The concept of accident-tolerant fuels has been proposed and widely investigated over the past decade. The contribution is focused on one of the near-term approaches -the modification of the surface of existing Zr-alloy claddings by protective coatings. The studied specimens are based on Zr-1%Nb substrate with chromium coating, multicomponent chromium nitride and chromium coating and multi-layer chromium nitride/chromium coating deposited by a physical vapor deposition process. Thermomechanical tests were designed to understand the cladding deformation and the burst conditions during the LOCA phenomena. Presented results show both the positive and the negative effect of coating on cladding behavior. All coatings exhibited a reduction of ballooning size and prolongation of time to burst. On the contrary, coating can be connected with larger opening size after burst, higher hydrogen content and deterioration of the local mechanical properties.
Zirconium fuel claddings act as a first barrier against release of fission products during nuclear power plant operation and interim storage of the spent fuel. During the reactor operation, cladding tubes are exposed to different stress level at elevated temperatures and neutron irradiation in corrosive environment. It causes a material degradation by corrosion, cladding embrittlement by hydrides and radiationinduced damage or radiation growth and creep of the fuel rods. The irradiation damage effects mainly contribute to the loss of material ductility. In our study, microstructure of as-received (non-irradiated) Zr-alloys used in LWR (Zr1Nb, Zr-1Nb-1.2Sn-0.1Fe, Zr-1.5Sn-0.2Fe-0.1Cr) were examined by electron microscopy methods. Transmission electron microscope (TEM) was used to describe the microstructure of claddings used in different reactor conditions and identify the radiation-induced damage, which is presented on Zr1Nb irradiated to one standard campaign in the VVER-1000 active zone. Following Electron Backscatter Diffraction (EBSD) method on transparent foils complements the TEM results in larger area, i. e. by grain size and orientation or analysis of local misorientation after irradiation. Radiationinduced damage was observed in Zr1Nb metallic matrix as type dislocation loops, presence of radiation-induced precipitates or partial amorphization of the secondary phase particles. EBSD method showed no changes in crystallographic orientation, but a local increase of dislocation density can be affected by neutron irradiation.
Zirconium-based alloys are commonly used as a material for nuclear fuel claddings in the light water reactors. The cladding material must function to fix a huge number of pellets, while conducting heat into the coolant that flows turbulently around the fuel rods. Cladding tubes can contain gaseous fission products that escape the fuel. Thus, by functioning as a sealed unit, it prevents a contamination of the coolant water with high-radioactive fission products. The integrity of claddings is always a critical issue during reactor operation and wet or dry storage and transport of the spent fuel rods. Moreover, the role gains importance at Loss of Coolant Accidents (LOCA). After Fukushima accident, cladding materials are widely studied with the purpose to reduce the high-temperature oxidation rate and enhance accident tolerance. In our contribution, we introduce the studies on Zr-1Nb (E110) cladding tubes after high-temperature steam oxidation at 1350 °C. During the testing of claddings, microscopy analytical methods play an important role in experimental verification of pseudo-binary phase diagram Zr1Nb-O, i. e. particularly in oxygen content determination at phase transitions. Wave Dispersive Spectroscopy (WDS) with complementary nano-indentation method were used to characterize the Zr1Nb microstructure formed after LOCA. It includes the regions from an oxide and oxygen-stabilized α-Zr(O) to the acicular prior β-Zr phase. The decrease of hardness and Young's modulus corresponds with oxygen content measured in line-profiles by WDS. The oxygen level at transition points was partly determined from Fe, Nb β-stabilizers and significant change in mechanical properties in fine-grained prior β-Zr. The slight fluctuation of oxygen values in adjacent grains can be caused by preferential oxidation through the favorably oriented α-Zr(O) grains studied by WDS+EBSD. As well, the non-uniform oxygen-rich α-Zr(O) phase adjacent to the oxide was characterized by EBSD & WDS. Increasing hydrogen content in specimens, 10, 700 and 1000 ppm H, caused increasing solubility of oxygen in prior β-Zr phase upon high-temperature and the cladding material hardening.
Nickel-based alloys are considered promising materials for primary circuits of high-temperature gas reactors (HTGRs), specifically for gas turbines. The primary helium (He) coolant in the gas-turbine-based HTGRs is expected to reach temperatures of up to 900 °C; therefore, the selected materials should adequately perform over a long service life at such an environment. A promising manufacturing method in the production of reactor components is precision casting, where the content of revert (recyclate) material in the alloy differs and can influence the material behavior. In our study, Inconel alloy 738 was manufactured by casting 50% and 100% of revert material and tested in HTGR conditions to examine the influence of helium coolant on the material’s properties. Tensile specimens were exposed at 900 °C for 1000 h in helium containing a specified amount of gaseous impurities. Scanning electron microscopy (SEM) with energy dispersive spectroscopy (EDS), in combination with X-ray diffraction (XRD) and nano-, microhardness methods, were used for material characterization after performing the tensile tests at room temperature. The presence of three types of layers was observed: a thin layer formed by aluminum and chromium oxides on the surface; non-uniform surface oxides Ti3O5 with inner (Al,Cr)2O3; and the inner fine-grained Inconel Cr-enriched phase (approx. 10–20 µm below the surface), which can act as a protective surface layer. Mechanical properties of both revert materials decreased after exposure to HTGR conditions but did not show a significant difference as a result of the content of the revert material. The increase of nano-hardness in line profiles throughout the specimen’s cross-section was observed locally at the surface oxides and in the precipitates and grain boundaries. After exposure, Rp0.2 values decreased by 20% and 17.7%, and Rm values by 12.3% and 20.8% in samples with 50 and 100% revert content, respectively. Furthermore, a decrease in microhardness values (HV0.1) was detected by 4.98% in longitude and 5.80% in cross-section for samples with 50% revert material and by 3.85% in longitude and 7.86% in cross-section for samples with 100% revert material. It can be concluded that both revert materials have similar corrosion resistance in HTGR conditions. The presented results complement the knowledge about the degradation of alloys in the coolant environment of advanced gas-cooled reactors.
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