查看更多>>摘要：In the present study, Ti3AlC2 samples were irradiated at room temperature by Fe ions, He ions, sequential Fe and He ions. Our results demonstrate the evolution of irradiation defects with the damage level and sequential two sorts of ions irradiation. A large number of stacking faults and a small amount of twins are formed under the damage level of ～8 displacements per atom during Fe ions irradiation. The former contributes much to the formation of the latter. In the sequential irradiated samples, the following He ions irradiation promotes a further evolution of the defects induced by Fe ions irradiation resulting in significant decreases in the intensity of GIXRD; the fluence of 1 × 1016 He/cm2 gives rise to a high density of dislocation loops parallel to the basal plane and meanwhile a few of He bubbles are observed.
查看更多>>摘要：The addition of small concentrations of alloying elements to pure zirconium can prevent nuclear fuel cladding material from undergoing unstable oxide growth in aqueous environments at light water reactor operating temperatures. The role of alloying elements in stabilizing the oxide growth is examined in this paper, to better understand the oxide growth stabilization mechanism. To this end, a set of initial, short-duration corrosion experiments were performed, followed by oxide layer characterization. Specifically, ten model Zr alloys were selected to test the effect of small alloying additions on the alloy corrosion rate and corrosion breakaway. These alloys were corrosion tested in pure water in an autoclave at 360 °C for up to 70 days. The alloys included crystal bar Zr, sponge Zr, and model alloys with small concentrations of Sn, Fe, and Cr. After testing, the alloys were characterized using scanning electron microscopy (SEM) and synchrotron μ-X-ray fluorescence (μXRF) to study how the structure of the oxide and alloying element distribution related to unstable oxide growth. Initial results in the 360 °C water environment showed breakaway oxidation may be caused by unstable oxide growth due to heterogeneous distribution of the alloying elements. Heterogeneous distribution of alloying elements was correlated to the occurrence of unstable oxide growth (either nodule-like, grain boundary penetration, or differential grain-to-grain growth). It is possible that this heterogeneity, made possible by low alloying element content, can cause breakaway corrosion, but further study is warranted.
查看更多>>摘要：Forged Reactor Pressure Vessel (RPV) heads are manufactured from large ingots that are more than a hundred tonnes in weight, although the forging industry has implemented important measures to reduce carbon macrosegregation in these components, residual chemical inhomogeneities may still be present in the final product. Such remaining chemical inhomogeneities, along with other manufacturing variables, will translate into mechanical properties differences within the part. The present paper describes the development of predictive models that can be used to assess the effect of carbon macrosegregation on some of the most conventional mechanical properties of nuclear RPV heads. The particularity of these models is the use of a Machine Learning method to create them, in this case the neural networks technique built within a Bayesian framework was used to develop tensile property models (UTS, YS and Elongation). A first attempt was also performed to create models for more complex properties such as the Charpy impact toughness and the fracture toughness. The results show that the tensile property models, and to a less extent the Charpy impact and fracture toughness models seem to have captured reasonably well the variable interactions describing the problem involved, this was observed in spite of the reduced number of input variables used to create the models. This is because neural network models can cope with complex input variable interactions more efficiently than conventional linear relationships. The models can therefore be used to explore the effect of specific individual manufacturing variables, as well as variable interactions on the mechanical properties of nuclear RPV heads of steel 16MND5 (equivalent to SA-508 cl 3 steel).
查看更多>>摘要：A new interpretation of the breakaway swelling observed in high-density dispersion fuels (U6Fe, U3Si, etc.) under irradiation in research reactors is proposed on the base of Mansur's pore coalescence mechanism for randomly distributed immobile pores owing to their growth and pair impingement. This mechanism was further developed in the previous papers of the authors by considering triple and multiple collisions and was modified in the current paper in application to equilibrium gas-filled pores (bubbles). New analytical solutions in the mean-field approximation for the case of pair and triple collisions of equilibrium pores at the early stage of irradiation (with the swelling less than 50–60%) are presented. For higher swellings it is necessary to use a more accurate statistical approach, based on kinetic Monte Carlo calculations and considering multiple collisions of growing pores. It is shown that at high burnups, a sharp increase in the growth rate of the fuel swelling begins at a relatively high value of the fuel swelling (about 60%), in a qualitative agreement with observations for various intermetallic compounds.
查看更多>>摘要：During operation of a light water reactor, waterside corrosion of the Zircaloy nuclear fuel cladding causes hydrogen pickup. The absorbed hydrogen can redistribute in the cladding driven by existing concentration, stress, and temperature gradients. When the concentration reaches the solubility limit, hydrides precipitate. These hydrides can be more brittle than the Zircaloy matrix, so they can endanger the cladding integrity during a transient if their concentration is too high. In recent years, extensive efforts have been made to understand hydrogen behavior and to develop simulation tools able to predict hydrogen diffusion and hydride precipitation and dissolution. These efforts led to the development of the Hydride Nucleation-Growth-Dissolution (HNGD) model and its implementation into the nuclear fuel performance code Bison. While it offers a significant improvement and accurately predicts the amount of precipitates, this model fails to predict the thickness of the hydride rim under a temperature gradient. The current work presents the limitation of the HNGD model and proposes two hypotheses to improve the model's accuracy. The first hypothesis introduces a time dependency to the supersolubility to reduce the nucleation barrier as hydrogen atoms find more favorable nucleation sites. The second one introduces a hydride content dependency to the solubility. These hypotheses were validated and implemented into Bison and are now available to the user community. The modified HNGD model accurately predicts the hydride rim thickness, and it was demonstrated that this updated model can be used in Bison to model Zircaloy cladding with a zirconium inner liner. Finally, potential experimental and numerical methods are discussed to further validate these hypotheses.
查看更多>>摘要：Tensile mechanical properties of two AlxCoCrFeNi (x = 0.3 and 0.4) high-entropy alloys (HEAs) have been tested in liquid lead-bismuth eutectic (LBE) at 350 and 500 °C. The results show that LBE has a mild effect on the mechanical properties of Al0.3CoCrFeNi HEA which is a pure FCC structure, but some traces of liquid metal embrittlement (LME), such as cleavage, are detected at 350 °C and a tendency of intergranular (IG) cracking at 500 °C is also discovered. The mechanical properties of Al0.4CoCrFeNi HEA are strongly temperature-dependent. At 350 °C, the degradation effect of LBE is weak, while after the testing temperature increases to 500 °C, the mechanical properties are severely deteriorated even without the help of LBE. The mechanism responsible for this severe degradation phenomenon can be attributed to a network of thin (Ni, Al)-rich precipitates existing at the grain boundaries of the Al0.4CoCrFeNi HEA. As temperature increases, the precipitate/matrix interphase boundaries (IBs) become weak and can be perfectly wetted by LBE, leading to prevalent IG cracking facets on the fracture surface.
查看更多>>摘要：Post-irradiation annealing (PIA) is beneficial in recovering degraded mechanical properties of irradiated materials due to the evolution of irradiation-induced defects. To understand the underlying responsible mechanisms, isothermal PIA experiments were carried out in this work on a neutron-irradiated 9%Cr reduced-activation ferritic-martensitic steel (EUROFER97) using in-situ transmission electron microscopy. In general, PIA at 600 °C led to an appreciable reduction of the dislocation loop density and increment in the loops mean size. This is found mainly associated with the vacancy-mediated loops shrinkage phenomenon. The driving force for this process is loop's curvature force. Nevertheless, the spontaneous evolution of the loops neighboring microstructure substantially alters their annealing kinetics. Therefore, in this study, loops’ annealing kinetics in EUROFER97 is discussed in terms of various scenarios originating in terms of the presence or absence of the local external sources and sinks. In addition, influence of loops type, alloying elements segregation and free surface are also discussed.
查看更多>>摘要：Neutron-irradiation induced prismatic dislocation loops contribute to the degradation of materials’ mechanical properties. Recent post-irradiation annealing (PIA) investigations on neutron-irradiated reduced-activation ferritic-martensitic (RAFM) steel “EUROFER97” have revealed an almost complete recovery of the mechanical properties. To identify underlying recovery mechanisms, thick-foil annealing experiments on neutron-irradiated EUROFER97 were carried out. It is observed that PIA at 550 °C results in a reduction of initial dislocation loops density with a simultaneous increase in their mean size. This is linked to the dislocation loops shrinkage/annihilation as well as their coarsening/merging phenomena. Moreover, a clear coarsening of the quasi-circular 〈100〉 loops into their rectilinear versions and simultaneous shrinkage/annihilation of the ?<111> loops/clusters are observed. The coarsening phenomenon is a non-conservative climb process, which is discussed in terms of the movement of jogs whilst considering the change in the iron's elastic anisotropy with temperature and loops lowest energy directions. Furthermore, the observed interaction of the coarsened 〈100〉 loops and thereafter their merging to form large loops of various geometries are presented and discussed. Lastly, a rationalization justifying pronounced microstructural transition from dislocation loops dominated to dislocation tangles dominated microstructure with continued annealing is presented. These thick-foil PIA results complement our previous works and provide complete description of the dislocation loops annealing behaviour in neutron-irradiated EUROFER97.
查看更多>>摘要：The development and availability of a new 3rd generation SiC fiber, the Tyranno SA4 (SA4), are promising for the processing of higher neutron and/or corrosion resistant SiC/SiC composites. Despite its promising properties, especially the higher crystallinity and thermal conductivity than the Hi-Nicalon S fiber, the previous Tyranno SA3 (SA3) reinforcement leads to low damage tolerant SiC/SiC composites, restraining its use as a reinforcement. This is the consequence of very high interfacial shear stress, whatever the pyrocarbon interphase thickness. In this work, tubular samples where produced with both reinforcements and with two different pyrocarbon interphase thicknesses for tensile mechanical characterizations to access the potential benefit of the new SA4 fibers. The tensile mechanical properties of SA4 composites are highly enhanced compared to SA3 composites. The low damage tolerance drawback of SA3 composites is solved with higher failure strain for SA4-based composites. Tensile mechanical tests also highlight an unusual influence of pyrocarbon interphase thickness on the composites tensile modulus and proportional limit stress. The thinner interphase (≈ 60 nm) is the most interesting for repeatable mechanical properties and induces high proportional limit stress. Unloading – reloading cycles during tensile mechanical tests also highlight the benefit of this new fiber compared to Hi-Nicalon S. This work demonstrates that the substitution of SA3 by the new SA4 SiC fiber reinforcement in the processing of SiC/SiC composites is a great opportunity for the ceramic matrix composites development and especially for nuclear applications.
查看更多>>摘要：The influence of recrystallization on the recovery of radiation damage in tungsten (W) was investigated. Polycrystalline cold-rolled W foils were annealed in the temperature range between 1100 and 1600 °C and recrystallization and grain growth was observed. Subsequently, the specimens were irradiated with 7 MeV protons at cryogenic temperature. The induced radiation damage and its recovery during isochronal annealing treatments was monitored by in-situ measurements of the electrical resistivity. We observe that the recrystallized samples exhibit (a) a significant reduction in total defect recovery and (b) changes in characteristic features of the recovery spectra with respect to the as-received ones. These results are discussed and correlated to the effect of grain boundaries and dislocations on the radiation defect reactions.
NSTL
Elsevier