[en] Highlights: • Fe–Cr–Si alloys are considered as a potential ATF fuel cladding coating. • Fe–Cr–Si could offer similar benefits to FeCrAl without as much weld cracking. • Oxidation resistance improves with higher chromium wt. % in Fe–Cr–2Si alloys. • At/above 16 wt % Cr, Fe–Cr–2Si alloys resist corrosion similar to FeCrAl. • Further refinement of Cr and Si could resist oxidation and prevent intermetallics. - Abstract: Fe–Cr–Si alloys are potential candidates as accident tolerant fuel (ATF) cladding, which could offer similar benefits as Fe–Cr–Al alloys. The corrosion and oxidation behaviors of three Fe–Cr–2Si alloys with various chromium content (12 wt. %, 16 wt. %, 20 wt. %) in simulated primary water chemistry of a pressurized water reactor (PWR) were investigated. Post-test characterization included weight change, scanning electron microscopy (SEM), and X-ray Photoelectron Spectroscopy (XPS). The results showed that Fe12Cr2Si had the worst corrosion/oxidation resistance among the three alloys, while Fe16Cr2Si and Fe20Cr2Si showed excellent oxidation resistance due to their thin, continuous, dense oxide layers grown in simulated PWR conditions. These results bring insight into the corrosion behavior of Fe–Cr–Si alloys with varying chromium content exposed to the primary water environment during normal operating conditions found in PWRs.

[en] Highlights: • Thin film Cr₂AlC MAX phases were deposited by magnetron sputtering. • The films were irradiated in situ within a TEM with 320 keV Xe⁺ ions up to 90 dpa. • At 300 K irradiation temperature the film fully amorphises above 3.3 dpa. • At 623 K irradiation temperature no visible amorphisation up to 90 dpa was observed. High radiation hardness at 623 K is possibly based on presence of nano crystallinity. • The films showed ductile behaviour and good adhesion to Inconel 718. - Abstract: Cr₂AlC MAX phases were deposited using magnetron sputtering. The synthesis was performed via layer-by-layer deposition from elemental targets onto Si wafer and polished Inconel® 718 superalloy substrates at 650 K and 853 K. Transmission Electron Microscopy (TEM) characterisation showed that the thin films had a thickness of about 0.8 and 1.2 μm for Si and Inconel® substrates, respectively, and a MAX phase crystalline structure. Depositions onto Inconel substrate was performed in order to measure film mechanical properties. The films have hardness at around 15 GPa, reduced Young's modulus at around 260 GPa, do not delaminate and showed characteristic ductile behaviour during nanoscratching. Ion irradiations with in situ TEM were performed with 320 keV Xe⁺ ions up to fluence 1 × 10¹⁶ ions·cm⁻² at 300 K and 623 K. At 300 K the Cr₂AlC started to amorphise at around 0.3 dpa. At displacement levels above 3.3 dpa all crystalline structure was almost completely lost. Conversely, irradiations at 623 K showed no recordable amorphisation up to 90 dpa. It is discussed that the presence of many grain boundaries and low defect recombination energy barriers are responsible for high radiation hardness of Cr₂AlC MAX phase at 623 K. The thin film Cr₂AlC MAX phases have mechanical and radiation stability which makes them a candidate for fuel rod coating as Accident Tolerant Fuels (ATF) material for the next generation of nuclear reactors.

[en] Post-irradiation annealing (PIA) was conducted on a 304L stainless steel irradiated to 5.9 dpa in the Barsebäck-1 BWR reactor, to investigate its effect on the mitigation of irradiation-assisted stress corrosion cracking (IASCC) susceptibility. IASCC susceptibility was measured for the as-irradiated and four PIA conditions (500 °C: 1 h and 550 °C: 1, 5, and 20 h) via interrupted constant extension rate tensile and four-point bend experiments under simulated BWR-NWC conditions. The annealing treatments were observed to progressively reduce IASCC susceptibility, as measured by the final intergranular fracture fraction (tensile) and crack length per unit area (four-point bend), with full removal of IASCC susceptibility being observed following annealing at 550 °C: 1 h for tensile specimens and 500 °C: 1h for four-point bend specimens. Among the microstructure and mechanical property parameters measured as a function of PIA, the average dislocation channel spacing was observed to decrease by ∼25% and ∼40% from the as-irradiated condition after annealing at 500 °C: 1 h and 550 °C: 1 h, respectively. The mitigation of IASCC susceptibility correlated well with the decrease in the average dislocation channel spacing and is consistent with a process in which crack initiation is controlled in part by the high tensile stress at dislocation channel-grain boundary intersections.

[en] The effects of helium (He)-vacancy clusters on the stress-strain behavior of polycrystalline iron (α-Fe) are investigated by a mechanistic finite element (FE) approach using a continuum damage mechanics (CDM) description of the material behavior informed by molecular dynamics (MD) data. First, MD analyses of a single crystal (loading normal to {332} plane) and a bicrystal system containing a Σ11{332} grain boundary (GB) were performed to compute the uniaxial tensile response of an Fe single crystal and a system with a GB. MD results were then used in FE analyses of the same systems to identify parameters for the CDM constitutive relations for the crystal and the traction-separation law for the GB depicted by cohesive elements. Next, a 3D FE model of an α-Fe bicrystal system with an imperfect GB subjected to uniaxial tensile loading was developed. This model includes an equivalent hollow sphere under internal pressure in the middle of the GB to model the effects of pressurized He bubbles at 5 K, room temperature (RT) and 600 K on stress, strain and damage distributions. The radius of the equivalent sphere was determined assuming the presence of two vacancies in the system. Finally, MD stress/strain data of the same bicrystal system with He-vacancy clusters were compared to the corresponding FE results to validate this mechanistic approach that appears to be efficient in terms of computational time. FE model predictions of system strength and fracture strain are in fairly good agreement with the MD results at all three temperatures. Our results show that small and highly pressurized He-vacancy clusters reduce GB strength and fracture strain more significantly at 5 K than at RT and 600 K.

[en] Highlights: •LIBS was used for depth profile analysis of impurity deposition on surface of graphite tile removed from EAST lower divertor. •It is observed that a number of chemical species are presents in impurity layer on tile surface. •The Mo from first wall and W from upper divertor were sputtered by high energy particles and get deposited on the divertor. •The plot of signal intensity gives the entire information about elemental composition and depth profile of impurity layer. •The content of impurity is non uniform on tile surface and decreases as the laser pulse penetrates in the depth. - Abstract: Laser-induced breakdown spectroscopy (LIBS) has been applied for the depth-resolved identification of impurity deposited on EAST divertor tile. LIBS spectra show the presence of impurity elements (W, Mo, Li, Na, Ca, Cr) and tile material (C, Si, Ti) in the layer deposited on the graphite tile surface. The analysis of impurity deposition was performed in different positions on tile surface. The results indicate that the interaction of high heat plasma with plasma facing components (PFCs) leads to erosion of Mo from first wall and W from upper divertor. Depth profiling measurements show decay of impurity signals with successive laser shots and non-uniform impurity deposition in the different positions on tile surface. The impurity deposition was also measured at the different depths in the tile. The measurements show a non-uniform deposition at various depths in the deposition layer. Surface morphology and composition of tile were verified by SEM and EDX techniques. The study indicates that LIBS has potential to monitor the erosion and deposition of PFCs in fusion devices. The results obtained from LIBS technique are important for long pulse of H-mode plasma operation in EAST.

[en] Incorporating Raman spectroscopy with transverse lift-out specimens is demonstrated to effectively characterize depth-dependent ion-irradiation damage in nuclear ceramics, such as SiC/SiC composites irradiated up to 1, 10 and 50 displacements per atom (dpa) at 350 °C using 10 MeV Au ions. Raman spectroscopy reveals irradiation-induced structural disorder saturation in both SiC-fiber and SiC-matrix at doses as low as 1 dpa, despite vastly different microstructures, inferred from similar longitudinal optical (LO) and transverse optical (TO) phonon peak shifts. Diamond (D) and graphitic (G) peaks from SiC-fibers disappear under irradiation, revealing irradiation-induced carbon packet loss. The irradiation-induced carbon packet loss is also verified by conducting TEM on same FIB foils used for Raman spectroscopy. In a previous study, the irradiation-induced SiC-fiber shrinkage is known to occur due to carbon packet loss in fibers.

[en] SiC ceramic matrix composites are a potential replacement for current light water nuclear reactor fuel cladding material. However, loss of fission gas via micro-cracks and corrosion remain an issue. Cathodic arc Cr, CrN, and TiN coatings were deposited on SiC tubes and plates to provide hermeticity and corrosion resistance. These coatings were characterized to determine as-deposited quality. Cross-sectional microscopy, X-ray diffraction, glow discharge optical emission spectroscopy, and scratch tests were performed to evaluate the purity, structure, and mechanical performance of the coatings. Nitride coatings had stable interfaces, but larger defects in the coatings as compared to the Cr coatings which showed cracking at the interface, but less deposition-induced defects. Despite the local state of the interface, the mechanical properties of the metallic coatings versus ceramic coatings enabled the Cr coatings to resist loads three times that of the nitride coatings during scratch tests. Glow-discharge optical emission spectroscopy showed that improvement in elemental purity is needed for future coatings.

[en] This work examines the radial evolution of a high burnup fast reactor mixed-oxide (MOX) fuel pellet irradiated at the Fast Flux Test Facility (FFTF) using scanning (SEM) and transmission electron microscopy (TEM). At 21.3% fissions per initial metal atoms (FIMA), the sample analyzed in this work is likely among the highest burnup oxide fuel samples ever evaluated using TEM. Initial SEM examination focused on the characterization of the fuel microstructure and solid fission product behavior as a function of radial position. The fuel pellet underwent extensive restructuring as demonstrated by columnar grain formation near the central void. Oxide fission product phases were confined to near the fuel centerline and decomposed from perovskite BaZrO₃ to the fluorite ZrO₂ type crystal structure. Metallic five metal precipitates (FMPs) were observed at all radial positions and the Pd-rich metallic phase increased in area fraction moving toward the pellet surface. TEM examination focused on the determination of dislocation density ranging from the fuel pellet center to the pellet surface. Dislocation density was highest in the cooler regions of the fuel pellet, with the outer edge experiencing grain refinement and high defect concentrations consistent with the rim structure. The results of the paper highlight the variable dislocation behavior as a function of radial position in high burnup MOX fuels, as well as describe the largely unexplored change in crystal structure of the oxide solid fission product phase.

[en] Highlights: •Simulated molten core concrete interaction debris with actinide tracers were synthesized. •Subsequently, these debris were used for leaching tests. •The solid-solution phases (Zr_yU_1-y)O₂, (Ca_yU_1-y)O₂, and (Ca_zZr_yU_1-y-z)O₂ were formed. •The cement components formed a glass-like coating on the debris, which suppressed actinide leaching. - Abstract: To understand the chemical structure and stability of molten core-concrete interaction (MCCI) debris generated by the Fukushima Daiichi nuclear power plant accident in Japan in 2011, simulated MCCI debris consisting of the U–Zr–Ca–Si–O system and other simpler systems were synthesized and characterized. ²³⁷Np and ²⁴¹Am tracers were doped for the leaching tests of these elements and U from the simulated debris. The MCCI debris were synthesized by heat treatment at 1200 °C or 1600 °C, in reductive (Ar + 10% H₂) or oxidative (Ar + 2% O₂) atmospheres. Subsequently, the debris were used for actinide leaching tests with water. Zr and Ca formed a solid-solution with the UO₂ matrix, such as (Zr_yU_1-y)O_2+x, (Ca_yU_1-y)O_2+x, and (Ca_zZr_yU_1-y-z)O₂, which stabilized the matrix and suppressed actinide leaching from the simulated debris. On the other hand, the cement components (CaO and SiO₂) in the debris formed a glass-like coating on the debris, which also remarkably suppressed the leaching of actinides.

[en] Highlights: •FGR in the micro-cell pellets is estimated by a simplified approach based on FGR modeling for the conventional UO₂ pellet. •Up to 10 times higher diffusivity, FGR in the ceramic micro-cell pellet is comparable to or lower than in the UO₂ pellet. •Up to 10 times higher diffusivity, FGR in the metallic micro-cell pellet is clearly lower than in the UO₂ pellet. •FGR in the micro-cell pellets would be determined by the fraction of perfect walls and the magnitude of gas diffusivity. - Abstract: Following the Fukushima accident in March 2011, Korea Atomic Energy Research Institute (KAERI) has developed micro-cell fuel pellets for the purpose of reducing gas release, both stable and radioactive, not only under accident conditions, but during normal operation as well. The micro-cell pellet is composed of many cells covered with a ceramic or metallic wall material, whose function is either to lower or block the movement of gas atoms out of the cells, thereby reducing gas release out of the pellet. Based on both a Halden irradiation test of the micro-cell pellets up to 16 MWd/kgU and experience of fission gas release (FGR) in the conventional UO₂ pellet, a simplified approach is proposed to estimate FGR in the micro-cell pellets when it is irradiated with a constant linear power of 30 kW/m up to 60 MWd/kgU under normal operating conditions. Considering the lack of measured data on diffusivity in the micro-cell pellets, sensitivity studies for FGR were performed for two cases: diffusivity in the micro-cell pellet is the same as and 10 times higher than in the UO₂ pellet, respectively. Even for the 10 times higher diffusivity, FGR in the ceramic micro-cell pellet is at least comparable to or lower than in the UO₂ pellet, and FGR in the metallic micro-cell pellet is clearly lower than in the UO₂ pellet, due to its lower fuel temperature and the impact of its walls on gas movement. In real situations, the degree of reduction of FGR in the micro-cell pellets would be determined by the fraction of perfect walls and the magnitude of gas diffusivity. Until PIE results become available for the irradiated micro-cell pellets, the current approach provides only rough estimates and therefore its results need to be taken as preliminary.