Journal of Nuclear Materials最新文献

Transient testing of fast reactor type fuel pins with DIN 1.4970 cladding has been performed in the TRIGA-ACPR Reactor at INR, Pitesti. The fuel test segments fabricated at SCK CEN resemble the fuel pins of the MYRRHA core, which is a lead-bismuth cooled fast reactor under development. The goal of the tests was to determine the cladding deformation as a function of the energy deposited in the UO₂ fuel, and to identify a potential cladding failure threshold (expressed in energy deposition) as a result of Pellet Cladding Mechanical Interaction (PCMI).

Based on the reactor physics assessment and the test fuel segment design, a dedicated new irradiation rig was constructed. Twenty UO₂ fueled test segments were manufactured at SCK CEN using recently fabricated DIN 1.4970 cladding tubes with two different degrees of cold working. The un-irradiated test segments have been designed such that they resemble, as much as possible, fuel pins at a high burnup level. This has been achieved by reducing the pellet-cladding gap size and a higher level of the cladding cold-working degree. A total of 8 irradiation tests were performed in the TRIGA-ACPR on groups of 2 or 3 pins, that were subjected simultaneously to a short power pulse. The deposited power in the fuel and the temperature of the cladding was recorded during the test and cladding profilometry was measured before and after.

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In this paper, bulk W-1 wt.% Y₂O₃–1 wt.% Ti (WYT) alloys with nanocrystalline (NC), ultrafine-grained (UFG), and fine-grained (FG) structures are fabricated using high-energy ball milling, spark plasma sintering and controlled annealing treatments. A systematic examination focuses on thermal load induced damages, particularly crack formation and mode, after repetitive thermal bombardments. The 1500 °C annealed WYT with UFG structure shows optimal thermal shock resistance, with only microcracks (∼0.29 μm width) after thermal bombardments at an absorbed power density (APD) of 0.33 GW/m². In contrast, both the NC-WYT and 1700 °C annealed WYT with FG structure demonstrate significant cracking with intergranular fracture at an APD of 0.22 GW/m², exacerbating with increasing the APD. By examining intrinsic microstructures, thermal conductivity, and mechanical properties, the potential mechanisms underlying the distinct thermal shock resistance of these alloys have been discussed. This work provides valuable insights into the thermal shock resistance of oxide-dispersion-strengthened W alloys with different microstructures.

Intensification of corrosion in short-term experiments on carbon (СS) and stainless (SS) steel as materials for underground radioactive waste disposal by thermophilic microbial communities isolated from neutral (Tuva region) and alkaline thermal radionuclide-enriched environments (Tau Tona mine) in the presence of uranium under anaerobic conditions was studied. The corrosion rate of carbon and stainless steel in the presence of the Tau Tona culture increased up to 14 and 3.7 times, respectively. The increase correlated with the predominance of metal-reducing bacteria in the culture. The culture from Tuva, dominated by fermentative bacteria, increased the corrosion rate 6 to 4 times, depending on the type of steel and organic substrate. The main mechanism of microbial corrosion of both types of steel was the dissolution of passivating corrosion products, mainly magnetite, due to their reduction by iron-reducing bacteria or chelation by metabolic products of fermenting bacteria. Surface corrosion was observed on carbon steel and pitting on stainless steel. In the presence of uranyl ions, the corrosion rate of both types of steel increased up to 19 `. Uranyl ions can be reduced to UO₂ by microorganisms and will not further affect steel oxidation, furthermore, accumulation of reduced forms of uranium in corrosion products may passivate steel corrosion. It was found that in the presence of organic substances in the environment it can intensify chemical corrosion processes of both types of steel due to the complexation of steel corrosion products, preventing the formation of a passivating corrosion layer. Thus, in the presence of acetate, the corrosion rate of black and stainless steel was 12–36 % higher, in the presence of trehalose it increased the corrosion rate by 1–24 %.

Next-generation nuclear power plants are generally characterized by higher operating temperatures, increased neutron fluences and energies, and distinct corrosive coolant environments versus the existing light water reactor fleet. Whether using existing materials in new environments, newly developed materials tailored for these environments, or new manufacturing methods, the traditional decades-long approach for materials qualification does not facilitate rapid deployment. Ion irradiation has demonstrated success in reproducing material microstructure and select property evolution resulting from neutron irradiation with three to four orders of magnitude reduction in time and cost, making it an ideal candidate for accelerated irradiation testing. Because microstructure has a large impact on bulk material properties, limited neutron irradiation data at lower damage levels can in principle be combined with accelerated ion testing results and modeling and simulation to form an accurate prediction of microstructure evolution and select properties under different neutron irradiation conditions and at higher damage levels. The objective of this work is to present a conceptual framework of specific steps to fulfill several technical challenges associated with qualifying materials for performance in radiation environments on an accelerated time frame. A brief review of the regulatory landscape for materials in nuclear environments is presented, followed by additional overviews to understand the current state of the art for correlation of materials properties across radiation environments using experimental and computational methodologies. Finally, the roles of academia, national laboratories, and industry in the advancement of this accelerated materials qualification framework are discussed as a path forward, with possible case studies presented.

The epsilon particles that result from nuclear fission of UO₂ fuel possess both advantages and disadvantages from a spent nuclear fuel (SNF) management perspective. In this study, the effect of epsilon particles, namely Ru, Mo, and Pd, inherent in simulated UO₂ pellets is examined. Various analytical methods have been used to explore the changes in the structural, surface, and electrochemical properties of which contain epsilon particles. A notable finding is that the epsilon particles are not evenly distributed and tend to clump together, transforming into a metallic state after sintering, as detailed in the X-ray diffraction analyses. Energy-dispersive X-ray spectroscopy analyses highlight interesting aspects of the distribution of elements, especially the disappearance of Pd during sintering, which is likely due to its high vapor pressure. Although the lattice structure of UO₂ remains unchanged, the sizes of the grains and pores visibly change, which may influence the tendency of UO₂ pellet-cracking. Despite the addition of the epsilon particles, the electrical conductivity analyses show no significant changes, suggesting that they act as minor impurities without affecting the structural lattice. However, their possible role as catalysts in electrochemical reactions opens new and interesting areas that require thorough investigation. Moreover, examining the anodic dissolution under various conditions provides detailed insights into UO₂ dissolution and oxidation, revealing how epsilon particles subtly influence the oxidative dissolution process. This study clarifies the basic interactions and effects of epsilon particles in UO₂ pellets and broadens the path for a deeper understanding and improvement of nuclear fuel matrices and steering advancements in the safe and effective use of nuclear energy.

Microstructure refinement is an effective way to improve the mechanical properties and radiation resistance of FeCrAl alloy, which may also increase the susceptibility to high temperature steam oxidation (HTSO) due to the increased surface chemical activity. To investigate the microstructural effects on the HTSO behavior, a FeCrAl alloy was subjected to controlled annealing strategies to produce three types of microstructures: “fine sub-grains + fine precipitates”, “coarse grains + fine precipitates” and “coarse grains + coarse precipitates”. The three alloys were exposed to high temperature steam during a heating and holding procedure. Large oxide nodules and rough multilayer-structured oxide films were formed during heating, which subsequently evolved into a single α-Al₂O₃ layer during the isothermal oxidation. Microstructure refinement can promote the oxidation and slightly increase the weight gain. However, the variation of isothermal oxidation parabolic rate constants brought by the microstructure refinement is 1-2 orders of magnitude lower than that brought by alloying. Microstructure refinement can improve other service performance of FeCrAl alloy without significantly deteriorating the HTSO resistance, indicating its advantages for engineering applications.

The impact of thermal oxidation on Helium (He) implanted pure iron (Fe) was investigated to experimentally evaluate how thermal oxidation influences the diffusion and distribution of He within the material. In case of the sample with the lowest dose (2 × 10¹⁷ ions/cm²), the thinnest oxide layer was observed compared to the non-implanted pure Fe. It is due to He nano bubbles implanted in the material, which hinder the diffusion of Fe ions necessary for forming the oxide layer. For the medium dose sample (5 × 10¹⁷ ions/cm²), the oxide layer was slightly thicker than that of the lowest dose, however it had more porous structure due to the numerous nano He bubbles. In addition, a large bubble region around 150–200 nm depth was generated. The highest dose sample (1 × 10¹⁸ ions/cm²) was observed that the escape of He was accelerated by the large amount of nano He bubbles from within the metal, forming a porous Fe matrix on the top surface and an oxide layer that is very porous but thicker than those of lower doses.

In 2011, the Great East Japan Earthquake and tsunami caused a hydrogen explosion at the Fukushima Daiichi nuclear power plant, which exposed radioactive materials to the atmosphere and had a very negative impact on the nuclear power industry. Since then, efforts have intensified around the world to make nuclear power safer. Accident-tolerant fuel (ATF) is being developed to prevent the rapid oxidation of zirconium cladding, which directly causes hydrogen explosions. ATF research can be divided into two main approaches: changing the cladding material, or coating the surface of the cladding. Coatings are easier to commercialize and apply to existing nuclear power plants, so most vendors have focused on this approach. Chromium is a popular coating medium because of its superior properties such as a low oxidation rate and excellent adhesion. Therefore, research has been focused on suppressing the rapid oxidation of zirconium cladding in the event of an accident by adding a chromium coating with an appropriate thickness in terms of both economy and effectiveness. Even if the thickness of the coating is fixed, the penetration of oxidizing substances can be further delayed by improving the microstructure of the chromium coating, such as by reducing the grain boundary area. In this study, chromium-coated zirconium cladding tubes were fabricated by the arc ion plating process. The microstructure of the chromium coating was adjusted by varying the negative voltage (0–125 V), which in turn controlled the incident energy at which the chromium ionic particles hit the surface of the cladding tube. Experiments were then performed in 1200 °C steam environment to determine the optimal microstructure for high-temperature oxidation resistance. The development of various material degradation phenomena that occurred during 1200 °C steam oxidation was observed to identify the oxidation mechanism and the main factors of the microstructure that affect the zirconium oxidation rate.

Retention of hydrogen isotopes (HI) in plasma-facing components (PFCs) is a crucial process influencing the operation of a fusion device. The dynamics of HI retention is mainly determined by irradiation conditions of PFCs. These conditions can change due to the onset of fast transient events, such as edge-localised modes (ELMs). The development of ELMs results in repetitive short-term plasma bursts on PFCs, affecting the exposure regime. In this work, the effect of ELM-like loads on the fuel retention is numerically analysed using a one-dimensional diffusion model, implemented in the FESTIM code. As a representative simulation case, the deuterium (D) retention in tungsten (W) is considered with the geometry and inter-ELM exposure conditions relevant for large fusion devices. Temporal dependencies of heat and particle fluxes during intra-ELM stages are accounted for by applying the free-streaming model for an ELM filament transport. The simulations were conducted for three inter-ELM exposure regimes with various properties of transient loads, D trapping sites, and D recombination rates. Compared to the ELM-free irradiation, the onset of transients is shown to mainly reduce the D retention rate because of significant material heating induced by the arrival of energetic ELMy particles. This relative difference can decrease if the material is characterised by strong trapping sites or limited desorption from a surface. The findings also demonstrate that additional heating during transients provides conditions for a faster D migration into the bulk. In addition, we discuss the possibility of using ELM-average loads to obtain quick assessments of the D content. The approach is shown to be applicable for the case of small ELMs and is used to derive the analytical expressions for the D distribution in W within the steady-state approximation.

This study investigates the suppression of ZrCr₂ formation at the Cr/Zr interface by introducing trace amounts of Mg, Zn, and Sn into Cr coatings. Combining the first-principles calculation and experimental analyses, the inhibitory effects of these dopants on ZrCr₂ are examined. First-principles calculations predicted that Zn, Mg, and Sn can elevate the formation energy of ZrCr₂, with Mg exhibiting the most significant effect, thereby exerting an inhibitory influence on ZrCr₂ formation. Experimental findings demonstrate that Sn notably inhibits ZrCr₂ formation, resulting in a reduction of ZrCr₂ approximately 10%. However, Zn and Mg do not exhibit a substantial inhibitory effect on ZrCr₂ due to their low yield resulting from the low vaporization temperature. These results from computational simulations, alongside experimental validations, underscore promising strategies for mitigating ZrCr₂ formation, offering valuable insights for enhancing performance in nuclear fuel cladding applications.