首页期刊导航|Journal of Nuclear Materials
期刊信息/Journal information
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
收录年代

    Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions

    Schulthess J.Kamerman D.Winston A.Pomo A....
    5页
    查看更多>>摘要：© 2022New fuels are being developed for light water nuclear reactors with the initial goal of improving accident tolerance. More recently, these new fuel systems are also being recognized for their potential to support higher fuel burnups and thus improved economics for reactor plants through improved fuel utilization. Fuel safety testing of these new fuels systems is being conducted under transient irradiation conditions in the Transient Reactor Test facility at Idaho National Laboratory. A test campaign including two rodlets of U3Si2 fuel in Zircaloy-4 cladding and two rodlets of U3Si2 fuel in Silicon-Carbide Composite cladding has been performed to begin the development of fuel-safety criteria in design-basis reactivity-initiated accident conditions. Post-Transient-Irradiation Examinations have been performed and found that the accident tolerant materials maintained their primary rod-like geometry. Elemental species diffusion was found to occur in the Zircaloy rodlets indicating rapid diffusion kinetics at the temperatures achieved. No elemental species diffusion was found in the Silicon-Carbide rodlets. However, through-wall thickness cracks were observed due to displacement loading of the cladding from thermal expansion of the fuel pellets. This work presents the first experimental evidence of the performance of these fuel systems under transient nuclear heating and representative reactivity-initiated accident energy depositions.
      原文链接:
    • NSTL

    Coupled modeling of irradiated fuel thermochemistry and gas diffusion during severe accidents

    Germain A.Sercombe J.Riglet-Martial C.Introini C....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.In this paper, a novel approach where irradiated fuel thermochemistry and gas release are coupled is presented in details and illustrated by the simulations of some tests of the VERCORS program characterized by increasing temperatures and varying gas composition in the furnace (oxidizing or reducing conditions). At each step of the tests, the oxidation/reduction of the nuclear fuel and the fission product chemical speciation are precisely assessed thanks to a thermochemical equilibrium calculation relying on the OpenCalphad thermochemical solver and on a built-in thermochemical database derived from the SGTE database and completed by a solid solution model for the U-O-fission product system. Fission product releases are estimated from the chemically reactive gases that form in the fuel (according to the thermochemical calculation) and from a gas diffusion model based on the equivalent sphere model. The gas diffusion model takes into account not only the noble gases available in the fuel prior to the test but also the chemically reactive gases that form during the test. It is shown that the proposed coupled approach provides a consistent estimation of fission product release (I, Te, Cs, Mo, Ba) during the VERCORS tests in spite of the simple gas diffusion mechanism considered in the simulations (no distinction between the fission products). The proposed coupled approach is used to test some thermochemical hypotheses to improve the calculated release of some fission products (Ba, Mo).
      原文链接:
    • NSTL

    High temperature hardness testing of δ-zirconium hydride: Yield stress estimation by analytical and numerical models

    Cherubin I.J.S.Long F.Daymond M.R.
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.The mechanical properties of δ-zirconium hydrides were studied using a combination of nanoindentation techniques and numerical and analytical models. Two different alloys, containing different forms of δ-zirconium hydride (blister and rim), were analyzed at room temperature and at elevated temperatures up to 300 ∘C. A reduction of hardness from RT to 300 ∘C showed an approximately linear relationship with the change in temperature. For the δ-zirconium hydride rim, the hardness reduced from 3.73GPa at RT to 1.49GPa at 300 ∘C, a reduction of 61%. For the δ-zirconium hydride blister, hardness decreased from 3.5GPa at RT to 1.91GPa at 300 ∘C, a reduction of 45%. The yield stresses were calculated using two approaches: a numerical model proposed by Johnson to calculate values at RT, and another proposed by Dao for values up to 300 ∘C. At RT, the yield stress for the δ-zirconium hydride rim obtained was 883±21MPa, 224MPa at 300 ∘C. The δ-zirconium hydride blister showed a reduction from 879.5±10.5MPa at RT to 324MPa at 300 ∘C, a reduction in 61% and 45% for the hydride rim and blister, respectively. Through this combination of experimental and modeling techniques, this study showed the variation of hardness and yield stress in complex systems, and in different environments, using nanoindentation as the main characterization technique.
      原文链接:
    • NSTL

    Oxidation properties and microstructure of a chromium coating on zircaloy-4 fuel cladding material applied by atmospheric plasma spraying

    Du P.Zhang R.Li Q.Song P....
    16页
    查看更多>>摘要：? 2021To achieve a full protective coat on welding joints at both ends of zircaloy cladding tubes, we studied the rapid deposition of a thick chromium coating on zircaloy-4 cladding tubes by atmospheric plasma spraying. Microstructure, mechanical properties and high-temperature oxidation tests of chromium coatings were also studied. The initial screening oxidation test of the specimen was carried out at 1100 °C, 1200 °C and 1300 °C in dry air, and the oxidation properties of the specimen were explained. The nanohardness and modulus of the sprayed chromium coating were higher than those of the zircaloy-4 substrate. The nitride phase of Cr2N was formed after the high-temperature oxidation test. After oxidation at 1300 °C, there were multilayer structures: surface Cr2O3, residual Cr coating, thin (Zr,Cr)O2 layer, Cr-Zr layer and zircaloy substrate. There were bubble folds on the surface of the chromium coating and voids in the Cr2O3 layer. The oxidation test showed that the chromium coating can effectively slow the severe oxidation of zircaloy-4 cladding tubes. In summary, an APS-thick chromium coating has the potential to improve the oxidation resistance of zircaloy-4 cladding tubes in light water reactors.
      原文链接:
    • NSTL

    Cold sprayed Cr-coating on Optimized ZIRLO™ claddings: the Cr/Zr interface and its microstructural and chemical evolution after autoclave corrosion testing

    Fazi A.Stiller K.Andren H.-O.Thuvander M....
    5页
    查看更多>>摘要：© 2022Cr-coated Optimized ZIRLO™ cladding material fabricated with the cold-spray deposition process is studied. Microstructure and chemistry of this material are investigated before and after exposure to autoclave corrosion testing with scanning electron microscopy, energy dispersive spectroscopy analysis, electron backscattered diffraction, transmission electron microscopy and atom probe tomography. The results are used to assess what changes have occurred upon autoclave exposure. The formation of a compact, 80 – 100 nm thick Cr2O3 layer is observed on the surface of the exposed samples. Nucleation of ZrCr2 intermetallic phase is discovered at the Cr/Zr interface. This Laves phase nucleates inside the intermixed bonding layer that can be found in both pristine and exposed samples, and decorates the interface in the form of small particles (less than 50 nm in size). Using transmission electron microscopy and atom probe tomography the growth of a Zr-Cr-Fe phase was detected. This phase is found in the region of the Zr-substrate immediately adjacent to the coating, up to a few hundred nanometres distance from the Cr/Zr interface. A small degree of recrystallization occurs upon autoclave exposure in the 1-2 µm thick nanocrystalline layer produced on the Zr-substrate by the cold spray deposition method utilized for the fabrication of the Cr-coating.
      原文链接:
    • NSTL

    Ion-irradiation-induced clustering in Fe-Mn-Ni-(Si) steels: Nucleation, growth and chemistry evolution

    Sha G.Xue F.Xue J.Jin S....
    5页
    查看更多>>摘要：© 2021This research addresses kinetics and thermodynamics of clustering in Fe-Mn-Ni-(Si) steels under high-energy Fe ions irradiation at 623 K, a dose rate of (4.6 ± 0.3) × 10-4 dpa/s and doses up to 1.5 dpa, to gain deep insights into nucleation, growth and chemistry modification of solute clusters using atom probe tomography. The investigation, for the first time, reveals that the clustering in the steels involves two kinetic processes i.e. Mn-Ni-(Si) cluster chemistry modification and size growth. The chemistry modification by which the clusters reach a stable chemistry is accomplished prior to cluster fast growth. The slow chemistry modification makes the Mn-Ni-(Si) clusters exhibit rapid growth subsequently (via coalescing adjacent clusters). Si addition slows down the cluster chemistry modification, with 0.6 dpa required for the Mn-Ni clusters to approach a stable composition, less than 1.5 dpa for the Mn-Ni-Si clusters. The Si addition enhances nucleation of Mn-Ni-Si clusters, prolongs cluster composition modification, and delays the commencement of their fast growth.
      原文链接:
    • NSTL

    Influence of grain size on the high-temperature creep behaviour of M5Framatome1 zirconium alloy under vacuum

    Trego G.Brachet J.C.Vandenberghe V.Portier L....
    5页
    查看更多>>摘要：© 2021The effect of grain size on the viscoplastic behaviour of M5Framatome zirconium alloy thin sheets was investigated at high temperature under uniaxial tension, using a variety of equiaxed microstructures with controlled grain sizes. In the α phase domain, a Coble diffusional creep regime and a dislocation creep regime were observed, in agreement with the literature. A negative sensitivity of the strain rate to temperature was highlighted in the upper part of the α+β two-phase temperature range, consistently with the literature. For the first time, a linear creep regime was evidenced in the β phase domain. In this regime, a sensitivity of the strain rate on the third power of the grain size is observed, suggesting a Coble regime with diffusion along grain boundaries. Modelling with multi-mechanism Norton power-law rate equations, including dependence on grain size and a homogenous strain-rate assumption (Taylor model), enabled to satisfactorily reproduce the experimental results over the 700–1100 °C temperature range, especially the negative sensitivity of the strain rate to temperature between 880 and 930 °C. Very good agreement was obtained with a second order self-consistent and full field homogenization schemes.
      原文链接:
    • NSTL

    Distinct He-induced damage evolution in nickel-based alloys irradiated at elevated temperatures

    Zhu Z.Ji W.Huang H.
    5页
    查看更多>>摘要：© 2022 Elsevier B.V.The nickel-based alloy Inconel 617 and Alloy 800H have been envisaged as the leading candidate for the higher-temperature Molten Salt Reactors (MSRs) due to their superior high-temperature properties. To evaluate their potential applications, both alloys were irradiated by He ion at 750 and 850 °C with corresponding He ion fluence. Same irradiation tests were performed on Hastelloy N alloy to study its microstructure evolution at elevated irradiation temperature, as well as being the reference alloy to reveal candidate alloys irradiation tolerance. Transmission electron microscopy (TEM) reveals the depth distribution of irradiation induced He bubbles and Frank loops. The TEM graphs and quantitative analysis shows that when irradiated at 850 °C, the formation of precipitated clusters lead to denser and boarder distribution of He bubbles in Inconel 617, while the bubble-loop complexes form and contribute to larger volume fraction of He bubbles in Hastelloy N alloy. The nanoindentation technique is used to measure their mechanical changes, and the results show that the Alloy 800H has the lowest degree of irradiation induced hardening when irradiated at 850 °C, whereas that of all alloys irradiated at 750 °C are basically similar. Further, the yield strength increment is calculated and deviations between experimental and calculated values are caused by the formation of precipitated clusters in Inconel 617 and the selection of smaller obstacle strength in Hastelloy N alloy. By comparison, the Alloy 800H demonstrates a greater potential to function as the structural material for higher-temperature MSRs due to smaller volume fraction and better resistance to irradiation induced hardening.
      原文链接:
    • NSTL

    Study of the hardness and Young's modulus at the fuel-cladding interface of a high-burnup PWR fuel rod by nanoindentation measurements

    Schneider C.Fayette L.Zacharie-Aubrun I.Blay T....
    5页
    查看更多>>摘要：© 2022 Elsevier B.V.During irradiation in a nuclear reactor, the fuel swelling and the cladding creep are at the origin of the contact between the fuel and the cladding, which in turn leads to the initiation of Zr oxidation by the UO2 fuel. In high burnup fuels, the zirconia layer formed at the Fuel Cladding Interface (FCI) presents numerous circumvolutions and a heterogeneous microstructure within its thickness. In this paper, nanoindentation measurements are carried out in the fuel-cladding interface of a PWR fuel rod irradiated up to a high burnup of 61 GWd/tU. The measured hardnesses and Young's moduli of the internal zirconia layer are compared to reference measurements obtained on a non-irradiated sample of commercial Y-TZP. The study of the evolution of mechanical properties across the zirconia layer led to the identification of four distinct zones related to the zirconia microstructure (phase, grain size, porosity and fission product recoil). Close to the cladding, the zirconia hardness HZrO2tends toward the zirconium hardnessHZr. Then, the zirconia hardness increases significantly with the grain size, until reaching a plateau at the middle of the zirconia layer, where the maximum hardness is obtained. Following the same trend, the elastic modulus progressively increases in the zirconia layer and reaches a plateau near the fuel High Burnup Structure.
      原文链接:
    • NSTL

    Characterization and qualification of neutron radiation effects – Summary of Japan-USA Joint Projects for 40 years –

    Muroga T.Hatano Y.Katoh Y.Clark D....
    5页
    查看更多>>摘要：© 2021 Elsevier B.V.The Joint Projects under the Japan-USA Fusion Cooperation Program started in 1981 and has continued for more than 40 years. In the Joint Projects, although a wide range of fusion materials and engineering issues were covered, neutron radiation effects on fusion reactor materials have always been the major research emphases, and the neutron irradiation facilities in the US were jointly used by Japanese and US researchers. Japanese test facilities including neutron and charged particle irradiation facilities were complementarily used. The initial focus of the Joint Projects was on fundamental fusion neutron radiation effects and irradiation correlation. Systematic comparison of fission and fusion radiation effects in comparable damage levels and the effects of transmutation-induced helium were investigated. The collaboration was then focused on the effect of dynamic irradiation effects in variable conditions. In addition to the relatively fundamental studies, the Joint Projects contributed largely to development of candidate materials such as RAFM steels, vanadium alloys, SiC/SiC composites, and tungsten alloys, through a mechanism-oriented approach. The Joint Projects also covered issues specific to materials application to fusion blankets and plasma-facing components, including neutron radiation effects such as tritium retention and permeation of neutron-irradiated plasma-facing materials. Various irradiation technologies were developed and applied to the irradiation experiments, including those for in-situ testing. Considering that high energy neutron sources, such as A-FNS and IFMIF-DONES, now have high viability, the research supporting the neutron source programs is essential. The knowledge obtained through the Joint Projects is valuable and should be advanced for this purpose. To this end, it is of urgent necessity to launch an international scientific program accumulating knowledge of fusion neutron radiation effects, including their fundamental aspects.
      原文链接:
    • NSTL