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Journal of Nuclear Materials
North-Holland
Journal of Nuclear Materials

North-Holland

0022-3115

Journal of Nuclear Materials/Journal Journal of Nuclear MaterialsSCIISTPEIAHCI
正式出版
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    Predictions of ingress of hydrogen isotopes for zirconium alloys should not be based on solubility limits.

    McRae, G. A.Coleman, C. E.Nordin, H. M.
    5页

    Synthesis of U 3 O 8 and UO 2 microspheres using microfluidics *

    Nelson, Andrew T.Kurleya, J. MatthewHunt, Rodney D.McMurray, Jake W....
    6页
    查看更多>>摘要:Uranium-bearing microspheres below 50 mu m with a narrow size distribution allows for a wider variety of fuel forms. To accommodate the smaller size, gel microspheres with a composition of UO 3 center dot nH 2 O center dot mNH 3 were synthesized using microfluidics and subsequently converted to U 3 O 8 and UO 2 . To accommodate the slower flow rates required by microfluidics, a more stable broth was established. The gelation studies resulted in a broth that was stable for more than two days at 0 degrees C and for close to 3 h at room temperature while still gelling within 25 s. Synthesis of gel microspheres with a narrow size distribution lasted for 5 h and produced -0.5 g of air-dried material. The gelled microspheres were converted to U 3 O 8 and UO 2 and with sizes of 50 and 40 mu m in diameter, respectively.

    Influence of elasticity on the morphology of fcc-Cu precipitates in Fe-Cu alloys. A phase-field study

    Nizinkovskyi, RostyslavHalle, ThorstenKrueger, Manja
    12页
    查看更多>>摘要:Cu-rich precipitation ensures high strength and satisfactory ductility of steels at low carbon content. Al-though in an over-aged state, it leads to the degradation of mechanical properties. It is of prime impor-tance for nuclear reactor vessel steels, which are exposed to extreme conditions at elevated temperature for a long time. The precipitates at this state have an fcc structure and possess a characteristic elongated morphology. The invariant-line model was mostly used to explain the morphology of fcc-Cu precipitates in the bcc-Fe matrix, although it has several limitations. It idealizes morphology and does not consider the elastic properties of both precipitate and matrix. Also, this method does not reveal the influence of the eigenstrains on kinetics and the formation of the nanostructure. Besides, the influence of the applied stress on the precipitation process can't be revealed with this method. With this in mind, we developed a phase-field model to investigate the formation of the elongated mor-phology of fcc-Cu precipitates in the bcc-Fe matrix. The effect of diffuse interface and discretization on the behavior of solution was verified with a sharp-interface model and invariant-line prediction. It was found that simulated precipitates agree qualitatively with available experimental data. The role of a min-imal mismatch line on the morphology formation was verified. Also, it was shown that the anisotropy of elastic properties has a minor effect on morphology formation. The applied stress with different mag-nitudes and directions has a minor effect on the morphology of precipitates. Nevertheless, it results in variant selection due to the difference in interaction energy. The bifurcation diagram of interaction of two precipitates was constructed. The equilibrium configurations reveal that elasticity enhances the ki-netics of the reaction and that there is a twin-like configuration, which enhances the accommodation of shear-like eigenstrain and reduces the elastic interaction energy.

    Effect of rhenium addition on deuterium retention in neutron-irradiated tungsten

    Toyama, T.Matsumoto, A.Shimada, M.Oya, Y....
    8页
    查看更多>>摘要:The effects of rhenium (Re) addition on deuterium (D) retention in neutron-irradiated tungsten (W) were investigated. Pure W and W-5Re (5 at.%) alloy samples were irradiated with neutrons at High Flux Iso-tope Reactor using MFE-RB-19 J capsule. The sample temperature and the damage level were 864 K and 0.35 dpa for pure W and 792 K and 0.26 dpa for W-5Re alloy. A portion of the samples was ex-posed to D plasma at Tritium Plasma Experiment at Idaho National Laboratory at 823 K to a fluence of 5 x 10 25 m(-2). Vacancy-type defects in neutron-irradiated samples were examined using positron an-nihilation spectroscopy (PAS); D retention after plasma exposure was evaluated by thermal desorption spectrometry (TDS). TDS measurements revealed that D retention in the neutron-irradiated W-5Re alloy was similar to that in the unirradiated W sample, whereas a significant increase in D retention was observed in neutron-irradiated W. Thus, Re addition significantly suppressed the increase in D retention after neutron irradi-ation. This effect was attributed to the suppression of vacancy-type defect formation, as confirmed by PAS. (C) 2022 Elsevier B.V. All rights reserved.

    Synergistic effects of Si and Y on corrosion behavior of cast cladding steels by pre-laying Y powder for nuclear applications in static liquid LBE

    Ren, HaoZhang, XiaoxinChen, YingxueZeng, Xian...
    15页
    查看更多>>摘要:Ferritic/martensitic (F/M) steels with (9-12)wt.% Cr are potential candidates for structural applications in Lead-cooled Fast Reactors (LFR) due to their low solubility in liquid Lead-Bismuth Eutectic (LBE) coolant. How to improve corrosion resistance of 9-12Cr F/M steels has been a challenge. Here we cast three steels by pre-laying yttrium (Y) powder on the bottom of mold to obtained Y-bearing steels. Precursory work on this family of Y-bearing steels has previously highlighted them with the longer steady state creep stage and the lower minimum creep rate at 650 & DEG;C under 120 MPa, which has motivated their study under more extreme conditions, such as corrosion behavior in liquid LBE. We find a synergistic effect of Si and Y on both optimizing structure and reducing thickness of oxide scale at various oxygen concentrations, exposing temperatures and test times. The oxide scale changed from dual layers to single spinel layer, the thickness reduced from 36 mu m to 4 mu m, dissolution corrosion and PbBi penetration were inhibited. Low diffusion coefficient of ions in SiO2 and Y segregation on grain boundaries of steels hinder the outward diffusion of cations, and strong affinity between Y and O leads to the formation of Y2O3 preferentially, acting as nucleation sites of SiO2 and Cr2O3, accelerating a dense and continuous oxide scale. This work enables the development of structure steels with high corrosion resistance for various high temperature application, including boiler steel, gas turbine, and heat exchanger tubes.(C) 2022 Elsevier B.V. All rights reserved.

    Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

    Cappia, FabiolaTeng, FeiMurray, Daniel J.Yao, Tiankai...
    13页
    查看更多>>摘要:Fuel cladding chemical interaction (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the high-Cr tempered martensitic HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature (PICT) reached to 650 C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic & USigma;-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe2Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).(c) 2022 Elsevier B.V. All rights reserved.

    No ball milling nee de d: Alternative ODS steel manufacturing with gas atomization reaction synthesis (GARS) and friction-based processing

    Darsell, J. T.Wang, J.Zhang, D.Ma, X....
    11页
    查看更多>>摘要:Oxide dispersion strengthened (ODS) steels are promising structural materials for future fusion reactors. The high-density ( -10 23 /m 3 ) of highly stable Y-(Ti)-O nano-oxides provide high sink strength for radiation resistance and high-temperature ( > 650 degrees C) creep strength. Concomitantly, helium management is enabled by trapping high density ( -10 23 /m 3 ) of small ( < 3 nm) helium bubbles in the vicinity of nano-oxides. However, conventional route of making ODS steels involves prolonged ball milling, canning, degassing, and laborious thermo-mechanical processing (TMP). Such route, especially the batch-by-batch ball milling step, faces persistent challenge with scalability and high costs. Gas atomization reaction synthesis (GARS) method has demonstrated the potential of making precursor ODS steel powders without ball milling, but the nano-oxide density was around 10 21 /m 3 in the final consolidated form by conventional TMP. Taking advantage of GARS precursor powder, we use friction-based processing, including friction consolidation and extrusion, to manufacture ODS steel with further improved nano-oxide characteristics. Preliminary results showed that Y/Ti/O species were intimately mixed and rapidly reacted to form nano-oxides with a number density of -10 22 /m 3 .

    Characterization of microstructure and hardening of SLM nickel-based alloy irradiated by He ions

    Zhu, ZhenboHuang, HefeiMin, ShilingHou, Juan...
    12页
    查看更多>>摘要:For the application of additively manufactured nickel-based alloys in molten salt reactors (MSRs), it is crucial to evaluate their performance in nuclear environments, especially under He-induced damage. Therefore, in this study, the irradiation damage behaviour of selective laser melted (SLM) nickel-based GH3536 alloy was studied by irradiating it with He ions at various doses. Traditionally manufactured (TM) materials were used as reference materials to evaluate the resistance of SLM GH3536 to irradiation. Transmission electron microscopy (TEM) was performed to reveal the depth distribution of irradiation-induced He bubbles, as well as the characteristics of dislocation loops, using both on-zone scanning transmission electron microscopy (STEM) and rel-rod modes. Nanoindentation tests were performed on both alloys to investigate their irradiation hardening. The TEM graphs and quantitative analysis showed that, owing to the cellular sub-grain structures, the He bubbles in the SLM alloy exhibited a lower number density but larger size than those in the TM alloy. Additionally, the SLM alloy showed inferior He tolerance but better resistance to irradiation hardening, as revealed by the nanoindentation test results. Consistency between the experimental hardness increment and calculated yield stress increment showed that the irradiation hardening of the sample was mainly caused by the presence of irradiation-induced defects. A comparison between the on-zone STEM and rel-rod modes suggested that the rel-rod mode is more effective if only the effect of irradiation-induced defects on mechanical properties is studied. Furthermore, this study provides insight into the design of SLM nickel-based alloys for nuclear industry applications.(c) 2022 Elsevier B.V. All rights reserved.

    Influence of tunable interfaces on radiation tolerance and nanomechanical behavior of homogeneous multi-nanolayered Al 1.5 CoCrFeNi high entropy alloy films

    Pu, GuoLin, LiweiRen, DingGan, Kefu...
    16页
    查看更多>>摘要:Introducing high-density defect sinks, e.g., grain boundaries (GBs) and interfaces, to capture helium (He) atoms and irradiation-induced defects under He + irradiation is a promising strategy for constructing novel nuclear-materials with enhanced radiation tolerance. To understand the vital roles of GBs and interfaces on the irradiation response in high entropy alloys (HEAs), high-density GBs and homogeneous interfaces between layers were firstly introduced into multi-nanolayered Al 1.5 CoCrFeNi HEA films by tailoring the nominal monolayer thickness ( h ) ranging from 2 to 100 nm. The multi-nanolayered HEA films were then irradiated by 60 keV He + at a fluence of 1 x 10 17 cm -2 . The results show that the He bubbles with size of 1-2 nm were preferentially distributed along the GBs within the films and the phase structure kept stable in the irradiated HEA films with less interfaces (such as h = 100 nm). This suggests that the He bubbles mainly distributed at GB sinks due to the high-density GB intersections acted as dominant defect-sinks to trap He atoms. While both the distribution of He bubbles throughout the interfaces and the partial BCC -> FCC phase transformation due to radiation-induced segregation were disclosed by introducing numerous interfaces in the irradiated multi-nanolayered HEA films ( h = 10 and 30 nm). The corresponding intrinsic mechanism of He behavior was elucidated based on the interactions between the He defects and the nano-GBs or homogeneous interfaces. The underlying mechanisms of irradiation-induced phase transformation were explained by the relief of compression stress and the radiation-induced segregation behavior by vacancy-flux. Also, the roles of interfaces and GBs on the irradiation swelling and nano-mechanical properties of the HEA films were also revealed by the systematic investigations.

    Helium interaction with solutes and impurities in neutron-irradiated nanostructured ferritic alloys: A first principles study

    Ke, HuibinEdwards, Danny J.Setyawan, WahyuPitike, Krishna Chaitanya...
    15页
    查看更多>>摘要:Density functional theory (DFT) calculations are performed to explore the binding between He and alloy-ing solutes, impurities, and transmutation products expected in neutron irradiated nanostructured ferritic alloys (NFAs), here 14YWT is taken as an example. Elements that exhibit significant binding (attraction) with an interstitial He are: Y (binding energy = 0.46 eV), Mg (0.32), O (0.33), Ti (0.16), and C (0.15). Those that provide significant binding to a substitutional He are: O (1.44), Y (1.24), N (0.73), H (0.56), Mg (0.52), Ti (0.34), Si (0.34), C (0.33), Al (0.32), Ni (0.26), Ta (0.23), and Mn (0.16). The presence of these elements in Fe matrix could reduce the transport of He towards oxide particles, dislocations, and internal boundaries, and could promote He bubble nucleation in the matrix. For convenience, we compile existing binding en-ergy data of He with He-n and He-n V (He-vacancy) clusters. Dissociation pathway analysis reveals that, in general, the most likely dissociation of a He-n V cluster is by a sequential emission of individual He atoms. Furthermore, larger bubbles are more prone to dissociation than smaller ones. In addition, we estimate the binding energy (segregation energy) of He in bulk Y2Ti2O7 (YTO) single crystal, YTO/Fe interface, and YTO particle embedded in Fe, with respect to interstitial He in Fe, from existing formation energies of He in these structures. We also compile available data of He binding with Fe self-interstitial atom (SIA), SIA clusters, and edge and screw dislocations. Note that given the absence of DFT data, the binding with SIA clusters and dislocations are gathered from simulations with empirical potentials. The data presented in this paper is important to inform multiscale simulations of He bubble accumulation. (C) 2022 Elsevier B.V. All rights reserved.