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  • 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
  • Phase reversion kinetics of thermally decomposed (α + γ′) phases to γ-phase in U – 10 wt% Mo alloy
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-31
    Ryan Newell; Abhishek Mehta; Dennis D. Keiser; Yongho Sohn

    Thermally induced phase reversion kinetics of decomposed α + γ′ phases to γ phase was investigated for the U – 10 wt% Mo (U–10Mo) hypoeutectoid alloy using X-ray diffraction (XRD), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). The as-cast U–10Mo alloy was homogenized to the γ phase at 1073K for 96 h, and decomposed at sub-eutectoid temperature of 773K as a function of time from 240 to 1200 h. The decomposition of γ phase initiated via two mechanisms: (1) discontinuous precipitation (DP) or cellular precipitation along the grain boundaries and inclusions; and (2) continuous precipitation (CP) within the γ grains as γ′ needles. After 720 h at 773K, the DP colonies completely consumed the continuous γ matrix, after which discontinuous coarsening (DC) was observed. The completely-decomposed U–10Mo alloy samples were then annealed at 843, 853, and 863K as a function of time to document the kinetics of α+γ' → γ phase reversion. The reference intensity ratio analysis was employed to XRD patterns to quantify the phase amount after each heat treatment. A combinatorial kinetic model, consisting of the classical JMAK nucleation and growth, and spherical particle dissolution, was employed to describe the α+γ' → γ phase reversion as a function of time and temperature. This phenomenological model provided a good approximation for the phase reversion within the temperature and time ranges examined in this study. Furthermore, a diffusion enhancement approach was employed for a set of typical irradiation parameters to estimate the irradiation effect on the α+γ' → γ phase reversion kinetics.

    更新日期:2019-12-31
  • The characterization of electrodeposited chromium barriers for nuclear reactor cladding application
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-30
    Sunghwan Yeo; Jun Hwan Kim; Sung Ho Eom

    A layer of Cr coating on the inner surface of HT9 cladding is a promising barrier to hinder fuel and cladding chemical interaction (FCCI) in sodum-coolded fast reactors (SFRs). Cr barriers obtained with various current densities and electrolyte temperatures used for the electrodeposition of HT9 disks, were investigated for thickness, Vickers hardness, and the existence and length of microcracks. The crack-free Cr barrier electrodeposited at 1.6 A/cm2 and 80 °C for 70 min showed excellent barrier performance in the diffusion couple test with Ce/Nd alloy at 580 °C for 25 h. These conditions were applied to the electrodeposition of HT9 cladding tubes. The Cr barrier with a thickness of 27.24 ± 1.25 μm having sound microstructure was successfully coated on the inner surface of HT9 cladding tubes using a developed electrodeposition system for the tubes. Burst and tensile mechanical tests verified the excellent adhesion and ductile characteristics of the Cr barriers.

    更新日期:2019-12-30
  • Measurement of PuO2 film thickness by electron probe microanalysis (EPMA) calibration curve method
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-28
    J.A. Stanford; N. Teslich; S. Donald; C.K. Saw; R. Gollott; L.N. Dinh

    Plutonium (Pu) surfaces are highly reactive toward oxygen containing species and, therefore, invariably covered with oxides (e.g. PuO2) during transport and handling. The actual thickness of the surface oxide may dictate if a plutonium part is suitable for a certain application. As a result, a cost-effective, quick, non-destructive, yet reliable means to measure the oxide layer thickness formed on Pu samples is desirable. In this study, the cross-sections of a series of room temperature grown oxides on Pu samples were trenched by focus ion beam (FIB) then observed by scanning electron microscopy (SEM) to measure the surface oxide thicknesses, which were then combined with the corresponding oxygen k-ratios provided by electron probe microanalysis (EPMA) to form calibration curves. Oxide thickness measurements for the calibration curves were made on samples within the typical SEM observable range for PuO2 (35–400 nm). The portion of the calibration curve in the thinner oxide region (<35 nm) were approximated via Pouchou and Pichoir's φ(ρz) theory. Two specimens with micrometer-thick PuO2 standards (one formed at room temperature and the other at higher temperature with a higher level of crystallinity) were made for the k-ratios in this study, allowing EPMA users to choose the standard that best suits their needs. If the surface corrosion is known to be PuO2 (from the environment in which the Pu sample is stored) or if the stoichiometry of the surface oxide is confirmed by a preliminary/compliment technique, these calibration curves allow EPMA users to quickly and efficiently determine PuO2 thicknesses from the measured oxygen k-ratios of their samples. The methodology presented in this study can also be used as a template for creating calibration curves for oxides grown on other actinides.

    更新日期:2019-12-29
  • Experimental evaluation of orientation and temperature dependent material stress-strain curves of Zr2.5%Nb Indian pressure tube material and development of a suitable anisotropic material model
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-25
    M.K. Samal; A. Syed; D. Sen; J. Chattopadhyay

    Because of their hexagonal close-packed (HCP) structure and easy prismatic slip, the tensile properties of Zirconium based alloys show strong texture dependence and hence, anisotropy. The type of texture, microstructure and hence, the mechanical properties depend upon the sequence of manufacturing process followed in fabrication of components, such as pressure tubes (PT) in pressurized heavy water type reactors. Due to small wall thickness (i.e., 3.75 mm) of the pressure tubes, designing and machining of specimens in all the three directions, especially in the radial direction, is a challenge and this explains the reason for scarcity of work in this area in research literature. In this work, different types of specimens have been machined from transverse, longitudinal and radial orientations of the Zr2.5%Nb Indian PT following a novel design in order to study the anisotropy in the plastic deformation and hardening properties. In addition, shear specimens have been extracted from longitudinal-transverse plane to study the plastic deformation in shear. Tests on tensile and shear specimens have been carried out at room temperature (25), 100, 200 and 300 °C respectively. Using the experimental data, the parameters of Hill's anisotropic yield function have been evaluated. In addition to evaluating the parameters at initial yield point, the nature of variation of these parameters with equivalent plastic strain, has also been established. This new material model where the anisotropic parameters evolve with plastic strain, shall be useful for design (by analysis) and integrity assessment of the pressure tubes for different types of postulated loading conditions.

    更新日期:2019-12-26
  • Simulation of reactivity initiated accident thermal transients on nuclear fuels with laser remote heating
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-24
    T. Vidal; L. Gallais; J. Faucheux; H. Capdevila; J. Sercombe; Y. Pontillon

    The investigation on the behaviour of nuclear fuels under thermal loads representative of RIA (Reactivity Initiated Accident) conditions in nuclear reactors is a significant challenge. In this work we introduce a new concept of a laboratory experiment relying on high power lasers to reproduce the temperature gradients experienced by nuclear fuel pellets during such events. The concept is experimentally demonstrated on UO2 samples, supported with numerical simulations to estimate the temperature gradients within the samples. This is a first step in the perspective of applying the technique on irradiated fuels.

    更新日期:2019-12-25
  • Conductive inserts to reduce nuclear fuel temperature
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-24
    Pavel G. Medvedev; Robert D. Mariani

    In an effort to develop better performing nuclear fuels, placement of conductive inserts in nuclear fuel pellets is being considered. Conductive inserts incorporated in the fuel pellets will dissipate heat more efficiently, allowing the fuel to operate at a lower temperature while producing the same amount of power. Calculation results reported in the present paper, show that placing a 6-finned molybdenum insert in a UO2 pellet will result in an 842 C reduction of the peak fuel temperature. Likewise, placing multiple equidistantly spaced 50 μm thick molybdenum discs in a UO2 pellet will result in a 995 C reduction of the peak fuel temperature. In both cases, the volume of the insert is limited to 5% of the volume of the pellet. While these results are preliminary in scope, the impact of the reduced peak temperature on expected fuel performance warrants initiation of experimental work that will include fabrication and irradiation of these and similar fuel designs at the Idaho National Laboratory.

    更新日期:2019-12-25
  • 更新日期:2019-12-23
  • A improved equation of state for Xe gas bubbles in γU-Mo fuels
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-20
    Benjamin Beeler; Shenyang Hu; Yongfeng Zhang; Yipeng Gao

    A monolithic fuel design based on a U–Mo alloy has been selected as the fuel type for conversion of the United States High-Performance Research Reactors (HPRRs). An issue with U–Mo monolithic fuel is the large amount of swelling that takes place during operation. The accurate prediction of fuel evolution under irradiation requires implementation of correct thermodynamic properties into mesoscale and continuum level fuel performance modeling codes. However, the thermodynamic properties of the fission gas bubbles (such as the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) are not well known. This work studies Xe bubbles in γU-Mo from a diameter of 3 nm up to 8.5 nm and from 400 K up to 700 K. The energetic relationship of Xe bubbles with regard to voids and Xe substitutional atoms is described. The transition is also determined for when a bubble becomes over-pressurized. Finally, an equation of state is fit to the pressure as a function of molar volume and temperature.

    更新日期:2019-12-20
  • The sporadic history of rubidium and its role in corrosion of steel related to nuclear material storage
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-20
    R. Matthew Asmussen; James J. Neeway

    The influence of rubidium (Rb), both in its metallic and oxidized states, on the corrosion resistance of steels has direct relevance to the storage of radioactive Kr-85 containing nuclear materials. Kr-85 undergoes a β-decay to generate stable, metallic Rb. The literature to date is sparse, contained primarily in company reports and divisive as to whether Rb metal, or its oxides, can lead to corrosion of steel materials. In both the direct storage of Kr-85 and used nuclear fuel that will contain Kr-85, if Rb is corrosive then continual decay of the Kr-85 will generate an increasingly corrosive environment for the steel canister it is stored in. This review was written to consolidate the known data to date on Rb corrosion of steels to identify consistencies, contradictions and remaining gaps in our understanding of the Rb: steel system.

    更新日期:2019-12-20
  • Efficient deuterium permeation reduction coating formed by oxidizing the Fe–Cr–Al ferritic steel in reduced oxygen atmosphere at 973 K
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-20
    Yi-Ming Lyu; Yu-Ping Xu; Xin-Dong Pan; Hao-Dong Liu; Xiao-Chun Li; Hai-Shan Zhou; Zhong-Shi Yang; Guang-Nan Luo

    Alumina is regarded as one of the most promising candidate tritium permeation barrier (TPB).Through thermal oxidization of Al-contained alloys Fe–Cr–Al, alumina layer with high tritium permeation reduction ability can be obtained. After bonding this kind of materials with structural materials such as reduced activation ferritic/martensitic (RAFM) steels, it can serve as TPB. In this work, efforts have been done to enhance the hydrogen isotope permeation reduction ability of the alumina layer on the Fe–Cr–Al ferritic steel by optimizing the oxidation process. The oxidation temperature of the Fe–Cr–Al ferritic steel is set to 973 K, which is lower than the final heat treatment temperature of the RAFM steel. Three different atmospheres have been employed for the oxidation process of the Fe–Cr–Al ferritic steel. Gas driven permeation (GDP) experiments have been performed to examine the deuterium permeability of the oxidized Fe–Cr–Al ferritic steel. The deuterium permeability of the Fe–Cr–Al ferritic steel oxidized in argon with 1700 ppm oxygen is 104 times lower than that of the RAFM steel at 823 K. The microstructure and chemical composition of the oxide layer of the Fe–Cr–Al ferritic steel oxidized in three different atmospheres has been clarified using Scanning Electron Microscopy (SEM) and X-ray Photoelectron Spectroscopy (XPS) methods. It is suggested that oxidizing the Fe–Cr–Al ferritic steel at reduced oxygen atmosphere could inhibit the growth of the iron oxide and chromium oxide in the oxide layer, thereby leading to a dense and compact alumina layer that has excellent hydrogen permeation reduction performance.

    更新日期:2019-12-20
  • Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON1
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-20
    D. Schappel; N.R. Brown; K.A. Terrani

    This research presents BISON simulations of stress, temperature, and failure probability of tristructural isotropic (TRISO) fuel particles subjected to transient power pulse conditions in the Nuclear Safety Research Reactor (NSRR). By modifying the default elastic properties of the PyC and the UO2 coefficient of thermal expansion correlation, BISON was found to produce suitable agreement with the observed and independently simulated results when appropriate shape and scale Weibull parameters for SiC failure were chosen. The Weibull parameter effects were also explored within the range of SiC failure and were applied using experimental data from hemispherical crush testing.

    更新日期:2019-12-20
  • Utilization of Artificial Neural Network to explore the compositional space of hollandite-structured materials for radionuclide Cs incorporation
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-18
    Dipta B. Ghosh; Bijaya B. Karki; Jianwei Wang

    Hollandite with the general formula A2B8O16 is known for its potential to immobilize radionuclide Cs in the tunnel along the z-axis of the crystal structure. The effective Cs incorporation in a hollandite phase with an optimal loading capacity and the long term stability depends significantly on the B-site cations, which, in addition to providing optimal structural compatibility, must ensure the phase's resistance to chemical weathering in an aqueous environment that includes external thermodynamic conditions such as temperature and solution chemistry. Based on the importance of the B-site cations, we explored in detail the possible B-site compositions by employing Artificial Neural Network (ANN) simulations and crystal chemistry principles. With a set of 91 experimentally determined data collected on hollandite that is available in open literature, we trained the network and subsequently tested the predictive power of the trained network. Relying on the successful outcomes of the trained network at the testing phase, we further utilized the trained network to map the dependence of the tunnel size, which was used as a criterion for Cs compatibility in the channel, in a wide compositional space encompassing eighteen 3 + cations and fifteen 4 + cations. By combining the Cs compatibility and the structural tolerance factor for hollandite structure, the predicted B-site compositions, comprising of cations spanning across the depth and breadth of the periodic table, can be employed as a guide in the search for optimal hollandite composition for Cs immobilization.

    更新日期:2019-12-19
  • An equation of state for xenon/krypton mixtures confined in the nuclear fuels
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-17
    A. Jelea

    In the present paper, one proposes an equation of state allowing to reasonably describe the behavior of the xenon/krypton mixtures, in any proportion, confined in the form of nanoscale bubbles in the UO2 and MOX nuclear fuels. The temperature range covered by this equation lies between the room temperature and the fuel melting point. The equation of state, which was built based on molecular dynamics results obtained on a large density-composition-temperature range, could be easily implemented in the nuclear fuel performance codes.

    更新日期:2019-12-18
  • Effects of plutonium dioxide encapsulation on the physico-chemical development of Portland cement blended grouts
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-17
    Sarah A. Kearney; Bliss McLuckie; Kevin Webb; Robin Orr; Ian A. Vatter; Antonia S. Yorkshire; Claire L. Corkhill; Martin Hayes; Michael J. Angus; John L. Provis
    更新日期:2019-12-18
  • Assessment of phase stability of oxide particles in different types of 15Cr-ODS ferritic steels under 6.4 MeV Fe ion irradiation at 200 °C
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-16
    Peng Song; Akihiko Kimura; Kiyohiro Yabuuchi; Peng Dou; Hideo Watanabe; Jin Gao; Yen-Jui Huang
    更新日期:2019-12-17
  • Oxidation and anion lattice defect signatures of hypostoichiometric lanthanide-doped UO2
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-16
    Travis A. Olds; Samuel E. Karcher; Kyle W. Kriegsman; Xiaofeng Guo; John S. McCloy

    A series of sintered UO2 pellets doped with lanthanide (Ce, Nd, Yb) elements were investigated using powder X-ray diffraction, Raman spectroscopy, thermogravimetric analyses and differential scanning calorimetry. A combination of electron microprobe and thermogravimetric analyses, for oxygen content, enabled precise determination of the hypostoichiometry for lanthanide-doped samples at 1 and 5 atom percent. Two Raman laser wavelengths (785 and 455 nm) have afforded greater sensitivity to spectroscopic signatures of the phonon bands (1LO and 2LO) associated with oxidation of (U1-yMy)O2-x and the anion defects introduced by lanthanide substitution. Oxygen hypostoichiometry forces a reduction in the average coordination number surrounding (U,M) sites, which is compensated by a decrease in U–O bond length, and concomitantly the lattice parameter, consistent with the obtained Raman spectra. The evolution of O/M ratio up to (U1-yMy)O2 after oxidation was also examined using Raman spectroscopy, revealing that the ‘defect band’, including a component attributed to oxygen vacancies (∼540 cm−1) and the 1LO phonon (∼575 cm−1) increased in intensity with increasing dopant concentration and upon oxidation. The lanthanide dopants inhibited oxidation to U3O8, most prominently for Yb 5 at%, having been delayed by ∼180 °C. Thermogravimetric analyses reveal an early oxidation feature that may be related to influx of O to satisfy hypostoichiometry up to (U1-yMy)O2, possibly stabilizing a U4O9 or U3O7 intermediate, delaying oxidation to U3O8.

    更新日期:2019-12-17
  • Fission product speciation in the VERDON-3 and VERDON-4 MOX fuels samples
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-16
    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

    Within the framework of the International Source Term Programme, the VERDON-3 and VERDON-4 complementary tests were devoted to studying MOX fuel and fission product (FP) behaviour at very high temperature (>2300 °C) under oxidising and reducing atmospheres. Post-test qualitative and quantitative characterisations of the VERDON-3 and VERDON-4 samples were carried out in the LECA-STAR facility at the CEA Cadarache centre. This article focuses on the impact of the atmosphere on fission product speciation and more precisely on the behaviour of Cs, Mo and Ba. The well-known white inclusions composed of metallic Mo, Ru, Rh, Pd and Tc underwent transformations during the two tests implying melting and/or composition changes. The semi-volatile behaviour of Mo was once again confirmed. Almost no Cs was seen to remain in the samples after the VERDON-3 and VERDON-4 tests, which is consistent with its volatile behaviour. The behaviour of Ba, which is known to be released in high quantities in reducing atmospheres, provided complex to assess as significant releases seems to have taken place in both VERDON-3 and VERDON-4 tests.

    更新日期:2019-12-17
  • Regulating the helium bubble nucleation in the titanium tritides by environment temperature during the early aging period
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-16
    G.J. Cheng; B. Yao; W. Ding; L.Q. Shi; X.S. Zhou; X.G. Long; M. Chen; H.H. Shen; S.M. Peng

    The nucleation of helium bubbles in the early aging stage of tritium storage materials will have an important influence on the bubble growth behavior in the whole of aging. Herein, the nucleation of helium bubbles was adjusted by environment temperature at the first 15 days immediately after preparation of polycrystalline titanium tritide films. The influence of the environment temperature on the nucleation of helium bubbles were studied by X-ray diffraction (XRD), positron annihilation spectroscopy (PAS), thermal helium desorption spectroscopy (THDS) and scanning transmission electron microscopy (STEM). Results demonstrate that the helium bubble nucleation is significantly affected by the environment temperature of −60 °C, 20 °C and 120 °C. The higher the environment temperature, the lower number density of helium bubble in the titanium tritide films. The helium bubble nucleation mechanism is strongly correlated to the helium atom diffusion coefficient that is controlled by the environment temperature. During the following aging, the rate of helium atoms captured by a single helium bubble increases in titanium tritides with low number density bubbles, which accelerates the growth of helium bubbles in the matrix. However, the low number density bubbles also decrease the capture probability of free helium atoms by helium bubbles, resulting in the increase of helium atoms diffusing to infinite size defects such as grain boundaries macroscopically, and accelerating the evolution of helium in titanium tritides.

    更新日期:2019-12-17
  • Understanding the role of Fe, Cr and Ni in Zircaloy-2 with special focus on the role of Ni on hydrogen pickup
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-14
    Y.R. Than; R.W. Grimes; B.D.C. Bell; M.R. Wenman

    Ni as an alloying addition in Zircaloy leads to an increase in hydrogen pick-up fraction. Atomic scale simulations of tetragonal ZrO2, based on density functional theory, are used to identify a possible mechanism for this observation. First, defect formation energies associated with Ni but also Fe and Cr are used to predict relative defect cluster and defect charge concentrations using Brouwer diagrams. At low oxygen partial pressures (PO2), expected in the vicinity of the oxide metal interface, a cluster consisting of an oxygen vacancy adjacent to a charge neutral Ni0 atom is identified as the most populous cluster. Further simulations show that a hydrogen molecule will dissociate in the vicinity of this cluster. No other cluster is both sufficiently populous and acts in this way. This differentiates Ni from the other alloying elements.

    更新日期:2019-12-17
  • Behaviors of bubble-loop complexes in He-irradiated CLAM steels at elevated temperatures
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-13
    Fang Li; Yaxia Wei; Fengfeng Luo; Weiping Zhang; Xiong Zhou; Yiheng Chen; Cheng Chen; Liping Guo; Jingping Xin; Shaobo Mo

    The behavior of bubble-loop complexes formed in reduced activation ferritic/martensitic (RAFM) steels under He+ irradiation at elevated temperatures was investigated by transmission electron microscope (TEM), and corresponding simulations on the interaction between He bubbles and dislocation loops were also carried out by molecular dynamics (MD). China Low Activation Martensitic (CLAM) steels were irradiated under18 keV He+ with the fluence of 1 × 1020 ion/m2 at room temperature, 250 °C, 350 °C, 400 °C, 450 °C and 550 °C, respectively. Bubble-loop complexes were observed in all specimens when irradiated at 250 °C and above. Almost all of He bubbles were located inside the dislocation loops when the irradiation temperature reached up to 550 °C. Moreover, the formation of the complexes was independent of the type of dislocation loops since both a0 <100> and a0/2 <111> loops could form bubble-loop complexes. He bubbles were situated to be more stable inside the dislocation loops than that in the matrix according to the experimental results. The MD simulations showed that the dislocation loop can be locked by the He bubble, which could elucidate the mechanism of the observed bubble-loop complexes formed in He irradiated RAFM steels. The mean size of dislocation loops decreased when irradiation temperature increased to 550 °C, which can be ascribed to the high emission rate of point defects from dislocation loops at high temperature.

    更新日期:2019-12-13
  • Computational study of HCl adsorption on stoichiometric and oxygen vacancy PuO2 {111}, {110} and {100} surfaces
    J. Nucl. Mater. (IF 2.547) Pub Date : 2019-12-12
    Jonathan Collard; Helen Steele; Nikolas Kaltsoyannis

    The interactions between HCl and both the pristine and defect {111}, {110} and {100} surfaces of PuO2 are modelled using hybrid density functional theory, within the periodic electrostatic embedded cluster method. In the case of the pristine surfaces, adsorptions onto the {110} surface are the most stable, likely due in part to the presence of surface hydrogen bonds. In addition, results suggest that even when dissociatively adsorbed onto the surface, the proximity of the hydrogen and chlorine atoms has a significant impact on the stability of the system. The electronic structure of both the pristine and reduced surfaces of PuO2 and UO2 is also probed, with unpaired electrons left behind in a neutral oxygen vacancy defect site having a greater tendency to delocalize on the surface of UO2 than PuO2. HCl adsorptions on the reduced surfaces reveal that configurations in which the chlorine atom attempts to “heal” the gap left by the oxygen vacancy in the {111} surface are by far the most stable of all considered. Finally, molecular thermodynamics is employed to translate adsorption energies to HCl thermal desorption temperatures, for each geometry considered. Particularly when a chlorine atom is embedded in the surface of PuO2, the temperatures required for thermal desorption to occur are high, implying that although thermal treatment is likely to remove some chlorine contamination from PuO2 samples, some will likely remain bound to defect sites within the material.

    更新日期:2019-12-13
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