查看更多>>摘要：Considerable effort is being devoted to development and regulatory qualification of low enriched fuels for research and test reactors by many agencies worldwide. One promising fuel configuration being examined for United States higher power research and test reactors (USHPRRs) are plate-type fuels composed of a metallic uranium-molybdenum foil clad in an aluminum alloy. The two pieces of aluminum alloy cladding are bonded using a hot isostatic pressing method. The mechanical properties of the resulting bond line in the aluminum alloy cladding will vary by the HIP'ing parameters, requiring a need to characterize the bond line. Small scale mechanical testing can provide a path for evaluating the mechanical properties and deformation behavior of the bond line both prior to and following irradiation. In this research, room temperature micro-tensile specimens of non-irradiated and irradiated samples containing an Al alloy (AA 6061) bond line were tested to evaluate its strength and deformation behavior. Observations indicated that the strain rate did not affect the deformation behavior or strength and most of the micro tensile specimens failed in a ductile mode in grains around the bond line. There was no indication that the microstructural features from the bond line affected the mechanical properties of the micro-tensile specimens. An initial examination was performed on irradiated material but further systematic studies of the effects of irradiation can be performed in the future.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The concept of high-entropy alloys (HEAs) focusing on tuning the overall chemical complexity represents a novel alloy design strategy. In contrast, alloying of a metallic matrix with minor doping elements with limited and localized tunability has been a common practice to improve material performance. Combining the idea of globally engineering defect energy landscape in HEAs and the localized doping strategy in dilute alloys, in this work, we explore doping effects of minor elements in a HEA matrix to further enhance the overall and localized chemical tunability, aiming to improve its irradiation resistance. Specifically, we study the influence of minor Al, Cu, Ti, and Pd substitutional doping elements on defect energetics in a CoCrFeNi model HEA based on density-functional theory (DFT) calculations. The DFT results indicate that the formation and migration energies of vacancies can be strongly influenced when a dopant is introduced at the first nearest neighbor shells around a vacancy. On the other hand, interstitial energetics are only slightly affected. Among the four elements, Ti and Pd generally decrease vacancy formation energies and increase vacancy migration energies more significantly than Al and Cu. The doping effects become more pronounced when the concentration of the substitutional dopants increases. Based on the energy distributions obtained from DFT, we build a kinetic Monte Carlo (kMC) model to assess the impact of dopants on vacancy-mediated diffusivity in the doped HEAs. Our results suggest that Ti and Pd can lower the tracer diffusivity in the considered HEAs and act as trapping sites, whereas Cu may enhance the atomic transport. This work indicates that substitutional doping in HEAs is an effective strategy in metallurgy to further tune the defect and transport properties of complex alloys. (C)& nbsp;2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Results of known detailed microstructural studies of the Zircaloy-4 at various creep stages were used to validate the conventional microstructure-based MKE approach, modeling the kinetics of the production and recombination of dislocations. The developed model is considered a tool to interpret the governing mechanisms of deformation. The strain rate in non-stationary dislocation structures was determined by the mobility of dislocations in an external stress field, lowered by the back-stress imposed dislocation structure. The rate of microstructure recovery determines the steady-state regime's strain rate. As the root of power-law creep is still a disputable issue, microstructure recovery is analyzed in the view of conservative and non-conservative climb processes. It is shown that at least in materials where dipole instability is observed, the conservative climb may cause a change in activation energy and power-law exponent value close to the fifth power law.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：A B S T R A C T Liquid metals Li, Na and NaK are the candidate heat transfer mediums which are installed in the irradiation capsule of advanced fusion neutron source. The compatibility issue of F82H and 316L in the candidate heat transfer mediums was studied by means of the corrosion tests at the temperature up to 823 K. The depletion of Cr, C and Fe from the steel was caused in liquid Li. The Li diffusion into the steel matrix was analyzed by TOF-SIMS. The corrosion in liquid Na was caused by the Na diffusion into the steel matrix and the surface oxidation. The corrosion in NaK was similar with that in liquid Na.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：An outstanding challenge in the manufacturing and joining of oxide dispersion strengthened steels is retaining the nanofeatures in the alloy throughout the fabrication and welding process. MA956 was friction stir welded with two different sets of welding parameters, resulting in a medium and high heat input. After welding, 5 MeV Fe ++ ion irradiations were performed at doses ranging from 50 to 200 dpa in the temperature range of 400 to 500 degrees C. Post-irradiation characterization was performed with scanning transmission electron microscopy and energy-dispersive x-ray spectroscopy to investigate the Y-Al-O dispersoids, voids, and dislocations. After welding, the dispersoid microstructure coarsened, resulting in fewer and larger dispersoids regardless of heat input. After irradiation, the dispersoid behavior in the welded material was sensitive to temperature, exhibiting growth behavior attributed to Ostwald coarsening at 500 degrees C but a mixture of nucleation and more muted growth at 400 and 450 degrees C, attributed to competing mechanisms of radiation-enhanced diffusion and Ostwald coarsening. Void swelling correlated to heat input; being more prevalent in the welded conditions occurring at lower doses and in higher values relative to the base material. The low values of swelling despite microstructure coarsening caused by welding demonstrate the excellent swelling resistance of MA956, even after welding with the highest swelling values of 0.5% noted in the stir zone high heat input condition at 450 degrees C, 200 dpa. The dislocation behavior was inconsistent: the strongest trend was that network density was higher for welded versus base material, and an increase in loop diameter with temperature was observed. A rate theory model based on the observed microstructure suggests at high temperature interstitial loss to sinks was more likely to be dominant compared to mutual annihilation via point defect recombination, because of an increase of the radiation diffusion coefficient with temperature regardless of initial welded microstructure.Published by Elsevier B.V.
查看更多>>摘要：Zirconium alloys in nuclear power plants operate in high-pressure water at temperatures between 250 and 350 degrees C. Hydrogen (or deuterium) ingress due to waterside corrosion and if the solubility is exceeded H precipitates as a brittle hydride phase. Degradation mechanisms involve the accumulation of these brittle hydrides at cold spots or crack tips, as a result of H redistribution in response to thermal and stress gradients, respectively. Knowledge of H diffusion coefficients at operating temperatures is central to evaluating the rate of hydride accumulation and crack growth velocity. We determine the diffusion coefficients of H in Zircaloy-2 and Zr-2.5%Nb rolled plates at 250 degrees C, 300 degrees C and 350 degrees C along the rolling and normal directions by neutron imaging experiments with sensitivity of 5 wt ppm H for a spatial resolution 0.04 mm x 2 mm. These values were evaluated from H concentration profiles measured at room temperature on specimens of dimensions 10 x 10 x 4 mm(3) containing a hydride layer on one face, after annealing treatments between 60 and 600 min. This allowed the identification of a transition zone of 200-300 mu m between the hydride layer and the Zr alloy material, composed by large, sparsely distributed hydrides. In Zircaloy-2 plates, no substantial differences were observed in H diffusion along different directions or metallurgical conditions, and diffusion coefficients (0.6 +/- 0.1 10(-10) m(2)/s at 300 degrees C). By contrast, in hot rolled Zr-2.5%Nb plates the diffusion along the rolling direction (5.5 +/- 0.5 x 10(-10) m(2)/s at 300 degrees C) was much faster than along the normal direction (2.5 +/- 0.7 10(-10) m(2)/s at 300 degrees C), very likely due to H diffusing along the continuous network of beta filaments. After a thermal treatment of 3 h at 860 degrees C the plate microstructure changed generating radically changed H diffusion coefficients, resulting in H diffusion being much faster along the normal direction (4.0 +/- 0.5 10(-10) m(2)/s at 300 degrees C) than along the rolling direction (1.4 +/- 0.5 10(-10) m(2)/s at 300 degrees C). (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：During Loss of Coolant Accident (LOCA) in Pressurized Water Reactor (PWR), the oxidation of zirconium alloy cladding has a significant impact on the core degradation process. However, empirical parabolic rate correlations applied in accident analysis codes only provide oxide thickness obtained by the hypothesis of stoichiometric zirconia, which has limitations for simulating complicated oxide behaviors under accident conditions. The diffusion model with moving boundaries proposed in previous studies can give the distribution of oxygen concentration. However, due to simplified assumptions cladding expansion was not properly considered. In this study, an oxidation model for zirconium alloy cladding considering oxygen diffusion, zirconium conservation and mass density variation with temperature and oxygen concentration is developed. The model is validated by experimental data and shows good agreement. To analyze cladding oxidation during LOCA transient, QUENCH-05 and QUENCH-SR test are simulated. The model can give reasonable simulation results. The simulation results of QUENCH-05 test confirm the fact that beta-Zr plays an important role in maintaining the integrity of cladding under quenching process. The simulation results of QUENCH-SR test indicate that steam starvation can still lead to cladding embrittlement due to the consumption of beta-Zr matrix and thus increase the risk of cladding failure. (c) 2022 Elsevier B.V. All rights reserved.
Marshall, D. V.Eftink, B. P.Wang, Y. Q.Bourne, G. R....
11页
查看更多>>摘要：The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T(m)) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300 & nbsp;C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (delta HV expressionpproximexpressiontely equexpressionl to & nbsp;100) relative to the Low N specimen (delta HV expressionpproximexpressiontely equexpressionl to = 160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications. (C)& nbsp;2022 Elsevier B.V. All rights reserved.
Costa, Diogo RibeiroLiu, HuanLopes, Denise AdornoMiddleburgh, Simon C....
19页
查看更多>>摘要：UN-UO2 composite fuel is considered an advanced technology fuel (ATF) option to overcome the low oxidation resistance of the UN fuel. However, the interaction between UO2 and UN limits the performance of such composites. A possible way to avoid this interaction is to encapsulate the UN fuel with a material that has a high melting point, high thermal conductivity and reasonably low neutron cross-section. Amongst many candidates, refractory metals can be the first option. In this study, detailed investigations in UN-X-UO2 composite systems (X = V, Nb, Ta, Cr, Mo, W) were performed using SEM/FIB-EDS. The systems were heat-treated at 1773 K and 80 MPa for 10 min in vacuum using the spark plasma sintering method as a pressure-assisted diffusion apparatus. The results suggest that Mo and W are the most promising coating candidates to protect the UN fuel against interactions with UO2. Both metals are inert to N migration and preserve sharp interfaces with the nitride fuel. V, Nb, Ta and Cr strongly interact with UO2 and UN and form their respective nitrides V2N/V8N, Nb2N, and Cr2N. The formation of TaNx was not observed but Ta reacts with UO2 and forms two phases at the UO2-Ta interface (UTa2O7 and Ta2O5), while O from UO2 + x diffuses throughout the Ta foil and oxidise the UN pellet via grain boundary attack. This oxidation mechanism also occurs at the V, Nb and Cr-UN interfaces. Our recent atomic scale modelling of the X-UN interfaces also proposes Mo and W as the optimal candidates. Therefore, these results validate the coating candidates for the UN fuel and may guide further experimental/modelling development in UN-X-UO2 advanced technology fuel. (C)& nbsp;2022 The Author(s). Published by Elsevier B.V.& nbsp;
查看更多>>摘要：This paper presents simulations of the VERDON 1 and RT6 tests (temperature increase up to fuel-clad melting, oxidizing and/or reducing conditions within the furnace) performed with high burnup UO2 fuel (i.e., up to 72 GWd/tU) and considering a coupling between irradiated fuel thermochemistry and a fission gas release model. The thermochemical calculations rely on the Thermodynamics of Advanced Fuels -International Database (TAF-ID) for the description of the phases likely to form from the 15 fission product considered in the fuel (Ba, Ce, Cs, I, La, Mo, Nd, Np, Pd, Rh, Ru, Sr, Tc, Te, Zr) and on the Open-Calphad solver for the minimization of the Gibbs energy of the system. The gas release model describes a diffusion process for the gases within equivalent spherical grains. It accounts for the gases formed by reaction between the fission products as well as the chemically inert noble gases. The progressive fission product depletion of the fuel due to their release in gas form closes the coupling. The impact of the fission product release kinetics on the thermochemical equilibria within the fuel is then studied by coupling the thermochemical calculations with a gas diffusion model. The coupled simulations led to a very good agreement with the release kinetics of various fission products (I, Te, Cs, Mo, Ba). The released fractions of the low-/non-volatile fission products and of uranium, plutonium are also well reproduced. The observed differences in the fuel-clad melting temperatures in the two tests (2200 degrees C during RT6 and 2600 degrees C during VERDON 1) are however not reproduced by the simulations. (c) 2022 Elsevier B.V. All rights reserved.
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