasing implantation dose, resulting in a total loss of sp2 character at-6 atomic % Cs+. A striking increase in higher binding energy components (285.8-289.4 eV) in the C 1s spectra cannot be attributed solely to the presence of C-O species, instead indicating a dramatic re-ordering of the graphitic lattice to accommodate and neutralize the embedded Cs. XPS analysis indicates the formation of cesium carbonate within the graphitic material; this is reinforced by analysis of a Cs2CO3 reference sample. Additionally, thermodynamics equilibrium simulations, supported by simplified density functional theory (DFT) calculations, are performed, leading to strong agreement with the XPS findings; that is, the most stable form of Cs within the graphite matrix is expected to be cesium carbonate (Cs2CO3), which is in thermodynamic equilibrium with degraphitised HOPG and oxygenated degraphitised HOPG.

A 30 MeV Ni-58(5+) ion beam was used to irradiate fine-grained graphite grade IG110 and ultrafine-grained graphite grade G1 at 400 degrees C to study their microstructure evolution under irradiation. Taking advantage of the depth dependence of the damage rate and cumulative damage of ion irradiation, the microstructure change exploration of graphite with multiple damage levels within a single fluence specimen was achieved by characterizing the sample cross-section with HR-TEM and micro-Raman. The Raman 2D maps of the cross-sections of the graphite samples irradiated with various fluences and displacement damages up to similar to 18 displacements per atom were analyzed. Evidenced by the saturation of the intensity ratio of graphite Raman D and G band (I-D/I-G), the irradiation damage and annealing equilibrium was observed. Moreover, I-D/I-G in combined with the full width at half maximum of the G band (FWHM (G)), shows an inverse evolution compared with the graphitization process, which is also supported by the HR-TEM observation. Demonstrated by the increasing rate of I-D/I-G with FWHM (G) and the saturation I-D/I-G, the microstructure changes of fine-grained graphite IG110 and ultrafine-grained graphite G1 was distinguished. (c) 2021 Elsevier B.V. All rights reserved.

BackgroundThe control rod guide tube (CRGT) is needed to facilitate the movement of the control rod in the molten salt reactor (MSR). Graphite material is resistant to neutron radiation, high temperature and molten salt corrosion, and has small neutron absorption cross section, hence it is an ideal CRGT material except for the comparatively low strength and toughness. It is believed that reinforcing the graphite CRGT with carbon fiber fabric (CFF) is a promising way to promote its application in molten salt reactor.PurposeThis study aims to optimize the preparation parameters of the CFF reinforcement for CRGT.MethodsFirst of all, the CFF was pre-impregnated with precursor solution in a vacuum environment, and wound on the graphite tube with a certain tension. Secondly, samples with different densification cycles and winding layers were prepared through multiple pressure impregnation, curing and carbonization cycles of the Precursor-Infiltration-Pyrolysis (PIP) process. Then, the mechanical properties of carbon fiber cloth reinforced graphite tube samples and the reference samples were tested. Finally, the microstructure and failure mode of the damaged samples were analyzed for further optimizing the preparation process to improve the material performance.Results & ConclusionsThe results show that the CFF winding and PIP method enhances the strength and toughness of the graphite tube, and the preparation parameters have been optimized that make the reinforced graphite CRGT with carbon fiber fabric (CFF) possible to be applied to MSR.

Knowledge about the microstructure of nuclear graphite is critical to the understanding of its irradiation behavior in the reactor. Using micro X‐ray diffraction (μXRD) two‐dimensional (2D) maps, the crystallite character of IG110 and reactor pebble graphite was characterized at the submillimeter range with a spatial resolution of about 5 μm. Various structures in the nuclear graphite were identified by comparing the X‐ray diffraction peak intensity, position, and full width at half maximum (FWHM) 2D maps. The two‐peak feature of the FWHM histograms seen in pebble graphite may be related to the raw coke and natural graphite used as raw materials. With these results, it can be concluded that the μXRD 2D map is an effective method to characterize the microstructure of nuclear graphite.

As an outstanding material for moderator and reflector, graphite is broadly applied in nuclear reactors. Irradiation will change the pore structure and content, which is intimately related to the properties of graphite. However, the irradiation effect of nuclear graphite on pores requires a more profound understanding, especially the evolution on a three-dimensional(3D) view. In this work, FIB-SEM tomography intuitively reveal the pore structure and porosity evolution of ion irradiated graphite grades IG110 in a 3D view, which was irradiated with 30 MeV Ni5+ ion at 400 degrees C. Besides, the quantitative statistic of porosity indicates a negative correlation between porosity with the ion irradiation damage and confirms the consistency of ion deposition sites and irradiation damage peak. Moreover, individual microcracks of the different ion irradiated depths and virgin graphite are extracted and displayed in various 3D perspectives, which exhibit a much more complex morphology than lenticular structure and demonstrates the closed trend of the microcracks under irradiation. Through the display of 3D pore structure and quantitative statistics of porosity, aims to provide a further understanding of ion irradiation behavior of nuclear graphite. (C) 2021 Elsevier B.V. All rights reserved.

The Brazilian disc test has been used widely to measure the tensile strength of brittle materials, such as rock and ceramic. However, the Brazilian disc method developed for rock has not been applicable for nuclear graphite due to the large ratio of tensile to compressive strength in the material. According to the results of stress analysis and Griffith’s strength criteria, the graphite compressive stress near the contact area must be decreased for the Brazilian tensile strength method. In this paper, the Brazilian disc test has been modified by using a newly developed anvil with a V-shaped cavity. According to finite element analysis, the stress concentration around the contact area between graphite samples and anvils was significantly reduced by the new design of anvils. The design has been proved to be effective by comparing with the direct methods. The method also has been proved to be valid for different samples sizes considered in this study, which indicates that it has a wide applicability.

As one of the six General IV nuclear reactors, the molten salt reactor (MSR) has received an increasing attention around the world owing to its high energy conversion efficiency, inherent safety feature, and application of the thorium fuel. However, major components for MSRs are facing great challenges due to the extreme environments of high temperature, corrosive molten fluoride salt, and neutron irradiation inside the MSRs. In this paper, requirements for the MSR materials are analyzed; the development history, current status, and existing problems of the key MSR materials (including alloy structure materials and nuclear graphite) are depicted; key technical problems to be solved during the current construction of experimental MSRs are pointed out; and the development strategy for the MSR materials is proposed, based on the research and development plan for the MSRs.

Binderless nanopore-isotropic graphite (NPIG) produced from mesocarbon microbeads by isostatic pressing method was irradiated with 7 MeV Xe26+ to a total peak dose of 0.1, 0.5, 2.5 and 5.0 dpa. The effect of the irradiation on the microstructure and physical properties of NPIG was then evaluated and compared against the performance of isostatic nuclear graphite (IG-110, TOYO TANSO CO., LTD). Ion-irradiation has a different effect on the microstructure (crystallinity, crystallite size) and mechanical properties (hardness, Young's modulus) of NPIG and IG-110 graphite. At lower irradiation doses, the surface morphology of NPIG was fragmented in a similar way to that of IG-110, but the NPIG gradually balled at 5 dpa. X-ray diffraction results show that NPIG has a lower degree of graphitization than IG-110. Raman studies indicated that NPIG reached saturation at lower doses. The nanoindentation showed that the hardness and Young's modulus of the NPIG and IG-110 increased after irradiation. Transmission electron microscopy images also provide clear evidence for an irradiation-induced increase in the number of basal dislocations and defects. Thus, although NPIG is generally more sensitive to irradiation than IG-110, its hardness is actually less affected.

Graphite has been identified as a structural and neutron-moderator material in the molten salt reactor (MSR). In order to support the structural material development for MSR, the influence of molten salt infiltration on the mechanical strength of graphite was studied. The high-temperature mechanical strength of graphite grades IG-110 and NG-CT-10 decrease with the increasing of weight gain ratio during the molten FLiNaK salt infiltration. It shows a noticeable high-temperature mechanical strength degradation, which can be evidenced by the softening coefficients. The smaller softening coefficient of compressive strength compared with that of tensile strength at the same infiltration pressure indicates the compressive strength is more vulnerable to the molten salt infiltration than the tensile strength. The major fragments analysis of the samples after compressive strength tests shows that the molten FLiNaK salt infiltrated graphite samples experience are longitudinal splitting fracture rather than the shear fracture of virgin graphite. The increase of thermal strain after molten salt infiltration indicates an extra stress in infiltrated graphite, which may be one of the reasons for the failure mechanism change.

The massive molten salt-graphite interface produced by liquid infiltration or vapor penetration in a molten salt reactor suggests the importance of irradiation-induced mixing and diffusion. We employed He+ ion beam irradiation to simulate neutron irradiation-induced mixing of Na and K in graphite. The depth distributions of both Na and K were measured by RBS (Rutherford backscattering spectrometry) and XPS (X-ray photoelectron spectroscopy) depth profiling. Both the mixing rate and the diffusion rates of the Na and K mark layers in the graphite induced by the ion beam irradiation were calculated. The results indicate that the diffusion rates of Na and K in the graphite reflector and/or moderator were several hundred nanometers per dpa irradiation dose. This could induce a change in the properties and cohesion of graphite.

钍基熔盐堆(Thorium Molten Salt Reactor,TMSR)是以核石墨为反射体及慢化体、2LiF-BeF2(FLiBe)熔盐为主冷却剂的反应堆.在TMSR中,核石墨直接与熔盐接触.由于石墨的多孔特性,熔盐有可能渗入石墨的孔隙中,引发其力学、热学性能的变化.研究熔盐在TMSR环境下是否渗入候选核石墨及其浸渗量,对于反应堆的运行安全至关重要.基于自行研制的熔盐浸渗实验装置,采用静态熔盐浸渗试验方法,测试TMSR候选核石墨T220在不同压强下的熔盐浸渗量,并研究了温度、时间对T220、NBG-18及IG-110石墨材料熔盐浸渗行为的影响.研究结果表明:T220石墨的临界浸渗压强介于600～700 kPa之间,这说明在TMSR工况下(<500 kPa)该石墨不发生FLiBe熔盐浸渗.温度(600°C和700°C)及时间(20～2000 h)对三个牌号石墨熔盐浸渗行为影响不大.

In thermal Molten Salt Reactors, the nuclear graphite core is in direct contact with the molten salt coolant. Due to the porous nature of nuclear graphite, the molten salt may infiltrate the nuclear graphite, which may affect the mechanical strength and irradiation behavior of the nuclear graphite. In order to evaluate the infiltration behavior of molten salt in nuclear graphite, both FLiNaK and FLiBe salts were used to infiltrate two typical nuclear graphite grades: IG110 and NBG18. The pressure dependence of the infiltration weight gain ratio was measured. The influence of molten salt infiltration on the thermal properties of these two graphite grades, such as their thermal expansion behavior and thermal conductivity, was also measured. The mechanical strength of the FLiNaK-infiltrated graphite was measured at room temperature and elevated temperature, and showed that the mechanical strength of the nuclear graphite was enhanced at room temperature and weakened at elevated temperature by molten salt infiltration. Finally, the thermal expansion coefficient and the fracture surface analysis measured after FLiNaK infiltration indicated that the stress induced by molten salt infiltration could be one of the reasons for the graphite property changes.

Presented in this article are mechanical property and micro-structural data for fluoride molten salt infiltrated graphite at high temperature. Four infiltration pressures (0 kPa, 450 kPa, 600 kPa, and 1000 kPa) and two kinds of graphite (IG-110 and NG-CT-10) were used during molten salt infiltration. After fluoride molten salt infiltration, compression testing and tension testing were performed at 700 degrees C to determine compressive strength, tensile strength, softening coefficient, stress-strain curve, and absorbed energy. Utilizing scanning electron microscopy (SEM) applied to fracture fragments, SEM micrographs for the fracture surface of molten salt infiltrated graphite and virgin graphite were determined. (C) 2018 The Authors. Published by Elsevier Inc. This is an open access article under the CC BY license (http://creativecommons.org/licenses/by/4.0/).