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  • Three-dimensional bubble reconstruction in high burnup UO2
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-12
    Casey McKinney; Rachel Seibert; Grant Helmreich; Assel Aitkaliyeva; Kurt Terrani

    In light water reactor (LWR) UO2 fuels, the evolution of volatile fission products is one of the critical areas of fuel behavior that is yet to be fully understood. In UO2 irradiated to high burnups, it is well known that most released fission gases come from the central region of the fuel as opposed to the highly porous high burnup structure (HBS) on the periphery of the pellets. However, fuels with and without interconnected bubble networks at the fuel center showed high to moderate release fractions, which conceals the mechanisms responsible for the gas release in the latter scenario. In this work, focused ion beam tomography was used to investigate the three-dimensional bubble structure in an irradiated LWR UO2 fuel pellet with high degree of fission gas retention so that the degree of bubble interconnection could be assessed. Six radial locations with different burnups and temperatures were serially sectioned and imaged to reconstruct the three-dimensional bubble structure. As expected, the highest porosity was observed at the periphery of the fuel (HBS). The porosity then decreased towards the pellet center, except for the centermost location. This location had a slightly higher porosity than its adjacent mid-radial location, which was attributed to the temperature difference between the two locations. This study provides a first-time volumetric evaluation of the porosity at different radial locations on a UO2 fuel pellet. During this investigation, no significant bubble interconnection was noted at any of the six radial locations.

    更新日期:2020-02-12
  • GD-OES study of the influence of second phase particles on the deuterium depth distribution in dispersion-strengthened tungsten
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-10
    E. Lang; C.N. Taylor; J.P. Allain

    Dispersion-strengthened tungsten materials represent a new class of W-based materials to be investigated for use as plasma-facing component in nuclear fusion reactors. However, the retention and permeation characteristics of these materials under low energy deuterium (D) irradiation need to be elucidated before the efficacy of these materials can be judged. Due to possible deep penetration of D in W, depth profile techniques such as glow discharge optical emission spectroscopy (GD-OES) are needed to probe D concentrations many microns beneath the material surface. In this study, the D retention behavior of W materials containing 1–10 wt% TaC, TiC, or ZrC are investigated with GD-OES. After exposure to a 5 × 1018cm−2 D ion fluence at 100 °C, D was observed beyond the D implantation depth the surface in many specimens, and the D depth profile was found to depend on the type and concentration of the added second phase. Combined with in-situ X-ray Photoelectron Spectroscopy (XPS) studies, the effects of impurity oxygen atoms on the D retention is considered, as an increasing oxygen content correlates with decreased D retention. The influence of grain size, second phase particles, and oxygen content on the retained D depth and concentration in these complex W-based materials is discussed.

    更新日期:2020-02-10
  • A better nanochannel tungsten film in releasing helium atoms
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-08
    Guo Wei; Jingwen Li; Yonggang Li; Huiqiu Deng; Changzhong Jiang; Feng Ren

    Discovering highly radiation resistant plasma facing materials (PFM) is an urgent target for nuclear fusion reactor. In our previous work, a new PFM (nanochannel tungsten film) that consists of many tungsten crystal columns was found having excellent radiation tolerant properties. In the present work, three tungsten columns whose top surfaces are {100}, {110} and {111} -oriented were studied by molecular dynamics simulations and density functional theory calculations. We find that the tungsten columns whose top surfaces are {100}-oriented retain fewer helium atoms than the tungsten columns whose top surfaces are {110} or {111}-oriented. Moreover, the microstructural changes of tungsten columns whose top surfaces are {100}-oriented are smaller than others after high helium fluence exposure, which indicates that the nanochannel tungsten film maybe more suitable than bulk tungsten for plasma facing material in releasing helium atoms and delaying the formation of “fuzz” structure. The results reported here can help us to design a better radiation resistant plasma facing material in the future.

    更新日期:2020-02-10
  • Effects of temperature on helium bubble behaviour in Fe–9Cr alloy
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-08
    Zhanfeng Yan; Tengfei Yang; Yanru Lin; Youping Lu; Yue Su; Steven J. Zinkle; Yugang Wang

    In the present study, Fe–9Cr model alloys were irradiated by 275 keV He+ ions at 500 to 800 °C to a peak damage level and implanted helium content of 0.5 dpa and 0.75 at% He, respectively. Effects of temperature on He bubble behaviour were studied by transmission electron microscopy (TEM) to reveal the temperature effects. The bubble spatial distribution changes from homogenous at 500 and 600 °C to heterogeneous at 700 and 800 °C, due to preferential bubble precipitation on dislocations and grain boundaries at high temperature. Faceted bubbles with low energy faces along the {001} planes were present at all four temperatures and are attributed to surface energy minimization. Bubble precipitation on the GBs became more pronounced with increasing temperature. Meanwhile, the cavity swelling increased with increasing temperature, from 0.59% at 500 °C to 8.21% at 800 °C with a sharp increase above 700 °C. By analysing the temperature dependence of bubble size and density, a transition temperature between low and high temperature regimes was obtained in terms of different activation energies which are related to different bubble nucleation and formation mechanisms.

    更新日期:2020-02-10
  • 更新日期:2020-02-10
  • Sputtering of nanostructured tungsten and comparison to modelling with TRI3DYN
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-07
    R. Stadlmayr; P.S. Szabo; D. Mayer; C. Cupak; T. Dittmar; L. Bischoff; S. Möller; M. Rasiński; R.A. Wilhelm; W. Möller; F. Aumayr

    He-induced nanostructured tungsten (so-called W-fuzz) was bombarded with Ar ions under 60° and the dynamic erosion behaviour was experimentally investigated. By using a highly sensitive quartz-crystal-microbalance technique in a particle catcher configuration, the sputtered particles distribution of W-fuzz could be evaluated. In contrast to a flat sample, where sputtered particles are emitted primarily in forward direction, we find that W-fuzz samples emit sputtered particles preferably in backward direction (i.e. in the direction of the incident ion beam). After continuous Ar irradiation of a W-fuzz sample the distribution approaches that of a flat sample. In addition to experimental data we also show modelling results obtained with a state-of-the-art Monte-Carlo (MC) binary collision approximation (BCA) code TRI3DYN in full 3D. Surface morphology changes as monitored by SEM as well as the dynamic sputtering behaviour can be well reproduced by the full 3D MC-BCA code.

    更新日期:2020-02-07
  • SCIANTIX: A new open source multi-scale code for fission gas behaviour modelling designed for nuclear fuel performance codes
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-06
    D. Pizzocri; T. Barani; L. Luzzi

    Bridging lower length-scale calculations with the engineering-scale simulations of fuel performance codes requires the development of dedicated intermediate-scale codes. In this work, we present SCIANTIX, an open source 0D stand-alone computer code designed to be included/coupled as a module in existing fuel performance codes. The models currently available in SCIANTIX cover intra- and inter-granular inert gas behaviour in UO2, and high burnup structure formation as well. Showcases of validation in both constant and transient conditions are presented in this work. As for the numerical treatment of the model equations, SCIANTIX is developed with full numerical consistency and entirely verified with the method of manufactured solutions – verification of different numerical solvers is also showcased in this work.

    更新日期:2020-02-07
  • Epsilon metal: A waste form for noble metals from used nuclear fuel
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-06
    Denis Strachan; Jarrod V. Crum; Chase C. Bovaird; Charles Windisch; Mac Zumhoff; Brian McIntosh; Xiaolei Guo; Gerald S. Frankel

    Epsilon metal (ε-metal) is the metallic phase that forms as inclusions at the grain boundaries in the UO2 fuel during reactor operation. This metal is composed of Pd, Mo, Rh, Ru, and Tc. These metallic inclusions are insoluble in strong acid and remnants of these metallic inclusions have been found in the UO2 matrix that remains from the natural reactors in Gabon that were active 1.8 billion years ago, therefore ε-metal should be an excellent waste form for the immobilization of the long-lived isotopes 107Pd (6.5 × 106 a) and 99Tc (2.13 × 105 a), with 99Tc being the isotope of interest for repository performance. Therefore, the chemical durability of this potential waste form is assessed in this study. Typically, corrosion rates for metallic materials are measured electrochemically because they are quick, inexpensive, and can reveal the mechanism by which a metal corrodes, at least initially. However, in a repository the waste form would be subjected to slowly flowing water without an applied electrical potential over long time periods. Therefore, the corrosion rates of ε-metal specimens were measured with both electrochemical tests and the single-pass flow-through test (SPFT). Potentiodynamic and potentiostatic polarization results suggest that a thin passive film exists on the alloy surface, which seems to be responsible for its high corrosion resistance. Additionally, X-Ray photoelectron spectroscopic results suggest that Pd oxides are significantly enriched in the passive film Results from the SPFT show that the dissolution rates were weakly dependent on pH. Only Mo and Re were found in solution and were used for the calculation of the dissolution rates. In general, the electrochemically determined corrosion rates agree reasonably well with the initial dissolution rate measured with the SPFT test, but they are about one or two orders of magnitude higher than the steady state rates. The causes for this discrepancy are discussed.

    更新日期:2020-02-07
  • Hydrothermal reactivity of neutron absorber composites
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-05
    Kirsten Sauer; Marlena Rock; Florie Caporuscio; Ernest Hardin

    The stability of neutron absorber composite materials at hydrothermal conditions was tested in a series of two-week experiments to mimic spent nuclear fuel disposal. Coupons, composed of boron carbide (B4C) sintered with and encased in aluminum, were increasingly altered in experiments at 150, 230, and 300 °C and pressures of 150 bar. Alteration of aluminum to boehmite (γ-AlO(OH)) and hydrogen gas generation occurred over the range of investigated temperatures, but is most significant at 300 °C. The formation of boron-bearing mineral phases was not detected; however, aqueous boron was present in the reaction fluids.

    更新日期:2020-02-06
  • Effects of neutron flux on irradiation-induced hardening and defects in RPV steels studied by positron annihilation spectroscopy
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-04
    T. Toyama; T. Yamamoto; N. Ebisawa; K. Inoue; Y. Nagai; G.R. Odette

    Neutron-flux effects on irradiation-induced hardening and microstructures in a reactor pressure vessel steel were studied. An A533B-type steel containing no Cu was neutron-irradiated with fluxes of 1 × 1014 n/cm2/s (high-flux) or 1 × 1012 n/cm2/s (low-flux) to the same fluence of approximately 3 × 1019 n/cm2, and the same temperature of approximately 290 °C. The recovery behavior of irradiation-induced defects and irradiation-hardening, ΔHv, was investigated by post-irradiation isochronal annealing from 275 to 450 °C. In both the high- and low-flux cases, the recovery behavior of ΔHv and the average positron lifetime, τave, corresponded well to the annealing, suggesting that defects in which positrons are trapped are the origin of irradiation-hardening. The values of ΔHv and τave in the high-flux sample started to recover at around 350 °C, while those in the low-flux sample started to recover at around 400 °C. Thus, in the high-flux sample, unstable defects transiently existing at low temperature but annealed out at around 350 °C, are indicated. Such defects are suggested to be defect-(Mn, Ni, Si) complexes, where the nature of the defect is that of a mono-vacancy and/or dislocation loops.

    更新日期:2020-02-04
  • Uranium nitride tristructural-isotropic fuel particle
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-04
    Kurt A. Terrani; Brian C. Jolly; Jason M. Harp

    A summary of activities around concept development, design, manufacturing, and early irradiation testing of a novel coated fuel particle, uranium nitride tristructural-isotropic fuel is provided. This fuel particle offers significantly higher uranium density over historic manifestations of coated fuel particles and may be more optimal for a range of advanced reactor applications. After reviewing the design process, steps involving production of this fuel form are discussed. Low burnup irradiation data on this fuel particle are now available, indicating that performance metrics have been met to date based on the original design expectations.

    更新日期:2020-02-04
  • Microstructural characterization of copper coatings in development for application to used nuclear fuel containers
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-04
    Weiwei Li; Bosco Yu; Jason Tam; Jason D. Giallonardo; David Doyle; Dominique Poirier; Jean-Gabriel Legoux; Peter Lin; Gino Palumbo; Uwe Erb

    In the design of used nuclear fuel containers for deep geological repositories, copper is considered to be a suitable and long-lived barrier for corrosion resistance. The microstructures of state-of-the-art copper materials used in this application produced through extrusion, a grain boundary engineered electrodeposition technique and cold spraying were studied via electron backscattered diffraction. Desirable microstructural characteristics for localized corrosion resistance of pure copper were compiled from the literature considering grain size, grain boundary character distribution, and crystallographic texture. The subject copper materials were found to have favourable microstructures for localized corrosion resistance, in particular, a high fraction of special grain boundaries, especially Σ3 twins, rendering them suitable for the given application.

    更新日期:2020-02-04
  • Temperature dependent electrochemical equilibrium diagram of zirconium-water system studied with density functional theory and experimental thermodynamic data
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-04
    Guangxi Jin; Canhui Xu; Shuanglin Hu; Xiaosong Zhou

    Zirconium (Zr) alloys are important cladding materials of nuclear fuel in nuclear power industry, which are commonly immersed in complicated aqueous environments. Their corrosion relates critically to the safe operation of nuclear reactors, and is affected by many factors. As a foundation to study such complicated corrosion behaviors, in this work, the electrochemical stabilities of various species in the Zr-water system are directly demonstrated with the electrochemical equilibrium (Pourbaix) diagrams. The diagrams are calculated by applying methods of combining first-principles calculations on solid state phases with experimental thermodynamic data of aqueous species. The calculated Pourbaix diagrams are generally consistent with the results of previous experiments, while the solid phases in the oxide form (passivation region) of the calculated ones are much richer than the experimental counterparts. The stabilities of the solid phases are significantly affected by temperature and acidity (pH value) of the environment, as well as slightly by ionic concentration. To construct the experimental consistent Pourbaix diagrams for the Zr-water system, it is found necessary to correct the relative chemical potentials between solids and aqueous ions. Such a scheme of correction is demonstrated to be robust for the Pourbaix diagrams obtained with different exchange-correlation functionals. This work may help us to understand the corrosion process more clearly and guide us to enhance the corrosion resistance of the cladding Zr in cooling water circuit environment.

    更新日期:2020-02-04
  • Oxygen trapping in defect clusters in Fe and FeCr alloy by ion channeling and ab-initio study
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-02-03
    Vairavel Mathayan; Jaiganesh Gnanasekaran; Lakshmanan Chelladurai; Sundaravel Balakrishnan; Rajaraman Ramalingam; Binay Kumar Panigrahi; Amarendra Gangavarapu
    更新日期:2020-02-03
  • Effect of deformation level and orientation on SCC of 316L stainless steel in simulated light water environments
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-31
    Donghai Du; Miao Song; Kai Chen; Lefu Zhang; Peter L. Andresen

    The stress corrosion cracking (SCC) behavior of 316L stainless steel (SS) forged to 0%, 12.6%, 21.4%, and 39.8% reduction in thickness was investigated at constant K in light water reactor environments. The yield strength, specimen orientation, and water chemistry were correlated with crack growth rate, and their dependencies are discussed. The crack growth rate (CGR) of 316L SS increased monotonously with yield strength irrespective of the specimen orientation or water chemistry. Higher CGRs were observed when cracks propagate along the plane parallel to forging plane than normal to forging plane. The effect of local deformation on the anisotropic cracking behavior for different orientations, crack paths and CGRs are also discussed.

    更新日期:2020-01-31
  • Thermodynamic modelling of Y–H and Y–Zr–H system aided by first-principles and its application in bulk hydride moderator fabrication
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-30
    Jiaqing Peng; Ming Wu; Fu Du; Fengli Yang; Jianyun Shen; Lijun Wang; Xinyu Ye; Guoqing Yan
    更新日期:2020-01-31
  • One-pot synthesis of Ln2Sn2O7 pyrochlore and MgAl2O4 spinel by soft chemistry route as potential inert matrix fuel system, and the microstructural analysis
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-30
    Linggen Kong; Tao Wei; Yingjie Zhang; Inna Karatchevtseva; Ilkay Chironi

    MgAl2O4–Ln2Sn2O7 (Ln = Y, La, Nd, Gd and Tb) composite materials with 50–70 vol% of MgAl2O4 as potential inert matrix fuels have been synthesized using a soft chemistry route. X-ray diffraction (XRD), Raman, scanning and transmission electron microscopies (SEM/TEM) are employed to investigate the bulk and micro structures of both ceramic phases. Heat treatment at 1400 °C in air for 24 h leads to the formation of Ln2Sn2O7 pyrochlore and MgAl2O4 spinel which are confirmed by XRD, Raman and selected area electron diffraction (SAED). More importantly, the component in one phase does not affect the crystalline formation of the other. Thermal analysis results show MgAl2O4 crystallizes prior to that of Y2Sn2O7 and the crystallization of Y2Sn2O7 delays due to the presence of MgAl2O4. SEM shows uniformly distributed pyrochlore particles in spinel matrix, especially for the sample in equal volume ratio, with grain sizes being 1–5 μm for both phases. HRTEM and SAED observations confirm that both phases possess a high level of crystallographic perfection at the atomic scale.

    更新日期:2020-01-31
  • Crystal and metal/oxide interface structures of nanoparticles in 15Cr–2W–0.1Ti–4Al–0.6Hf–0.35Y2O3 ODS steel
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-28
    Peng Dou; Qiao Ye; Akihiko Kimura

    FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials of generation IV nuclear fission reactors because of its excellent resistance to not only corrosion but also creep and irradiation due to the ultrahigh density nanometer-scale oxides. Crystal and metal/oxide interface structures of the nanoparticles in FeCrAl-ODS steel with Hf addition, i.e., Fe–15Cr–2W–0.1Ti–4Al–0.6Hf–0.35Y2O3, have been studied by high resolution transmission electron microscopy (HRTEM). The phases of 153 nanoparticles, which have diameter mainly of 2–10 nm and peak number fraction and, therefore, represent the oxides contributing most significantly to the macroscopic properties of the ODS steel, were identified and, moreover, the proportions of various types of oxides were determined by statistical analyses. Relative to FeCrAl-ODS steel without Hf addition, i.e., Fe–15.5Cr–2W–0.1Ti–4Al–0.35Y2O3 ODS steel, the coherency of the oxide nanoparticles are considerably improved. About 51% of the nanoparticles were found to be consistent with Y2Hf2O7 oxide having anion-deficient fluorite structure whereas only about 32% of the nanoparticles are composed of Y–Al complex oxides, indicating the addition of 0.6 wt% Hf into FeCrAl-ODS steel inhibits the formation of Y–Al complex oxides remarkably while prompts the significant occurrence of Y–Hf complex oxides. Y–Ti complex oxides constitute ∼17% of the nanoparticles, indicating even a very small amount of Ti could promote the significant formation of Y–Ti complex oxides. Almost all of the nanoparticles are either coherent (∼98%) or semi-coherent with the matrix. The crystallographic orientation correlations of the oxides and matrix were determined. The formation mechanisms of various kinds of oxides and, moreover, the reasons of the unusual irradiation tolerance and thermal stability of the ODS steel are discussed based on the results.

    更新日期:2020-01-30
  • Improving impact toughness of heavy section reduced activation ferritic martensitic CLF-1 steel joints with electron beam welding
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-28
    Shikai Wu; Yilei Shi; Guoyu Zhang; Song Zhang; Hongbin Liao; Xiaoyu Wang; Zhiyun Qin

    In this paper, electronic beam butt welding of the reduced-activation ferrite/martenstic CLF-1 steel with a thickness of 32 mm was performed. Furthermore, the impact toughness, microstructure, phase compositions and impact fractures of the joints with different preheating, post welding heat treatment and heat inputs were investigated systematically and the influencing rules and mechanisms were analyzed. The experimental results shown that, with the optimization of the processes parameters, well-formed joints without any defects such as pores, incomplete fusion, and cracks was obtained by electronic beam welding, which also possessed favorable tensile and flexural performances. The impact absorbed energy of the joints was not remarkably improved with preheating or post welding heat treatment. As the heat input dropped from 2805 J mm−1 to 1440 J mm−1, the impact absorbed energy of the weld increased from 3.3–12.5 J–275 J after PWHT at 740 °C for 4 h. At a heat input of 1836 J mm−1, the impact absorbed energy of the weld was comparable to the value of BM, and the specimens were not fractured after impact test. Preheating and PWHT only changed the size of martensite laths and carbides without eliminating the residual δ-ferrites in the weld. In addition, the impact fractures exhibited brittle fracture characteristics. As the heat input reduced, the content of δ-ferrites in the weld decreased significantly, the martensite laths in the weld zone after thermal treatment were refined, and Ta-rich MX carbides in the laths were precipitated. Therefore, the impact toughness of the joints increased significantly and the impact fractures exhibited brittle fracture characteristics.

    更新日期:2020-01-30
  • Annealing behavior of hardening and ductility loss of a 16Cr–4Al ODS ferritic steel irradiated with high energy Ne ions
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-28
    Yuguang Chen; Chonghong Zhang; Zhaonan Ding; XianLong Zhang; Yitao Yang; Yin Song; Akihiko Kimura

    Annealing behavior of the hardening and ductility loss of an oxide-dispersion-strengthened (ODS) ferritic steel (16Cr–4Al) irradiated with high-energy heavy ions is studied. 20Ne ions with 123.4 MeV supplied by a cyclotron was used to produce a quasi-uniform atomic region of atomic displacement damage and Ne concentration from surface to a depth of 33 μm in the specimens. Two atomic displacement levels (0.16dpa, 0.70dpa) were approached at about 220 K. The irradiated specimens were subsequently thermally annealed at room temperature, 473 K and 673 K, respectively. Hardness and ductility of the specimens were investigated with nano-indentation technique and small-punch test. The data of the nano hardness were fitted by Nix-Gao model to obtain the bulk equivalent hardness values. The irradiated specimens show observable hardening, which recedes monotonously with the increase of annealing temperature. The Arrhenius plots show a good linearity in the entire temperature range investigated, with an apparent activation energy of 0.13 ± 0.01 eV for both the two damage levels. Assuming that the hardening is caused by the self-interstitial-atom (SIA) clusters initially produced by the cascade damage, an activation energy of 0.70 ± 0.05 eV is deduced for the migration and coalescence process of the SIA-clusters. The small punch test (SPT) of the high-dose specimens shows that a minor ductility loss occurred under conditions as-irradiated or subsequently thermally annealed at 473 K, while a remarkable increase of the ductility loss was observed after the thermal annealing at 673 K, which coincides with the formation of nano-scale gas bubbles in high density in the specimen. It is indicated that the formation of bubbles has minor on the irradiation hardening. A comparison of three ODS ferritic steels Ne-ion-irradiated to 0.7 dpa/260 appm(Ne) shows that the ductility loss decreases with the increase of the sink strength of the oxide dispersoids/ferritic substrate interfaces.

    更新日期:2020-01-30
  • Effects of contents of Al, Zr and Ti on oxide particles in Fe–15Cr–2W–0.35Y2O3 ODS steels
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-28
    Peng Dou; Shaomin Jiang; Lanlan Qiu; Akihiko Kimura

    FeCrAl oxide dispersion strengthened (ODS) steel is one of the most promising candidate cladding materials of generation IV nuclear reactors because of its excellent resistance to not only corrosion but also creep and irradiation due to the ultrahigh density nanometer-scale oxides. Effects of the contents of Al, Zr and Ti on the crystal and metal/oxide interface structures of the particles in Fe–15Cr–3.7Al–2W–0.1Ti–0.6Zr–0.35Y2O3 (3.7Al–0.1Ti–0.6Zr) and Fe–15Cr–3.8Al–2W–0.12Ti–0.5Zr–0.35Y2O3 (3.8Al–0.12Ti–0.5Zr) were studied by high resolution transmission electron microscopy (HRTEM). For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, phase identification was accomplished on 175 and 213 particles, respectively, which have peak number fraction and, therefore, represent the oxides contributing most significantly to the macroscopic properties of the ODS steels. The proportions of various types of oxides were determined. For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, the proportions of Y–Zr complex oxides are ∼87.4% and ∼54.4% while the number fractions of Y–Al complex oxides are ∼3.5% and ∼5.2%, respectively, indicating that 0.5–0.6 wt% Zr inhibits the formation of Y–Al complex oxides remarkably while prompts the significant occurrence of Y–Zr complex oxides. For 3.7Al–0.1Ti–0.6Zr and 3.8Al–0.12Ti–0.5Zr ODS steels, the proportions of Y–Ti complex oxides are ∼5.7% and ∼36.6%, respectively, indicating it is effective for increasing the proportion of Y–Ti complex oxides by increasing the content of Ti from 0.09 wt% to 0.12 wt% with the content of Zr decreased from 0.63 wt% to 0.49 wt%. The crystallographic orientation correlations of the oxides and matrix were determined. The formation and refinement mechanisms of oxides and, moreover, the reasons of the unusual irradiation tolerance and thermal stability of the ODS steel were discussed based on the results.

    更新日期:2020-01-30
  • The HICU PIE results of EU ceramic breeder pebbles: General characterization
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-27
    M.H.H. Kolb; J.M. Heuser; R. Rolli; H.-C. Schneider; R. Knitter; M. Zmitko

    The HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment was performed in the High Flux Reactor (HFR) in Petten, NL, in order to irradiate candidate tritium breeder materials in a fusion relevant environment. The presented work focuses on the post-irradiation examination of the irradiated lithium orthosilicate based breeder pebbles. The pebble samples showed three different contents of Li-6 and were irradiated at two different temperatures and in mechanically constrained and unconstrained state. In this particular publication, the influences of the irradiation conditions on the pebble morphology, microstructure, porosity, and mechanical strength are addressed. The results indicate that in general a high irradiation temperature seems to be advantageous for maintaining the mechanical strength of the irradiated pebbles. A higher mechanical strength and a significantly lower closer porosity is observed for samples that were irradiated at high temperatures in comparison to pebbles that were irradiated at low temperatures. The effects on the pebble properties with respect to the Li-6 content are small in contrast to effects of the irradiation temperature. With an increased Li-6 content, no deterioration of the material properties was observed, especially for samples irradiated at high temperatures.

    更新日期:2020-01-27
  • Role of electronic and magnetic interactions in defect formation and anomalous diffusion in δ-Pu
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-26
    Sarah C. Hernandez; Franz J. Freibert; Blas P. Uberuaga; John M. Wills

    Previous experimental work has shown self-irradiation in Pu solids produces point defect populations that correlate with increases in local disorder, long-range structural changes, induced magnetic moments, and other thermo-physical property changes. Thermally activated kinetic processes drive these defects to diffuse and interact toward either damage evolution or lattice recovery. Using DFT and cNEB, as implemented in VASP, migration barriers for mono-vacancy and split-interstitial diffusion and Frenkel pair recombination were calculated in fcc δ-Pu. The results indicate the migration barrier of a monoclinic mono-vacancy is lower when compared to the migration barrier of a tetragonal split-interstitial in δ-Pu, contrary to typical fcc metal point defect migration. This fundamentally different diffusion mechanism is a result of local symmetry breaking induced by electronic and magnetic interactions leading to the development of Pu–Pu short bonds (<3.0 Å) within a many-atom complex defect forming and migrating. The migration of the monoclinic mono-vacancy maintains short bonds with anti-parallel spins throughout the transition; whereas, during the migration transition state for the tetragonal split-interstitial, formation of short bonds with parallel spins and a spin-flip of the migrating Pu interstitial occurs. The associated energy cost is reflected in an increase in the migration barrier energy. Frenkel pair recombination is not spontaneous at 0K, but correlates with magnetic moment interactions, leading to an energy barrier for recombination. From these results, it is concluded that migration of defects in unalloyed δ-Pu are highly dependent on the electronic and magnetic interactions that induce associated low-symmetry structures and consequently influence the diffusional properties. Typical fcc defect diffusion mechanisms do not apply to the monoclinic mono-vacancy and tetragonal split-interstitial in the complex 5f δ-Pu system suggesting that the experimental observation of radiation damage induced localized magnetic moments and anomalous diffusion properties measured in δ-Pu could be understood in terms of defect kinetics and interactions.

    更新日期:2020-01-26
  • The HICU PIE results of EU ceramic breeder pebbles: Tritium release properties
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-26
    J.M. Heuser; M.H.H. Kolb; R. Rolli; H.-C. Schneider; R. Knitter; M. Zmitko

    The former EU reference lithium orthosilicate based breeder pebbles were exposed to neutron irradiation in the HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment to test their stability and tritium release properties under DEMO relevant conditions. The samples, varying by three different Li-6 contents, were exposed to irradiation at two different temperatures and the pebbles were either pre-compacted or not. This second part of the post-irradiation examination is focussing on the tritium release behaviour of the ceramic breeder pebbles. The irradiation temperature has the strongest influence on the tritium release behaviour. The tritium inventory is significantly higher for samples that were irradiated at low temperatures. A clear trend regarding higher release rates with increasing Li-6 content was not observed. Tritium is released in a multi-staged process as HTO, HT or corresponding fragments. Fits based on the Wigner-Polanyi equation suggest that recombination reactions of tritium with adsorbed species on the pebble's surface play the dominant role in the release process. However, the probability for the recombination of two adsorbed T-species on the surface seems to be too low, as no reliable signal for T2 was detected.

    更新日期:2020-01-26
  • Oxidation and passivation of U(AlxSi1-x)3 alloy at elevated temperatures
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-25
    S. Cohen; M. Matmor; M. Vaknin; G. Rafailov; O. Appel; L. Shelly; N. Shamir; S. Zalkind

    The surface composition of U(AlxSi1-x)3 alloy (x = 0.57) and its interactions with oxygen, at elevated temperatures, were studied, utilizing Auger electron spectroscopy, X-Ray photoelectron spectroscopy and direct recoil spectrometry. Heating the alloy in ultra-high vacuum, results in aluminum (and some silicon) segregation to the surface, forming, above 700 K, a ∼0.6 nm self-assembly capping layer. Exposing the surface alloy to oxygen, at temperatures up to 500 K, causes oxidation of the uranium and the aluminum components, while silicon is only slightly oxidized. Above 600 K, only the aluminum segregated overlayer is oxidized, forming a passivation layer that inhibits further oxidation of the alloy.

    更新日期:2020-01-26
  • Synthesis and characterization of zirconolite-sodium borosilicate glass-ceramics for nuclear waste immobilization
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-25
    Hanzhen Zhu; Fu Wang; Qilong Liao; Yongchang Zhu

    Zirconolite-sodium borosilicate glass-ceramics were successfully prepared via slow-cooling methods, and the crystallization, microstructure and aqueous durability have been investigated with powder X-ray diffraction (XRD), backscattered scanning electron microscopy and energy dispersive spectroscopy (BSE-EDS), Raman spectroscopy and ASTM Product Consistency Test leaching method. The results show that the main crystalline phase in the prepared glass-ceramics is strip-shaped zirconolite phase and the quantitative fraction of the zirconolite phase is about 30 wt%, the chemical composition of zirconolite crystals grown from a glass matrix is determined by Rietveld refinement to be Ca0.93Zr0.76Ce0.31Ti1.95Al0.05O7 and 84.53% Ce are incorporated in zirconolite crystals. Moreover, the aqueous durability test shows the low normalized leaching rates of Si (LRSi), Ca (LRCa) and Ce (LRCe) of the glass-ceramics, and LRSi, LRCa and LRCe are about 4 × 10−4, 1 × 10−4 and 8 × 10−7 g m−2 d−1, respectively after 56 leaching days. The obtained conclusions provide useful information for the immobilization of high-level radioactive wastes by using borosilicate glass-ceramic as potential matrix through one-step method.

    更新日期:2020-01-26
  • Force-depth relationships with irradiation effect during spherical nano-indentation: A theoretical analysis
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-24
    Xiazi Xiao; Cewen Xiao; Xiaodong Xia

    A mechanistic model is developed for the force-depth relationship of ion-irradiated materials, which is conducted by spherical nano-indentation. With irradiation effect, the pop-in phenomenon almost disappears that is ascribed to the irradiation-induced defects serving as dislocation nucleation sites that facilitate the generation of new dislocations. After materials yielding, the evolution of statistically stored dislocations, geometrically necessary dislocations and irradiation-induced defects mutually contributes to the force-depth relationships with irradiation effect. Thereinto, the increase of loading force originates from the impediment of slipping dislocations by irradiation-induced defects. By comparing with the experimental data of Fe–12Cr alloy, a reasonable agreement is achieved.

    更新日期:2020-01-24
  • Significant growth of vacancy-type defects by post-irradiation annealing in neon ion-irradiated tungsten probed by a slow positron beam
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-23
    A. Yabuuchi; M. Tanaka; A. Kinomura

    Irradiation damage and its evolution in noble gas ion-irradiated tungsten have not been investigated in detail other than in the case of helium ion irradiation. In this study, irradiation-induced vacancy-type defects in helium ion- and neon ion-irradiated tungsten were investigated by using a slow positron beam, and their annealing behavior in the temperature range of 20∘C-900∘C was compared by characterizing the Doppler broadening of positron annihilation radiation spectra. In helium ion-irradiated tungsten, slight aggregation of irradiation-induced vacancy-type defects was observed upon annealing, but eventually, a large portion of the vacancy clusters was eliminated after annealing at 900∘C. In contrast, in neon ion-irradiated tungsten, irradiation-induced vacancy-type defects were observed to aggregate significantly at 300∘C and 600∘C. In addition, the large vacancy clusters formed by the aggregation survived even after annealing at 900∘C.

    更新日期:2020-01-24
  • Vacancy cluster growth and thermal recovery in hydrogen-irradiated tungsten
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-23
    M. Zibrov; W. Egger; J. Heikinheimo; M. Mayer; F. Tuomisto

    The thermal evolution of vacancies and vacancy clusters in tungsten (W) has been studied. W (100) single crystals were irradiated with 200 keV hydrogen (H) ions to a low damage level (5.8×10−3 dpa) at 290 K and then annealed at temperatures in the range of 500–1800 K. The resulting defects were characterized by positron annihilation lifetime spectroscopy (PALS) and positron annihilation Doppler broadening spectroscopy (DBS). Annealing at 700 K resulted in the formation of clusters containing 10–15 vacancies, while at 800 K and higher temperatures clusters containing about 20 vacancies or more were formed. Reduction of the defect concentration likely accompanied by further coarsening of the clusters started at 1300 K and ended at 1800 K with the complete defect recovery. The determined cluster sizes at 700 K and 800 K were larger than the estimated minimum cluster sizes that are thermally stable at these temperatures, indicating that the migration and ensuing coalescence of small clusters plays an important role in cluster growth.

    更新日期:2020-01-24
  • 更新日期:2020-01-23
  • MOX fuel microstructural evolution during the VERDON-3 and 4 tests
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-23
    C. Le Gall; S. Reboul; L. Fayette; T. Blay; I. Zacharie-Aubrun; I. Félines; K. Hanifi; I. Roure; P. Bienvenu; F. Audubert; Y. Pontillon; Jean-Louis Hazemann

    The VERDON-3 and -4 tests were part of the VERDON-ISTP programme that aimed at studying the fuel and fission products (FP) behaviour in severe accident conditions. The main objective of these two complementary tests was the study of MOX fuel behaviour and FP release under oxidising (VERDON-3) and reducing (VERDON-4) conditions at very high temperature (>2300 °C). Complementary to the on-line gamma spectrometry measurements performed during the two tests, post-test characterisations were carried out in order to tackle these tasks. The two samples recovered after the VERDON-3 and -4 tests were compared to a third one extracted from the same father rod and left as irradiated. This comparison enabled to highlight the effect of temperature and atmosphere on the fuel behaviour. These three samples were characterised by several techniques available at the LECA-STAR facility of the CEA Cadarache. Experimental observations showed that an interaction between the fuel and the cladding occurred in both types of conditions by interdiffusion mainly between U and Zr. This phenomenon led to the formation of a UyZr1-yO2±x cubic phase at the periphery of the fuel pellet which melted in the VERDON-4 test conditions, penetrating through the cracks of the sample and dissolving the fuel matrix. No liquid was formed during the VERDON-3 test despite the formation of a large fuel-cladding interaction zone.

    更新日期:2020-01-23
  • Corrosion studies of a low alloyed Fe–10Cr–4Al steel exposed in liquid Pb at very high temperatures
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-23
    Peter Dömstedt; Mats Lundberg; Peter Szakalos

    The aim of the work has been to study the corrosion resistance of a new low alloyed experimental FeCrAl steel, with the potential use as corrosion barrier in high temperature lead based energy applications. The exposures were conducted in liquid lead at 800 °C and 900 °C, with controlled oxygen environment, for up to 1760 h. The results demonstrate that the new experimental alloy had formed a protective oxide in both exposures, with no indications of lead penetration. The alloy showed better corrosion properties than that of the reference materials: Kanthal APM™, Kanthal APMT™ and AISI 316L. This indicates that the ductile Fe–10Cr–4Al-RE steel can be used as a corrosion barrier in liquid lead based clean energy applications, operating at very high temperatures.

    更新日期:2020-01-23
  • Quantification of the constitutive relationship of high-energy heavy-ion irradiated SS316L using the small punch test
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-22
    Xianlong Zhang; Chonghong Zhang; Zhaonan Ding; Yuguang Chen; Liqing Zhang

    Heavy-ion irradiation has been widely used to simulate neutrons-irradiation induced damage effects, due to its higher damage rate and lower activation of the post-irradiation manipulation. The high-energy ion beam is capable of producing a thickness of dozens of microns damaged region in steel, making it possible to evaluate the macroscopic mechanical properties of the irradiated specimens. In the present paper, a method to quantify the constitutive relationships of high-energy heavy-ion irradiated steels is proposed. Ni22+ ions with a kinetic energy of 357.86 MeV provided by a cyclotron were used to produce a quasi-homogeneous atomic displacement damaged layer (about 25 μm in thickness) in specimens of 316 L stainless steel. The temperature of the specimens were kept at about −50 °C during ion irradiation. Two damage levels of 0.16 and 0.33 displacement per atom (dpa) were approached. Small punch test of the unirradiated and irradiated ϕ3 mm disk samples was carried out to obtain the load-deflection curves. A series of finite element simulation of SPT of the laminated irradiated samples, in combination with sequential programming algorithm, was performed to characterize the constitutive relationships of the irradiation damaged layer of the samples. Finite element simulations with obtained constitutive relationships show agreement with the experimental results. Nanoindentation tests were carried out to verify the identified constitutive relationships. The nanoindentation results show an irradiation induced hardening in good agreement with that from the obtained constitutive relationships.

    更新日期:2020-01-22
  • Study on microstructure and mechanical property of linear friction welding on 9Cr reduced activation ferrite/martensite steel
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    Chenxi Liu; Yan Gao; Xiaohua Li; Wenchao Li; Kefu Gan

    The present work is concerned with the application of linear friction welding (LFW) process accessible for the weld-jointing of 9Cr reduced activation ferrite/martensite (RAFM) steels. Optical and electron microscopic characterization of microstructures were performed in various regions of the weld joint. Hardness, tensile and Charpy impact tests on the joined samples were processed to examine the mechanical property and reliability of the weld joints. It indicates that, hot plastic deformation induced by linear friction triggers continual dynamic recrystallization in the weld zone, along with high-density dislocation substructures formed by such severe deformation, which leads to a good combination of mechanical performances in the weld joint. Such linear friction welding comes beyond the rotational process restriction in conventional friction stir welding, and avoid significant oxide inclusions, porosities and coarsened grains brought by heat input as well. The work proves that the present LFW technique works well in the welding of 9Cr RAFM steels and inspires us of a future study on optimizing process parameters of the welding process for a better performance.

    更新日期:2020-01-22
  • Interatomic potentials of W–V and W–Mo binary systems for point defects studies
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    Yangchun Chen; Xichuan Liao; Ning Gao; Wangyu Hu; Fei Gao; Huiqiu Deng

    Interatomic potentials for tungsten–vanadium (W–V) and tungsten–molybdenum (W–Mo) binary systems have been developed based on Finnis–Sinclair formalism. The potentials are based on an accurate previously developed potential of pure W. Potential parameters of V–V, Mo–Mo, W–V and W–Mo were determined by fitting to a large database of experimental data as well as first principle calculations. These potentials were able to describe various fundamental physical properties of pure V and Mo, such as a lattice constant, cohesive energy, elastic constants, bulk modulus, vacancy and self-interstitial atom formation energies, stacking fault energies and a relative stability of <100> and ½<111> interstitial dislocation loops. Other fundamental properties of the potentials described included alloy behaviours, such as the formation energies of substitutional solute atoms, binding energies between solute atoms and point defects, formation energies and lattice constants of artificial ordered alloys. These results are in reasonable agreement with experimental or first principle results. Based on these results, the developed potentials are suitable for studying point defect properties and can be further used to explore displacement cascade simulations.

    更新日期:2020-01-22
  • Measurement of H and E within and in the neighborhood of a single hydride platelet in Zircaloy-2
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    K.O. Kese; U.D. Hangen; W. Grünewald; D. Jädernäs; A.-M. Alvarez; E. Broitman; J.K.-H. Karlsson

    A single hydride platelet and the matrix material next to it in a Zircaloy-2 cladding have been targeted for hardness, H, and Young's modulus, E, measurement using nanoindentation. The results were compared with those obtained in the matrix material far away from the hydride. The results show that hardness and Young's modulus in the hydride are higher than those of the matrix adjacent to the hydride, which are the same as those of the matrix far away from the hydride.

    更新日期:2020-01-22
  • Study on the immobilization of cesium absorbed by copper ferrocyanide using allophane through pressing/sintering method
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    Yang Cheng; Yiqi Wang; Hongji Sang; Yan Wu; Yuezhou Wei

    The development of the Cs stable immobilization method contributes to the treatment of insoluble ferrocyanide sludge containing radioactive Cs (Cs–CuFC) generated by the Fukushima nuclear accident. Cs–CuFC begins to decompose at 200 °C, and Cs in Cs–CuFC volatilizes after being oxidized to Cs2O in air at high temperature. Therefore, Cs has an immobilization ratio of less than 4% at 1000 °C. Meanwhile, Cs–CuFC generates poisonous gases, such as HCN, NO, and NH3, and the release of HCN is suppressed in active atmosphere. For the stable immobilization of Cs, a pressing/sintering method that uses allophane as an additive is examined. Allophane is mixed uniformly with Cs–CuFC in a mass ratio of 1:1, and the mixture is sintered at different temperatures to obtain solidified bodies. A stable crystal called pollucite is formed after sintering above 900 °C, and the immobilization ratio of Cs is approximately 100%. Part of pollucite is concentrated on some spots on the surface of solidified bodies. These bodies have good mechanical properties for geological storage. The leaching percentage of Cs for the solidified bodies sintered at 1100 °C in distilled water is less than 0.01% and 0.4% at 25 °C and 90 °C, respectively, thereby indicating that the solidified bodies have excellent immobilization properties and chemical stability.

    更新日期:2020-01-22
  • Investigation of anisotropic hardening response in a 12Cr-ODS ferritic steel subjected to 2.8 MeV Fe2+ irradiation
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    H.L. Yang; S. Kano; J.J. Shen; J. McGrady; Y.F. Li; D.Y. Chen; K. Murakami; H. Abe

    The anisotropic hardening response in a 12Cr-ODS ferritic steel was investigated before and after irradiation with 2.8 MeV Fe2+ at room temperature up to displacement damages of 0.5 and 15 dpa. For post irradiation examination, techniques of nano-indentation and orientation imaging microscopy were jointly applied to link crystal orientation with nano-hardness. Results showed that the averaged hardness of normal direction-transverse direction (TD-ND) specimen is less than that of rolling direction-transverse direction (RD-TD) specimen irrespective of the dose of irradiation damage. However, the amount of irradiation-induced hardening is observed to be weaker in TD-ND specimen relative to RD-TD specimen. These anisotropic phenomena are considered to be mainly attributed to the elongated grain structure with a very high grain aspect ratio of the present steel, in which smaller-sized grains are exhibited in TD-ND plane nevertheless fairly coarse grains are in RD-TD plane. In addition, the orientation dependent hardness and irradiation-induced hardening were confirmed. Specifically, the hardness of [001]-oriented grain is lower than that of [111]-oriented grain with and without irradiation. With the presence of irradiation, it was found that the extent of hardening is more obvious at [001]-oriented grain than [111]-oriented grain, which is attributed to a more activated primary {110}<111> slip beneath the indenter when tested at [100]-oriented grain compared with [111]-oriented grain.

    更新日期:2020-01-22
  • Proton irradiation and characterization of additively manufactured 304L stainless steels
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-21
    B.P. Eftink; J.S. Weaver; J.A. Valdez; V. Livescu; D. Chen; Y. Wang; C. Knapp; N.A. Mara; S.A. Maloy; G.T. Gray

    Irradiations were performed with 1.5 MeV protons to 0.6 dpa at 40–150 °C on additively manufactured (AM) 304L stainless steel and the changes in microstructure and mechanical behavior after irradiation were compared to wrought 304L stainless steel. All microstructural and hardness results after irradiation suggest the samples evolve toward a similar state, despite significant differences in the unirradiated microstructures and hardness values. A TEM and nanoindentation-based investigation of before and after proton irradiation at 40–150 °C is presented. Results are interpreted in terms of initial dislocation content, dislocation structures, and microstructural and chemical homogeneity.

    更新日期:2020-01-21
  • Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-20
    Riley J. Parrish; Fabiola Cappia; Assel Aitkaliyeva

    This work presents a comparison between the microstructural evolution of three annular fast-reactor mixed-oxide (MOX) fuel pellets irradiated to varying burnups at the Fast Flux Test Facility (FFTF). Fuel pellets irradiated to 3.4%, 13.7%, and 21.3% fissions per initial metal atom (FIMA) were examined using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) techniques. The cross-section of the low burnup pellet displayed minor structural changes, but the central annulus of the pellets at 13.7% and 21.3% FIMA shrank from their starting size. The high burnup fuel pellet featured streaking and porosity migration associated with columnar grain growth. The radial fission product distribution in each of the pellets had a higher number density of metallic particles >5 μm in diameter near the fuel centerline. Solid fission products in the fuel-cladding gap were observed in the low and intermediate burnup pellets. The low burnup sample showed minor accumulation of Ba in the gap, while the volatile Cs was primarily observed at the pellet surface. The intermediate burnup pellet displayed a porous mixture of fission products, consistent with the joint-oxide gain (JOG) that has been previously observed in fast-reactor MOX fuel pellets.

    更新日期:2020-01-21
  • Radiation damage tolerance of a novel metastable refractory high entropy alloy V2.5Cr1.2WMoCo0.04
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-18
    Dhinisa Patel; Mark D. Richardson; Bethany Jim; Shavkat Akhmadaliev; Russell Goodall; Amy S. Gandy

    A novel multicomponent alloy, V2.5Cr1.2WMoCo0.04, produced from elements expected to favour a BCC crystal structure, and to be suitable for high temperature environments, was fabricated by arc melting and found to exhibit a multiphase dendritic microstructure with W-rich dendrites and V–Cr segregated to the inter-dendritic cores. The as-cast alloy displayed an apparent single-phase XRD pattern. Following heat treatment at 1187 °C for 500 h the alloy transformed into three different distinct phases - BCC, orthorhombic, and tetragonal in crystal structure. This attests to the BCC crystal structure observed in the as-cast state being metastable. The radiation damage response was investigated through room temperature 5 MeV Au+ ion irradiation studies. Metastable as-cast V2.5Cr1.2WMoCo0.04 shows good resistance to radiation induced damage up to 40 displacements per atom (dpa). 96 wt% of the as-cast single-phase BCC crystal structure remained intact, as exhibited by grazing incidence X-ray diffraction (GI-XRD) patterns, whilst the remainder of the alloy transformed into an additional BCC crystal structure with a similar lattice parameter. The exceptional phase stability seen here is attributed to a combination of self-healing processes and the BCC structure, rather than a high configurational entropy, as has been suggested for some of these multicomponent “High Entropy Alloy” types. The importance of the stability of metastable high entropy alloy phases for behaviour under irradiation is for the first time highlighted and the findings thus challenge the current understanding of phase stability after irradiation of systems like the HEAs.

    更新日期:2020-01-21
  • Solubility and precipitation investigations of UO2 in LiF–BeF2 molten salt
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-18
    Hao Peng; Wei Huang; Leidong Xie; Qingnuan Li

    The solubility of UO2 in molten LiF–BeF2 (2:1 mol) (FLiBe) eutectic salt at 600 °C was studied by chemical and electrochemical methods. The results of dissolution experiment showed that the saturated solubility of UO2 in this melt was 2.37 × 10−3 mol/kg and its corresponding apparent solubility product (Ksp') was approximately 1.67 × 10−5 mol3/kg3. When more Li2O were added to the FLiBe–UF4 system, the cathodic peak current of U(IV) in the electrochemical cyclic voltammetry (CV) curve accordingly decreased because of precipitate formation. The precipitate corresponded to UO2 as determined by stoichiometric ratio (concentration variation of U4+ and O2−) and X-ray diffraction (XRD) analysis. Compared to the chemical analysis method, the CV technique was confirmed to be more feasible for accurate determination of concentration of U(IV). Meanwhile, the Ksp' value was also obtained to be 1.33 × 10−5 mol3/kg3 during the whole oxide titration procedure, which was highly consistent with that from the dissolution experiment. With the value of Ksp', the allowable amount of dissolved oxide ions (oxide tolerance) can be theoretically estimated in the FLiBe–UF4 system.

    更新日期:2020-01-21
  • Accelerated Monte Carlo method for calculation of sink strengths of absorbing surfaces for 3-D migrating particles in crystals of the cubic system
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-18
    A.B. Sivak; P.A. Sivak; V.M. Chernov

    An accelerated (compared to the standard “residence-time” Algorithm) Monte Carlo method for the calculation of the sink strengths of absorbing surfaces for particles in crystals of the cubic system has been suggested. On its basis, several algorithms have been developed which allow one to calculate the sink strengths either without any systematic inaccuracy or with its low and controlled magnitude. These algorithms have been tested by calculating the sink strengths of absorbing surfaces of different geometry (spherical, toroidal, cylindrical and planar) for particles (self-interstitial atoms, vacancies) in a crystal with BCC lattice. The use of the developed algorithms accelerates the calculations for low volume fractions of absorbers by orders of magnitude.

    更新日期:2020-01-21
  • Behaviour of (U,Am)O2 in oxidizing conditions: A high-temperature XRD study
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-17
    E. Epifano; R. Vauchy; F. Lebreton; A. Joly; C. Guèneau; Ch Valot; P.M. Martin

    Uranium–Americium oxides U1−yAmyO2±x are currently investigated as possible transmutation targets for next generation nuclear reactors. In the context of a comprehensive investigation of the thermodynamic and thermal properties of these materials, their behaviour in oxidizing conditions is here investigated for the first time. The results of high-temperature X-ray diffraction measurements in air are here presented. A wide composition domain of the solid solution has been investigated, measuring U1−yAmyO2±x oxides with Am/(Am + U) ratios ranging from 0.10 to 0.67. This allowed determining the effect of the americium content on the oxidation kinetics in air. Specifically, it will be shown that americium hinders the formation of the M4O9 and M3O8 phases.

    更新日期:2020-01-21
  • Corrosion of commercial alloys in FLiNaK molten salt containing EuF3 and simulant fission product additives
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-17
    Samuel W. McAlpine; Natasha Skowronski; Weiyue Zhou; Guiqiu Zheng; Michael P. Short

    In liquid–fuel molten salt reactor designs, salt–facing materials will be exposed to a molten salt containing a multitude of fission products and other corrosive species. Currently, little work has been done to understand the unique corrosion characteristics of materials in liquid–fuel systems. In this study, we conducted corrosion experiments up to 150 h in duration which exposed four commercial alloys (Hastelloy N, Incoloy 800H, 316L stainless steel, and Ni–201) to 3 molten salt compositions in order to better understand corrosion in liquid–fuel systems and inform reactor design going forward. It was found that the presence of simulant fission product species in a highly corrosive FLiNaK + EuF3 molten salt does not lead to any detectable increase in the extent of corrosion at reactor–relevant conditions. No penetration of simulant fission product species into the samples was detected. The unique corrosion morphology of each of the alloys tested in this work is discussed. In particular, Ni–201 was found to be an ideal salt–facing material in molten fluoride systems and is essentially immune to corrosion.

    更新日期:2020-01-21
  • Atomistic simulations of a helium bubble in silicon carbide
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-13
    L. Pizzagalli; M.-L. David

    Large scale molecular dynamics calculations have been carried out to investigate the properties of nanometric helium bubbles in silicon carbide as a function of helium density and temperature. A dedicated interatomic potential has been developed to describe the interactions between helium and SiC atoms. The simulations revealed that the helium density cannot exceed a certain threshold value, which depends on temperature, because of the plastic deformation of the SiC matrix. Both local amorphization at low temperatures, and nucleation and propagation of dislocations at high temperatures, have been identified as activated plasticity mechanisms. This work also predicts that very high pressure, up to 60 GPa could be reached in helium bubbles in silicon carbide.

    更新日期:2020-01-14
  • Effect of irradiation on nanoprecipitation in EM10 alloy - Comparison with Eurofer97
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-13
    O. Tissot; G. Sakr; C. Pareige; J. Henry

    Atom Probe Tomography investigations of EM10 and EUROFER alloys after both 1 MeV electrons and 2 MeV Fe2+ ions irradiations at 300 °C and up to doses of 0.6 dpa were performed. SiNiPMn(-Cu) enriched clusters were observed only in EM10 alloy. Phosphorus was found to be necessary for cluster formation. Segregation of Si, Ni, P, Mn elements were measured at a Grain Boundary and a dislocation. Cr clusters near a dislocation were noticed.

    更新日期:2020-01-14
  • Atomistic simulation study of clustering and evolution of irradiation-induced defects in zirconium
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-13
    Christopher Maxwell; Jeremy Pencer; Edmanuel Torres

    Zirconium (Zr) alloys have been widely used as structural materials for in-core components in water-cooled nuclear reactors. During normal operation, these materials are exposed to a neutron flux in the core of the reactor, resulting in material degradation such as irradiation-induced anisotropic growth, thus affecting their performance in the long-term. Experimental and theoretical studies have shown that irradiation-induced defects in zirconium lead to the formation of defect clusters and loops. The anisotropy in the migration of defects has been suggested to play an important role in irradiation growth in pure Zr and its alloys. However, the mechanisms that govern the microstructural evolution that lead to the observed anisotropic growth of Zr is still unclear. In the present work, we perform a molecular dynamics simulation study of irradiation-induced lattice defects in Zr to investigate the formation of clusters and loops. Irradiation-induced damage is modeled by constrained stochastic formation of vacancies and self-interstitial atoms in bulk Zr. Using this approach, the formation and evolution of defect clusters and loops were determined. The dynamic properties of lattice defect structures were investigated through the evaluation of their migration and diffusivity. We found that the diffusivity of vacancy and interstitial clusters is anisotropic and slow, while the diffusivity of large loops is relatively high and confined to the 〈a〉 plane.

    更新日期:2020-01-14
  • Temperature effect on fracture toughness of CLF-1 steel with miniature three-point bend specimens
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-13
    Yao Xie; Lei Peng; Wangzi Zhang; Hongbin Liao; Guian Qian; Yuanxi Wan

    Fracture toughness is one important mechanical property of reduced activation ferritic/martensitic (RAFM) steels which are primary candidate structural materials applied to fusion reactors. Temperature effect on fracture toughness of the Chinese low activation ferritic/martensitic (CLF-1) steel was investigated in the range of 25–550 °C with miniature three-point bend (3 PB) specimens, using the digital image correlation (DIC) method to measure load-line displacement. Results show that the fracture toughness J0.2(B) of CLF-1 steel decreases from 25 °C to 450 °C and increases from 450 °C to 550 °C. This changing trend with temperature is similar to that of some commercial ferritic/martensitic (F/M) steels and consistent with the temperature dependence of its ductility which is total elongation obtained from tensile testing. The fracture toughness minimum at 450 °C could be attributed to the deterioration of ductility, where the fracture surfaces with few typical dimples indicated quasi-cleavage-like features.

    更新日期:2020-01-14
  • Hot deformation behavior and processing map of Zr-4 alloy
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-12
    Jianjun Liu; Kelu Wang; Shiqiang Lu; Xiayun Gao; Xin Li; Feng Zhou

    The hot deformation behavior of Zr-4 alloy at the deformation temperature range of 750–1000 °C and the strain rate range of 0.001–10 s−1 was studied on a Gleeble-3500 thermal simulator. The results show that the flow stress increases with the increasing strain rates or the decreasing deformation temperature. Based on the experimental data, the strain-compensated constitutive equation was established to predict flow stress during different strains, strain rates and temperatures. Meanwhile, the hot deformation activation energy of Zr-4 alloy was respectively calculated to be 224.31 kJ/mol, 593.50 kJ/mol and 345.71 kJ/mol in α single-phase region, α+β two-phase region and β single-phase region, which are obviously much higher than the activation energy of pure zirconium (113 kJ/mol). It indicates that the main deformation mechanism is not the dynamic recovery but other deformation mechanisms. According to the dynamic material model and Murty instability criterion, Murty processing maps have been constructed at the true strain of 0.6 and 1.2. Moreover, by combining microstructural observations, the areas of 750–880 °C/0.01–0.32 s−1 and 900–1000 °C/0.03–1 s−1 are identified to be the optimum hot working parameters.

    更新日期:2020-01-13
  • Radiation damage in uranium dioxide: Coupled effect between electronic and nuclear energy losses
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-11
    Marion Bricout; Claire Onofri; Aurélien Debelle; Yves Pipon; Renaud C. Belin; Frédérico Garrido; Frédéric Leprêtre; Gaëlle Gutierrez

    A coupling between the nuclear and electronic energy losses occur in the nuclear fuel (UO2) during in-reactor operations. However, the underlying mechanisms involved are still to be investigated. In this work, synergistic effects of nuclear and electronic energy losses have been investigated by irradiating crystals with single (900 keV I ions or 27 MeV Fe ions) and dual (900 keV I ions and 27 MeV Fe ions, simultaneously) ion beams at the JANNUS-Saclay facility. The damage build-up kinetic was in situ characterized by Raman spectroscopy. The microstructure evolution was determined by transmission electron microscopy (TEM) observations and by X-ray diffraction (XRD) analysis. Results show that both crystalline disorder and strain level are lower under dual-beam compared to the single-beam ion irradiations. Indeed, the dual-beam irradiation induces a transition from the formation of dislocation loops to dislocation lines. This result can be explained, in the framework of the thermal spike model, by a local increase of the temperature along the high-energy ion path. This temperature increase likely induces an enhanced defect migration leading to defect rearrangement.

    更新日期:2020-01-11
  • A logical approach for zero-rupture Fully Ceramic Microencapsulated (FCM) fuels via pressure-assisted sintering route
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-11
    Caen Ang; Lance Snead; Yutai Kato

    A pathway to Fully Ceramic Microencapsulated (FCM™) fuel pellets showing absence of sintering-derived fuel “rupture” has been demonstrated. In the typical FCM manufacturing process, TRistructral ISOtropic (TRISO) particles display statistically significant rupture events. Rupture is caused by contact of particles during the axial shrinkage of fuel pellet that accompanies the pressure-assisted sintering process. To solve this, template SiC powder discs were fabricated to host planes of TRISO particles, and the disks are stacked to form a cylindrical “green” pellet. After sintering, it showed that up to ∼34% packing fraction of particles (Vp) is feasible without contact between planes. Sintering was shown to reduce the axial displacement between planes of TRISO, and XCT showed planes separated by a displacement of ∼100 μm. XCT, optical microscopy and SEM showed the very limited radial displacement of particles. However, the relative density of the FCM pellet was limited to ∼95%. The current results support this zero-rupture concept as viable, but perturbations to TRISO arrangements and limited matrix density require further effort, in order to improve FCM fuel performance.

    更新日期:2020-01-11
  • On the O-rich domain of the U-Am-O phase diagram
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-11
    E. Epifano; R. Vauchy; F. Lebreton; R. Lauwerier; A. Joly; A. Scheinost; C. Guéneau; Ch Valot; P.M. Martiny

    Uranium–Americium oxides U1−yAmyO2±x are promising candidates as possible transmutation targets for next generation nuclear reactors. In the context of a comprehensive investigation of their thermodynamic and thermal properties, the behaviour in oxidizing conditions is here studied. In a recent work, the behaviour in air of stoichiometric and sub-stoichiometric U1−yAmyO2−x compounds, with various Am content, was investigated by high-temperature X-ray Diffraction. Herein, the hyper-stoichiometric oxides obtained from that study are investigated by X-ray Absorption Spectroscopy. The new data, together with the previous XRD results, allow determining the exact compositions of the samples and hence obtaining phase diagram points in the O-rich domain of the U-Am-O system. Indeed, five phase diagram points at 1473 K are obtained: two tie-lines in the M4O9-M3O8 domain, for Am/(Am + U) = 0.10 and 0.15, one tie line in the MO2+x-M3O8 domain, for Am/(Am + U) = 0.28, and two points in the single phase MO2±x domain, for higher americium concentration. From these data, it is also concluded that trivalent americium has a small solubility in the M4O9 and M3O8 phases.

    更新日期:2020-01-11
  • Thermo-mechanical behavior of Zircaloy-4 claddings under simulated post-DNB conditions
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-10
    T. Jailin; N. Tardif; J. Desquines; P. Chaudet; M. Coret; M.-C. Baietto; V. Georgenthum

    The thermo-mechanical behavior of Zircaloy-4 claddings under simulated post-DNB RIA conditions was investigated. Around twenty experiments were performed in simulated post-DNB conditions, i.e. creep ballooning tests with heating rates greater than 1000 °C/s. Two different levels of pressure of 7 and 11 bar were tested for temperatures of interest ranging from 840 °C to 1020 °C. A complex creep behavior was highlighted in this range of temperature. It appears very well correlated to the phase content present within the material during fast thermal transients. Tests with low thermal transients were also performed and evidence a strong impact of the heating rate on the thermo-mechanical properties of the claddings.

    更新日期:2020-01-11
  • Simulation of the chemical state of high burnup (U,Pu)O2 fuel in fast reactors based on thermodynamic calculations
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-10
    Karl Samuelsson; Jean-Christophe Dumas; Bo Sundman; Jérôme Lamontagne; Christine Guéneau

    In this paper, the chemical state of fast reactor (U,Pu)O2 fuel at high burnup has been simulated using OpenCalphad software and the TAF-ID thermodynamic database. This has been done in order to evaluate the combination of software and database for further implementation into the Germinal fuel performance code of the Pleiades simulation platform. The results have been compared with post irradiation examinations (PIE) of fuel samples from the Phénix sodium-cooled fast reactor. The calculations performed from isotopic data compositions were able to predict all precipitates encountered in the PIE, as well as several other phases. When possible, the measured composition of the phases were compared with the simulations, and show a good similarity in this regard. Additionally, calculations based on measured composition in the fluorite (U,Pu) O2 phase have been performed at different temperatures and oxygen-to-metal ratios. Here, the calculations predict that the formation of fission product oxide compounds occurs to a greater extent in the cooler fuel periphery, which is also what experiments have shown. Oxygen potential has been calculated and compared with experiments with similar composition. The calculations are considered fast and reliable enough for implementation of the thermodynamic software into the fuel performance code.

    更新日期:2020-01-11
  • Electron microscopy characterization of fast reactor MOX Joint Oxyde-Gaine (JOG)
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-10
    F. Cappia; B.D. Miller; J.A. Aguiar; L. He; D.J. Murray; B.J. Frickey; J.D. Stanek; J.M. Harp

    The composition and crystal structure of the “Joint Oxyde Gaine” (JOG) has been investigated by means of electron microscopy. Microstructural characterization reveals a highly heterogeneous porous structure with inclusions containing both fission products and cladding components. Major fission products detected, other than Cs and Mo, are Te, I, Zr and Ba. The layer is composed by sub-micrometric crystallites. The diffraction data refinement, together with chemical mapping, confirm the presence of Cs2MoO4, which is the major component of the JOG. However, combinatorial analyses reveal that other non-stoichiometric phases are possible, highlighting the complex nature of the crystalline structure of the JOG. Fe is found in metallic Pd-rich precipitates with structure compatible with the tetragonal structure of FePd alloy. Cr is found in different locations of the JOG, in oxide form, but no structural data could be obtained due to local beam sensitization of the sample in those areas.

    更新日期:2020-01-11
  • Synthesis and characterization of iron phosphate based glass-ceramics containing sodium zirconium phosphate phase for nuclear waste immobilization
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-08
    Fu Wang; Jinfeng Liu; Yuanlin Wang; Qilong Liao; Hanzhen Zhu; Li Li; Yongchang Zhu
    更新日期:2020-01-08
  • Thermal conductivity of uranium metal and uranium-zirconium alloys fabricated via powder metallurgy
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-07
    Luis H. Ortega; Brandon Blamer; Karyn M. Stern; James Vollmer; Sean M. McDeavitt

    This study focused on the preparation of metallic uranium and uranium-zirconium alloys to measure the effect significant porosity had on thermal diffusivity from 20∘C to 300∘C. Precursor uranium powders were prepared through a hydride de-hydride process to obtain a < 70 mesh powder. The sample compositions were uranium, uranium 5 mass% zirconium and 10 mass% zirconium. Higher porosity decreased the material's thermal conductivity. Thermal conductivity was also reduced with increased zirconium content.

    更新日期:2020-01-07
  • H diffusion in excel measured by LIBS
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-03
    Rodolfo A. Perez; Carlos Ararat-Ibarguen; Manuel Iribarren

    H bulk diffusion in Excel (Zr–3.5% Sn–0.8% Mo–0.8% Nb) was measured using the Laser Induce Breakdown Spectrometry (LIBS) technique in 469-660 K (196–387 ºC) temperature range for the first time. D temperature dependence obeys the Arrhenius law with activation energy Q = (30 ± 3) kJ/mol and D0 = (3.0 ± 1.0)x10−8 m2/s. Those values are compatible with previous measurements of H diffusion in pure α-Zr and their alloys.

    更新日期:2020-01-04
  • The effect of temperature and fuel surface area on spent nuclear fuel dissolution kinetics under H2 atmosphere
    J. Nucl. Mater. (IF 2.547) Pub Date : 2020-01-01
    Ella Ekeroth; Michael Granfors; Dieter Schild; Kastriot Spahiu

    In this work we present the results of two spent nuclear fuel leaching experiments in simulated granitic groundwater, saturated with hydrogen under various pressures. The results show a large impact of the dissolved hydrogen already at 1 bar H2 and room temperature on the release of both the uranium and of the fission products contained in the fuel matrix. Based on the results of this study and on published data with fuel from the same rod, the importance of the oxidative dissolution of spent fuel under repository conditions as compared to its non-oxidative dissolution is discussed. The XPS-spectra of the fuel surface before the tests and after long-term leaching under hydrogen are reported and compared to reduced UO2 and SIMFUEL surfaces. The overall conclusion is that in spite of the unavoidable air contamination, hydrogen pressures of 1 bar or higher counteract successfully the oxidative dissolution of the spent nuclear fuel. The stability of the 4d-element metallic particles during fuel leaching under such conditions is also discussed, based on data for their dissolution. The metallic particles are also stable under such conditions and are not expected to release their component metals during long-term fuel leaching.

    更新日期:2020-01-01
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