18 Search Results

Asmore » an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. In this work, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr₂ phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)₆Fe phase (space group I4/mcm) and cubic (U,Zr)(Fe,Cr)₂ phase (space group Fd $\overline{3}$m). The (U,Zr)(Fe,Cr)₂phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm $\overline{3}$m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4₂/mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.« less

Here, to develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boost the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissionsmore » per initial heave metal atoms for initial screening purpose. After irradiation, the initial U-10Zr separated into an a-U annular region and an UZr_2+x center region with nanoscale spinodal decomposed microstructure. Because of the provided large amount of coherent interface areas, the fission gas atoms and vacancies generated in UZr_2+x phase are possibly pinned at the interface areas, leading to 20 times smaller fission gas bubbles than those in the neighboring a-U. The large bubbles in a-U become connected and merged into large pores that provide fast release paths for fission gas which prevents further fuel swelling. The fuel center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.« less

The microstructure of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750 was investigated. Both alloys were irradiated at low dose rates (~2 × 10^-8 dpa/s) to a neutron fluence of 6.9 × 10²² n/cm² (E > 0.1 MeV) at 371–389 °C. Different types of defects, including Frank loops, cavities, and γ' precipitates were characterized. The Frank loops in Type 304 stainless steel (SS) are larger in size (~50 nm in diameter) and lower in number density (2.58 × 10²¹ m^-3), compared to most previous higher dose rate neutron irradiation studies. The Frank loops in X-750 have an averagemore » size 26.0 nm of and a number density of 9.44 × 10²¹ m^-3. In 304 SS and X-750, cavities are of ~20 nm and ~14 nm in diameter, respectively. The swelling of both alloys was found to be insignificant. In 304 SS, Ni and Si were found enriched at the cavity surfaces and Ni,Si-rich precipitates were also found. Multivariate statistical analysis using non-negative matrix factorization reveals that these Ni,Si-rich precipitates contain only ~5.7 at.% Si, differing from the Ni₃Si γ' precipitates found in several previous studies. In X-750, L1₂-structured γ' precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni₃(Ti,Al)) of the precipitates and the superlattice reflections confirmed the stability of the crystal structure of these γ' precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.« less

Finely dispersed two phase microstructures resulting from a spinodal decomposition are of interest as they are associated with enhanced mechanical properties and excessive interfaces to mitigate defect related behavior. This study reports a spinodal-like phase decomposition in a U–50Zr alloy by thermal annealing at 620 °C and ion irradiation at 550 °C, with the latter temperature too low to initiate pure thermal phase transformation. Results of this study hold broad impact for U–Zr alloy systems and its application as advanced nuclear fuel.

The radiation response of a Fe–20Cr–25Ni austenitic stainless steel under self-ion irradiation at 500 °C was systematically investigated. The steel was irradiated at 500 °C by 3.5 MeV Fe2+ ions to 10, 50, and 150 peak dpa, respectively. In the 200–400 nm depth region, radiation-induced Frank loops were relatively stable in both size and number density from 10 to 150 peak dpa. Anisotropic distribution of Frank loops was observed in the 50 and 150 peak dpa specimens, possibly due to interaction of Frank loops and network dislocations with preferred orientations. Coarse voids were found only in the 50 and 150more » peak dpa specimens in depths less than 750 nm, suggesting that injected interstitials at deeper regions suppressed the void nucleation. The peak swelling was very low (~0.4%) for both 50 and 150 peak dpa irradiation. Radiation also led to the formation of intragranular plate-like Cr-rich carbides. Radiation-induced segregation of Ni and Si was found at various sinks: dislocation loops, void surfaces, and carbide-matrix interfaces. Finally, irradiation hardening was measured by nanoindentation and the results are consistent with microstructure-based calculations using the dispersed barrier hardening model. The major contributor to irradiation hardening changed from Frank loops at the lowest dose to network dislocations at the highest dose.« less

Proton irradiation induced Nb redistribution in Zr-xNb alloys (x = 0.4, 0.5, 1.0 wt%) has been investigated using scanning transmission electron microscopy/energy dispersive X-ray spectroscopy (STEM/EDS). Zr-xNb alloys are mainly composed of Zr matrix, native Zr–Nb–Fe phases, and β-Nb precipitates. After 2 MeV proton irradiation at 350 °C, a decrease of Nb content in native precipitates, as well as irradiation-induced precipitation of Nb-rich platelets (135 69 nm long and 27 12 nm wide) were found. Nb-rich platelets and Zr matrix form the Burgers orientation relationship, [$$1\bar{1}1$$]//[$$2\bar{1}\bar{1}0$$] and (011)//(0002). The platelets were found to be mostly coherent with the matrix withmore » a few dislocations near the ends of the precipitate. The coherent strain field has been measured in the matrix and platelets by the 4D-STEM technique. Here, the growth of Nb-rich platelets is mainly driven by coherency and dislocation-induced strain fields. Irradiation may both enhance the diffusion and induce segregation of interstitial Nb to the ends of the irradiation induced platelets, further facilitating their growth.« less

In this paper, the thermal-stability of 9Cr F/M ODS alloy was evaluated by the aging treatment in the air at 700 degrees C for 10 h, 100 h, 1000 h, and 2005 h. The variations of mechanical properties with different aging periods were tested in terms of tensile testing and vickers microhardness. And the microstructural evolution of strengthening particles and grain morphologies in the 9Cr F/M ODS alloys were characterized using field emission scanning electron microscopy, transmission electron microscopy and high-energy X-ray diffraction. It was found that the as-rolled 9Cr F/M ODS alloy slightly recovered without recrystallization during the initialmore » stage of less than 10 h aging time, resulting in the slight loss of tensile strength and microhardness and the increase of elongation. With the aging time prolonged to 2005 h, several of nano-sized Y-Ti-O particles coarsened and the composition changed to Y-Ti-Al-O, and the number density of large size nitride precipitations increased and the composition also changed from TiN to (Ti,AI)N. Meanwhile, some of grains tend to grow up in some preferential directions due to the zener pinning effect by the localized uneven distributed particles inside the grains, causing more particles, especially the particles with a size of larger than 40 nm, located at grain boundaries. These microstructural evolutions lead to the increasing of microhardness and a little bit decreasing of eleongation during 10 h-2005 h, and the increasing of tensile strength during 10 h-1000 h with the subsequent slight decreasing during 1000 h-2005 h. In conclusion, 9Cr F/M alloy shows a relatively good thermal-aging stability at 700 degrees C for 2005 h.« less

High entropy alloys (HEAs) have been considered for applications in nuclear reactors due to their promising mechanical properties, corrosion and radiation resistance. It has been suggested that sluggish diffusion kinetics and lattice distortion of HEAs can enhance the annihilation of irradiation-induced defects, giving rise to a higher degree of tolerance to irradiation damage. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages over conventional austenitic stainless steels (SS), we performed in-situ ion irradiation experiments with 1 MeV krypton at 300 °C on two HEAs and a 316H SS under an identical irradiation condition. Themore » irradiation introduced a high density of dislocation loops in all materials, and the microstructural evolution as a function of dose was similar for HEAs and 316H SS. Nanoindentation tests showed that the degree of irradiation hardening was also comparable between them. Furthermore, the similar microstructural evolution and irradiation hardening behavior between the HEAs and 316H indicate that, at low temperatures (≤300 °C), the irradiation damage of fcc alloys is not sensitive to compositional variation and configurational entropy.« less

In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 x 10(15) ions/cm(2) at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 x 1015 ions/cm(2), aboutmore » 51% of the loops were a(0)/2 < 111 > type for the 673 K irradiation, and the dominant loop type was a(0)< 100 > for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100-150 MPa. (C) 2017 Elsevier B.V. All rights reserved.« less

Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86more » x 10¹⁷ ions/cm², or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction (ARD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, Dⁿ - D$$n\atop{0}$$ = K phi, with n = 5.1. The Al-ZrN interaction products (Al₃Zr and AIN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance.« less