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  • Phase transition and compressibility study of UOs2 under pressure
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-18
    Balmukund Shukla, G. Shwetha, N.R. Sanjay Kumar, N.V. Chandra Shekar

    High-pressure X-ray diffraction studies on MgCu2 type phase of UOs2 have been carried out up to 36 GPa. The compound remains in its parent phase up to 12 GPa with bulk modulus 261 GPa. Beyond 12 GPa, a first order phase transformation is observed. The phase transition is sluggish in nature and completes at 20.1 GPa. The high-pressure phase is found to be a hexagonal cell with lattice parameters a = 3.013 Å and c = 4.267 Å. Charge density calculations show that uranium tetrahedra in the lattice are responsible for the phase transition to occur. The density of state plots at Fermi level, wherein a pseudogap originates at high pressures, confirms the existence of a high pressure phase and uranium site is found to contribute significantly to such changes in density of state. The retrievable high pressure phase is found to be the least compressible among uranium intermetallic compounds with its bulk modulus of 366 GPa.

    更新日期:2018-09-19
  • Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-18
    Chongchong Tang, Adrian Jianu, Martin Steinbrueck, Mirco Grosse, Alfons Weisenburger, Hans Juergen Seifert

    FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500 °C was examined. In case of transient ramp tests, catastrophic oxidation, i.e. rapid and complete consumption of the alloy, occurred during temperature ramp up to above 1200 °C for specific alloys. The maximum compatible temperature of FeCrAl alloys in steam increases with raising Cr and Al content, decreasing heating rates during ramp period and doping of yttrium. Isothermal oxidation resulted in catastrophic oxidation at 1400 °C for all examined alloys. However, formation of a protective alumina scale at 1500 °C was ascertained despite partial melting. The occurrence of catastrophic oxidation seems to be controlled by dynamic competitive mechanisms between mass transfer of Al from the substrate and transport of oxidizing gas through the scale both toward the metal/oxide scale interface.

    更新日期:2018-09-19
  • A first principle calculation on electronic properties of plutonium mononitride: Insights from dynamical mean field theory
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-18
    Ru-song Li, Ning-hua Tong, Jin-tao Wang, Du-qiang Xin, Shi-qi Huang

    We perform a first principles calculation on electronic properties of plutonium mononitride (PuN) using a many-body method merging local density approximation (LDA) with dynamical mean field theory (the so called LDA + DMFT scheme), taking into account the on-site Coulomb interaction between Pu 5f states and spin-orbit coupling due to high atomic number of plutonium element. We find that PuN is a mixed-valentnf = 4.823 moderately correlated system and a triplet of quasiparticle peaks below the Fermi level corresponding to multiplet of many-body quasiparticle peaks are due to valence fluctuations and the Pu atomic multiplet structure, which is agreement with the photo-electron spectroscopy observation. The calculation result reveals that the low energy scattering rate from the imaginary part of the self energy is very large, indicating that PuN is a bad metal, which is also in agreement with the density of states (DOS) analysis and other theoretical calculation. In order to compare with experimental angle-resolved photoemission spectrum (ARPES), we also calculate momentum-resolved electronic spectrum function, and analyze electronic excitation across the Fermi level.

    更新日期:2018-09-18
  • A hardening model for the cross-sectional nanoindentation of ion-irradiated materials
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-17
    Xiazi Xiao, Long Yu

    In this paper, a hardening model is developed for the cross-sectional nanoindentation of ion-irradiated materials. The model is based on the derivation of an average defect density within the formed plasticity affected region, which depends on the distribution of defect density in the irradiated region, indentation depth and distance from the irradiated sample surface. A succinct parameter calibration process is proposed by comparing the theoretical results with experimental data at a given indentation depth. A good agreement with experimental data can be achieved for both the fitted relationship between irradiation hardening and indentation distance from the irradiated sample surface under cross-sectional nanoindentation, and predicted hardness as a function of indentation depth under surface nanoindentation. Therefore, the rationality and accuracy of the proposed model are effectively verified. Based on the analysis of this proposed model, it is available to characterize the property of plasticity affected region and irradiation depth of ion-irradiated materials based on the experimental data measured through cross-sectional nanoindentation.

    更新日期:2018-09-18
  • Evolution of defects in Ti6-4 under Ti2+ ion irradiation: Focus on radiation-induced precipitates
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-18
    Sylvie Doriot, Emilie Jouanny, Joël Malaplate, France Dalle, Lucien Allais, Thierry Millot, Marion Descoins, Dominique Mangelinck, Moukrane Dehmas

    Ion irradiations on the Ti6-4 titanium alloy were conducted at the JANNUS French platform in two different conditions of temperatures, doses and fluxes, to simulate neutron irradiation damage. Quantification of defects and chemical microanalyses were carried out thanks to Transmission Electron Microscopy and Atom Probe Tomography. <a>-type loops and radiation induced precipitates (a vanadium-rich β BCC phase) were observed for both irradiation temperatures. During an irradiation at 300 °C, there was no notable influence of the dose and flux for the considered doses and fluxes ranges on the <a>-type defects. The influence of raising the irradiation temperature up to 430 °C was a lowering of their density and an increase of their mean diameter for both defects. In addition, a lower flux seemed to enhance this temperature effect. These phenomena were very significant for precipitates whereas it appeared very modest for <a>-type loops. The probable mechanism to explain the distribution of vanadium-rich β precipitates inside the α phase is the heterogeneous nucleation. The nucleation is dominated by the Radiation Induced Segregation (RIS) phenomenon at 430 °C and could be dominated by the mechanism of vanadium-rich clusters formation by ballistic effects in the cascades at 300 °C.

    更新日期:2018-09-18
  • Influence of formulation parameters of cement based materials towards gas production under gamma irradiation
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-14
    D. Chartier, J. Sanchez-Canet, L. Bessette, S. Esnouf, J.-P. Renault

    The release of radiolysis gas is a concern that may restrict the use of cement materials to condition intermediate level radioactive waste. Indeed, water naturally present in cement materials produces hydrogen gas (which can be explosive/flammable under some conditions) when it is exposed to ionizing radiation.The primary goal of the MATRICE (MAterials Resistant to Irradiation based on Cement) project is to identify and define formulations of cement materials in order to minimize the quantities of hydrogen gas released by radiolysis. The first approach is the minimization of water amount in standard Portland materials (calcium silicate-based cements) by addition of specific compounds (superplasticizers) to enable the preparation of wasteform. The second approach is to use “alternative” cement such as calcium sulfoaluminate cement. This cement was expected to release less hydrogen because the quantity of water needed for cement hydration is higher than Portland and moreover, their hydrates differ from those of hydrated calcium silicate mostly encountered in Portland based materials.Based on gamma irradiations with a60Co source, the results obtained demonstrate that the first approach is efficient but yet limited because the production of hydrogen of Portland pastes is about proportional to the total amount of water present in the materials. Thus, a tremendous drop of hydrogen production cannot be reach because rheological constraint does not allow a huge reduction of water, even with efficient superplasticizers. The second approach using calcium sulfoaluminate cements as an alternative binder provides results that are quite similar to Portland cement concerning the production of hydrogen under gamma irradiation.

    更新日期:2018-09-15
  • Molecular dynamics study of hydrogen-vacancy interactions in α-zirconium
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-14
    Christopher I. Maxwell, Edmanuel Torres, Jeremy Pencer

    Hydrogen is known to have detrimental effects on the mechanical properties of zirconium and its alloys, mainly due the formation of brittle hydride phases. The behaviour of hydrogen dissolved in zirconium is not fully understood at the atomic scale. In particular, the influence of hydrogen on the behaviour of defects in zirconium still needs to be clarified. In this paper, we report a molecular dynamics study of hydrogen behaviour in zirconium using an improved Zr-H interatomic potential. We determined the diffusivity of hydrogen in zirconium, and the interaction of hydrogen atoms with a single or a pair of atomic vacancies. The computed hydrogen diffusion is found to be in good agreement with experimentally determined values. Our results also indicate that hydrogen interactions with vacancies in zirconium promote vacancy clustering. Moreover, hydrogen is found to substantially decrease the mobility of divacancies. From the present simulation findings, it is concluded that the presence and concentration of hydrogen impurities may have a significant influence on the evolution of irradiation-induced defects in zirconium alloys.

    更新日期:2018-09-15
  • Fabrication process study of UCO composite ceramic microspheres with fructose as a carbon source by internal gelation and carbothermic reduction
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-14
    Xi Sun, Jingtao Ma, Xingyu Zhao, Shaochang Hao, Taowei Wang, Ziqiang Li, Changsheng Deng, Bing Liu

    A preliminary study was carried out on preparation of UC-UO2 (UCO) composite ceramic microspheres using fructose as the carbon source by combining internal gelation and carbothermic reduction. The effects of the pH of precursor solution, washing and drying processes on the gel microspheres and sintering process on the UCO sintered microspheres were analyzed. The physical properties, composition, surface morphology and internal microstructure of the UCO ceramic microspheres were also studied. Dense and crack-free UCO ceramic microspheres with good sphericity, uniform composition distribution, and high crush strength were successfully fabricated.

    更新日期:2018-09-15
  • Modeling primary creep for Zircaloy claddings during load reversals and drops in BISON
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-15
    Mudasar Zahoor

    It is desired to accurately predict creep strains in zirconium claddings, subjected to cases such as sudden load drops or reversals, using the BISON fuel performance code. It has been experimentally proven that primary creep reinitializes at sudden load drops and reversals, which is different than normal operating conditions, where stresses increase gradually. A new visco-elastic primary creep model is added to BISON, which is capable of handling situations of load drops or reversals. The model is coupled with secondary creep models namely Hayes-Hoppe and Limback-Anderson, available in BISON to make predictions on cladding creep strains. The results are compared to Halden in-pile experiments IFA-585 and IFA-699. Based on the comparison, it is recommended to use the visco-elastic primary creep model coupled with Hayes-Hoppe secondary model for consistency and performance.

    更新日期:2018-09-15
  • Micropillar compression study on heavy ion irradiated Zr-2.5Nb pressure tube alloy
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-13
    Qiang Wang, Chris Cochrane, Fei Long, Hongbing Yu, Mark R. Daymond

    In this paper, 40 MeV Zr ion irradiation on Zr-2.5Nb pressure tube alloy was carried out, delivering a damage layer 6.5 μm deep. Uniaxial micropillar compression experiments along the tube axial (AD) and transverse directions (TD) were conducted. Irradiation was found to alter the deformation mechanisms and to increase the yield strength more in the AD oriented pillars than TD for irradiation up to 0.6 dpa. This anisotropy in response is discussed in terms of the limitations of micropillar compression method and the different deformation mechanisms which are activated in each orientation. The anisotropic hardening effect of <a>-type dislocation loops on prismatic <a>, basal <a>, and pyramidal <c+a> slip systems is also discussed. The micropillar compression results were also compared with nanoindentation results conducted on the same material.

    更新日期:2018-09-14
  • Investigation of ion irradiation hardening behaviors of tempered and long-term thermal aged T92 steel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-12
    Dandan Zhao, Shilei Li, Yanli Wang, Fang Liu, Xitao Wang

    9Cr ferritic/martensitic steels are promising materials for in-core components in advanced Gen-IV reactors. In these applications, their long-term microstructural stability under thermal exposure and resistance to neutron irradiation are essential. Tempered (unaged) and long-term thermal aged T92 samples were used to evaluate the effects of thermal aging and ion irradiation on the microstructure and micromechanical properties of the steel. Both the tempered and aged samples were irradiated with 3 MeV Fe11+ ions to 0.25, 0.50, 1.00 and 5.00 dpa at room temperature. Using the nanoindentation technique, the irradiation hardening behaviors of T92 steel were investigated. The irradiation hardening effect was observed in both the tempered and aged T92 samples. To eliminate the soft substrate effect, the critical indentation depth was determined using the ratio of the average hardness of irradiated and unirradiated samples at the same depth. Under the same irradiation conditions, the macroscopic hardness values of the aged T92 samples after irradiation were lower than those of the tempered samples. The irradiation hardening effect was more significant in the aged T92 due to the decreased dislocation density and the coarsened martensitic lath after long-term thermal aging.

    更新日期:2018-09-13
  • Highly enhanced reduction of rare earth oxides in simulated oxide fuel in Li2OLiCl salt using lithium metal
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-12
    Eun-Young Choi, Jeong Lee

    The reducibility of rare earth oxides (REOs) in electrolytic reduction of nuclear used oxide fuel is poor as a result of their high oxygen affinity and thermodynamic stability. We show that the reduction of REOs to rare earth metals in Li2OLiCl salt can be enhanced significantly by using lithium metal. Specifically, REOs in the fuel are reduced to a high extent in the electrolytic reduction of simulated oxide fuel in 1.0-wt% Li2OLiCl salt following the application of electrical charge in order to induce the production of Li metal. We also demonstrate that the low reduction levels of REOs in reduced simulated fuel with high uranium metal content, which is produced through another electrolytic reduction run, can be increased by the immersion of the reduced simulated fuel in fresh LiCl salt containing Li metal at 0.3 wt%.

    更新日期:2018-09-13
  • Thermodynamic modelling of thoria-urania and thoria-plutonia fuels: Description of the Th-U-Pu-O quaternary system
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-11
    A. Bergeron, D. Manara, O. Beneš, R. Eloirdi, M.H.A. Piro, E.C. Corcoran

    In this work, a thermodynamic treatment for (Th,U)O2 and (Th,Pu)O2 fuels has been developed using the CALPHAD method to describe phase diagram and thermodynamic data of the phases in the Th-U-Pu-O sub-systems. The treatment builds on previously reported assessments of the U-Pu-O ternary oxide system and of the Th-U-Pu ternary metallic system. The thermodynamic and phase diagram properties calculated by the models herein are in good agreement with available experimental data for the binary and ternary subsystems, as well as new phase diagram data for the Th-O binary reported here. This treatment is useful to predict thermodynamic behavior of thoria-based nuclear fuels at various temperatures and compositions.

    更新日期:2018-09-11
  • Hydrogen diffusive transport parameters through CLAM steel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-11
    Hao Yang, Wei Wang, Man Jiang, Xiang Ji, Mingjie Zheng

    China Low Activation Martensitic (CLAM) steel has been selected as the primary candidate structural material for Chinese Helium Cooled Ceramic Breeder ITER Test Blanket Module. In the present work, the hydrogen transport parameters in CLAM steel have been experimentally measured with gas evolution permeation technique in the temperature range of 573–823 K at hydrogen pressures of 102–104 Pa. The pressure exponent n of permeation was between 0.48 and 0.55 and it showed that the permeation was approximately pure diffusion limited. The hydrogen transport parameters of CLAM were presented based on experimental results at 104 Pa, and the transport parameters were Φ(mol·m−1·Pa−1/2·s−1) = 3.11 × 10−8exp(-38.712/RT), D(m2·s−1) = 1.19 × 10−7exp(-16.404/RT), Ks(mol·m−3·Pa−1/2) = 0.27exp(-22.308/RT). All those transport parameters were compared with the available data corresponding to several other RAFM steels and the results shown the similar hydrogen transport behavior of the RAFMs.

    更新日期:2018-09-11
  • Characterisation of the spatial variability of material properties of Gilsocarbon and NBG-18 using random fields
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-08
    José David Arregui-Mena, Philip D. Edmondson, Lee Margetts, D.V. Griffiths, William E. Windes, Mark Carroll, Paul M. Mummery
    更新日期:2018-09-09
  • The study of hardening evaluation of pure Zr with δ-hydrides generation by the dynamic in-situ metallic structure observation and nano-indentation hardness test
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-08
    Kouichi Tougou, Takashi Onitsuka, Ken-ichi Fukumoto, Masayoshi Uno, Hiroaki Muta

    The in-situ TEM observations during tensile tests, electron backscatter diffraction (EBSD) analyses and nano-indentation hardness tests were performed to investigate the hardening mechanism by the relationship between the slip planes of dislocations and the habit planes of δ-hydrides in the pure Zr. The δ-hydride of habit plane which has crystal orientation relationship of (0001)α plane//{111}δ plane and <211¯0>α direction//<110>δ direction, was generated in pure Zr sample. The α-Zr phase was hardened on the crystal orientation of mainly (0001)α plane slip of dislocation. The hardening depends on crystal orientation and it was influenced by the interaction area (sheared area) of dislocations and δ-hydrides. It suggests that the hardening depends on the deference between the slip plane of dislocation and the habit plane of δ-hydride. Also, the hardening of pure Zr was influenced by the δ-hydride but not by the solid solution hydrogen. It is considered that the failure and embrittlement with the δ-hydrides generation in pure Zr is more influenced by the orientation of the long axis direction of δ-hydride than the interaction between the dislocation and δ-hydride in the α-Zr phase.

    更新日期:2018-09-09
  • STEM-EDS/EELS and APT characterization of ZrN coatings on UMo fuel kernels
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-08
    L. He, M. Bachhav, D.D. Keiser, J.W. Madden, E. Perez, B.D. Miller, J. Gan, W. Van Renterghem, A. Leenaers, S. Van den Berghe

    In the framework of the SELENIUM project, ZrN coated U-Mo fuel kernels were irradiated in the Belgium Reactor 2 of SCK•CEN to a plate average burnup of 48% 235U and a local maximum burnup just below 70% 235U. The microstructure and chemical composition of ZrN coating before and after neutron irradiation have been analysed using scanning transmission electron microscopy (STEM) equipped with energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS), and atom probe tomography (APT). The atomic ratio of N/Zr and fission product distribution determined by three techniques were compared and the combination of three techniques shows the advantages in characterization of chemical information for nuclear materials.

    更新日期:2018-09-09
  • Role of compression metallization in UO2 fission-product energy cascade track: Multiscale electron-phonon analyses
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-07
    Woong Kee Kim, Corey Melnick, Ji Hoon Shim, Massoud Kaviany

    While the electronic stoppage of charged fission fragments is relatively well understood, the subsequent energy cascade is not. Recent efforts to investigate this cascade and predict the resulting damage have used a two-temperature model (TTM) of the electronic and phononic systems coupled with a classical molecular dynamics (MD) simulation of the crystal lattice. In order to accurately predict the track radius produced by a fission fragment in UO2, this model (TTM + MD) requires that UO2, an insulator, have metallic properties, e.g., a substantial electron thermal conductivity and heat capacity. However, it has been predicted that UO2 becomes metallic under large pressures, and we perform ab initio (DFT-HSE) simulations to support this prediction. We show that the average U-U bond length decreases within and near the ion track during TTM + MD simulations, supporting the use of volume contraction to model the pressurized UO2 cell. Additionally, we evaluate the electron, phonon, and electron-phonon coupling properties of UO2 for variations in the pressure. In particular, we calculate the electronic heat capacity and thermal conductivity, and the electron-phonon energy coupling for use in subsequent TTM + MD simulations. The ab initio parameterized TTM + MD simulations provide a set of the track radii predictions which bracket and include the experimentally observed radii. The accuracy of the ab initio parameterized TTM + MD simulations depends on the pressure and degree of electron-phonon non-equilibrium assumed during the ab initio calculations. We suggest improvements to the current TTM + MD methodology in light of these results. Still, we show that the pressure-induced transition of UO2 from insulator to metal and subsequent energy transfer from the electronic to phononic systems can accurately explain radiation damage during swift, heavy ion stoppage in UO2. We make some additional observations regarding the accumulation and recombination of damage along the ion track and make comparison to the common SRIM model of ion stoppage and damage accumulation.

    更新日期:2018-09-07
  • Measurement of tensile strength of nuclear graphite based on ring compression test
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-07
    Xiaojuan Zhang, Yanan Yi, Haibin Zhu, Guangyan Liu, Libin Sun, Li Shi, Han Jiang, Shaopeng Ma

    Nuclear graphite is a type of quasi-brittle material, in which the ratio of the tensile strength to the compressive strength is higher than that of ceramic and hard rock materials. In this case, it is difficult to achieve a perfect splitting mode when the Brazilian splitting test is used to measure the tensile strength of such materials, leading to significant errors in the measurement. In this study, a ring compression test for measuring the tensile strength of the nuclear graphite material was proposed, and related principles and experimental verification were presented. The results showed that a regular tensile failure mode can be achieved using ring compression tests, and the tensile strength of nuclear graphite can be accurately measured. Furthermore, it was found that the accuracy obtained from the ring compression test was much better than that of the Brazilian splitting test.

    更新日期:2018-09-07
  • Ion-irradiation hardening accompanied by irradiation-induced dissolution of oxides in FeCr(Y, Ti)-ODS ferritic steel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-07
    Peng Song, Jin Gao, Kiyohiro Yabuuchi, Akihiko Kimura

    Irradiation effects on hardness and phase stability were investigated for an FeCr(Y, Ti)-ODS ferritic steel strengthened by Y-Ti-O nano-particles after irradiation with 6.4 MeV Fe3+ at room temperature (RT) up to nominal damages of 2, 10 and 50 dpa. With increasing displacement damage up to ∼20 dpa, nano-sized oxide particles slightly shrank, while the corresponding number density drastically decreased by almost two orders of magnitude compared to that of before irradiation. It is considered that ballistic dissolution should be responsible for such reductions in the particle size and number density. Dislocation loops consisting of 1/2<111> type (>80%) and <100> type were observed under weak beam dark field (WBDF) condition in the specimen irradiated to the nominal damage of 50 dpa. The average size and number density of all the dislocation loops were 2.8 ± 0.7 nm and (4.1 ± 0.7) × 1022 m−3, respectively, at the local damage of ∼72 dpa. Although the oxide particles were almost completely dissolved, nanoindentation hardness measurements revealed that the hardening went up continuously with increasing displacement damage and estimated to be 1.63 ± 0.39 GPa by the Nix-Gao model at the nominal damage of 50 dpa. The irradiation hardening accompanied by the dissolution of oxide particles was interpreted in terms of loss of oxide particles, solid solution hardening and formation of fine dislocation loops. The contribution of dislocation loops observed by TEM to the hardening was insufficient to overcome the loss of strengthening by dissolution, suggesting the importance of solid solution hardening and the larger strength factor of dislocation loops as a hardening contributor.

    更新日期:2018-09-07
  • Modeling and predicting total hydrogen adsorption in nanoporous carbon materials for advanced nuclear systems
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-07
    Stephen T. Lam, Ronald Ballinger, Charles Forsberg

    The retention of hydrogen in a variety of disordered carbon materials including graphene nanoplatelets and amorphous carbon was measured by volumetric adsorption and modeled using thermodynamic methods. To date, high temperature hydrogen adsorption research in carbon has predominately focused on graphite used as plasma-facing materials in fusion systems, moderators in high temperature gas reactors, and fuel components in molten-salt cooled reactors. Tritium, a hydrogen isotope that is produced in these systems, adsorbs on carbon surfaces that could be used effectively as sinks to remove tritium. Since hydrogen has chemical similarity to tritium, it can be used as a surrogate to simplify experimental studies, greatly expediting materials exploration. Thus, an understanding and prediction of hydrogen behavior on carbon materials is essential for evaluation of performance and safety of present and future reactor designs. Chemisorption experiments were conducted on 9 different materials with a range of surface areas, pore volumes, and pore size distributions. The results showed that carbons with higher surface area and pore volume exhibit higher hydrogen adsorption, which is attributed to larger quantities of active sites. Further, hydrogen adsorption was found to increase in materials with greater micropore (<10 Å) surface area, which was interpreted as being a result of in-pore trapping. In contrast to previous studies that assumed a single isotherm, four isotherms were examined including two- and three-parameter methods and homo- and heterogeneous surface methods—the Langmuir, Temkin, Freundlich and Sips isotherms. Analyses of non-linear models by ordinary least squares and Marquardt's percent standard deviation minimization were performed. It was found that constant order homogeneous models that have been used in the past lack the ability to describe pathways and reactions steps that occur. Further, it was found that Sips isotherm consistently provided the best minimization of error and most accurate prediction.

    更新日期:2018-09-07
  • Spark Plasma Sintering of fine-grained ceramic-metal composites YAG:Nd-(W,Mo) based on garnet-type oxide Y2.5Nd0.5Al5O12 for inert matrix fuel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-07
    L.S. Golovkina, A.I. Orlova, A.V. Nokhrin, M.S. Boldin, Е.А. Lantsev, V.N. Chuvil'deev, N.V. Sakharov, S.V. Shotin, A. Yu Zelenov

    Powders of garnet-type complex oxide Y2.5Nd0.5Al5O12 (YAG:Nd) – x vol.% W, Mo (x = 0, 10, 20) were obtained using wet chemistry techniques and precipitation of the metal phase from salt solutions. After synthesis, nanoparticles of YAG:Nd garnet cluster into submicron agglomerates. Spark Plasma Sintering (SPS) was used to obtain fine-grain ceramic-metal composites YAG:Nd – (W, Mo) with a relative density of ∼99%. The effect that metal concentration (tungsten, molybdenum) has on density, microstructure properties, and mechanical properties (microhardness, fracture toughness) of ceramics was studied. SPS activation energies were determined and it was demonstrated that the intensity of shrinkage of the fine-grained YAG:Nd – (W,Mo) ceramic-metal composites close to the activation energy of creep in tungsten (molybdenum). It has been demonstrated that the intensity of creep during SPS of fine-grained composites at medium temperatures is determined by the intensity of volume oxygen diffusion in the lattice of garnet nanoparticles, whereas at high temperatures it is determined by grain-boundary diffusion.

    更新日期:2018-09-07
  • Thermal conductivity degradation and recovery in ion beam damaged tungsten at different temperature
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-05
    Shuang Cui, Russ P. Doerner, Michael J. Simmonds, Chuan Xu, Yongqiang Wang, Edward Dechaumphai, Engang Fu, George R. Tynan, Renkun Chen
    更新日期:2018-09-05
  • In-situ observation of the dynamic behavior of cascade defect clusters formed by irradiation with high-energy self-ions at 50 K in Cu
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-05
    K. Ono, M. Miyamoto, K. Yamahaku

    We have observed the development of cascade defects formed by irradiation with 400 keV Cu+ ions in high-purity Cu at 50 K, and demonstrated the one-dimensional motion of freely or sporadically migrating small defect clusters during post annealing, using in-situ electron microscopy. The diffusivity for the back-and-forth one-dimensional motion of the freely migrating clusters during the post annealing at 50 K was evaluated based on the random walk theory. The free migration distance during the sporadic 1-D motion of the small clusters decreased with increasing the fluence of irradiated ions, which suggests trapping sites of radiation-induced defects. A significant number of cascade induced defects changed their positions and shapes at 50 K, and were annihilated by post annealing at 100 or 150 K. The present results for the free or sporadic 1-D motion of small defect clusters have a great impact on the current understanding of defect recoveries in stages ID+E and II for fcc metals.

    更新日期:2018-09-05
  • Forensic study of early stages of the Chernobyl accident: Story of three hot particles
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-05
    Andrey A. Shiryaev, Irina E. Vlasova, Vasily O. Yapaskurt, Boris E. Burakov, Alexey A. Averin, Ivan Elantyev

    Three contrasting hot particles ejected from the core of the 4th Unit of Chernobyl nuclear power plant at an early stage of Chernobyl accident were studied using complementary analytical methods: including γ-spectrometry, SEM-E(W)DX, EBSD, Raman spectroscopy and Secondary Ions Mass Spectrometry. The particles span range from dispersed UO2+x fuel to a fragment of zirconia with traces of U and to chemically and structurally complex ZrUO particle. These particles represent wide variety of processes in the reactor during the accident development and likely originate in spatially distinct domains. Whereas the fuel particle is virtually unaltered, the zirconia particle records interaction of zircaloy with fuel and structural steel, albeit rather short one. Finally, the last particle comprises remarkable mix of various phases and was most likely formed at an advanced stage of the accident when significant interaction of the fuel with surrounding materials and eventual displacement of the reaction products took place. This particle alone reproduces rather wide range of interactions encountered during various in-pile experiments. Ubiquitous presence of Fe in the ZrUO phases reveals interaction of spacer grids with ZrUO melt. Conditions leading to formation of these particles at early stages of the accident are discussed.

    更新日期:2018-09-05
  • Site specific, high-resolution characterisation of porosity in graphite using FIB-SEM tomography
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-04
    José David Arregui-Mena, Philip D. Edmondson, Anne A. Campbell, Yutai Katoh
    更新日期:2018-09-04
  • APT-studies of phase formation features in VVER-440 RPV weld and base metal in irradiation-annealing cycles
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-01
    S.V. Fedotova, E.A. Kuleshova, B.A. Gurovich, A.S. Frolov, D.A. Maltsev, G.M. Zhuchkov, I.V. Fedotov

    This paper presents a thorough APT study of VVER-440 RPVs WM and BM samples cut out from the inner RPV surface. The features of phase formation (number densities, sizes and composition) in VVER-440 RPV steels under primary irradiation, recovery annealing, re-irradiation, recovery re-annealing and subsequent accelerated re-irradiation are identified to justify its lifetime extension up to 60 years by recovery re-annealing.It is shown that primary and re-irradiation is accompanied by formation of radiation-induced Cu-P-based precipitates whereas recovery annealing leads to their partial dissolution and coarsening of the undissolved ones. At this, for both WM and BM under primary annealing, Cu dissolves in the matrix, while under re-annealing Cu almost doesn't return to the matrix. In WM P fully dissolves in the matrix under annealing while in BM there is gradual matrix P depletion under operation (including recovery annealing) due to formation of P grain boundary segregation. Under the third irradiation cycle, Cu almost doesn't contribute to precipitation: in WM radiation-induced precipitates are P-based, enriched with Si, Ni and Mn, while in BM there are Si-based precipitates, enriched with Ni and Mn. In BM number density of precipitates is lower than in WM during all operation stages that causes its lower radiation embrittlement.

    更新日期:2018-09-03
  • Chemical imaging and diffusion of hydrogen and lithium in lithium aluminate
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-01
    Weilin Jiang, Steven R. Spurgeon, Zihua Zhu, Xiaofei Yu, Karen Kruska, Tianyao Wang, Jonathan Gigax, Lin Shao, David J. Senor

    Tritium (3H) must be replenished for strategic stockpile and fusion reactors. 6Li-enriched γ-LiAlO2 pellets have been used for 3H production by thermal neutron irradiation. A fundamental study of 3H and 6Li diffusion processes in irradiated γ-LiAlO2 pellets is needed to assess and predict the long-term material performance. This study focuses on identifying the trapping sites and diffusion pathways of 1H (a surrogate for 3H) and Li atoms in polycrystalline γ-LiAlO2 pellets irradiated with both 4He+ and 1H2+ ions. A combination of STEM-EELS, Nano-SIMS and APT is employed for chemical imaging at nanoscale resolution. There is direct evidence for preferred Li diffusion pathways along grain boundaries. Isolated voids are identified as possible Li trapping sites in the irradiated γ-LiAlO2 pellets. Possible lithium hydrides and/or hydrates are precipitated on the surface of the irradiated pellets. This study improves our understanding of H and Li diffusion processes and provides data for modelling and simulation to predict material performance during neutron irradiation of γ-LiAlO2 pellets.

    更新日期:2018-09-03
  • Surface area and particle size distribution of plutonium oxides derived from the direct oxidation of Pu metal (LA-UR-18-22581)
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-01
    David M. Wayne, Larry G. Peppers, Daniel S. Schwartz, Paul C. DeBurgomaster

    The intent of this paper is to interrogate the physical characteristics of PuO2 powders and aggregates produced by direct metal oxidation to examine possible relationships between process parameters and the physical characteristics of the product. The majority of the PuO2 samples considered were generated during the pre-production and production phases of the Advanced Recovery and Integrated Extraction System (ARIES) program at Los Alamos National Laboratory (LANL). Oxide formed passively on Pu metal surfaces at ambient atmospheric conditions is also characterized. Plutonium oxide powders from the ARIES Direct Metal Oxidation (DMO) furnace and muffle furnace process lines are very consistent in terms of their bulk and tapped density, specific surface area (SSA) and particle size distribution (PSD). The SSAs of calcined PuO2 generated in the ARIES DMO-2 and muffle furnaces are between 0.1 and 0.5 m2/g. Oxides from both the DMO-2 and muffle furnaces also had similar tri-modal PSDs with maxima at 0.73–3.16 μm, 10.9–18.5 μm, and 41.3–63.0 μm. The PSDs of the oxide formed passively on Pu metal surfaces at ambient conditions were unimodal, with near-Gaussian distributions and D50 close to 1 μm. The oxide formed passively at ambient conditions also had significantly higher SSAs (6–8 m2/g).A documented process upset that originated from a heater element failure in DMO-2 resulted in significantly lower oxidation temperatures, and produced oxide having a slightly greater SSA, and lower bulk and tapped densities. Macroscopic (by quantitative sieving) and microscopic (by laser diffraction of aqueous suspensions) PSDs of the oxide produced during this interval are characterized by depleted large-diameter particle populations. We speculate that the lack of large particles in the aggregate emerging from DMO-2 during the temperature anomaly reflects a change in the thermal regime during oxidation which favors the generation of finer-grained aggregate.

    更新日期:2018-09-03
  • Molecular dynamics investigation on the local structures and transport properties of uranium ion in LiClKCl molten salt
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-01
    Jian-Xing Dai, Wei Zhang, Cui-Lan Ren, Han Han, Xiao-Jing Guo, Qing-Nuan Li

    The effects of temperature and composition on the structures and transport properties of UCl3LiClKCl salts were systematically investigated by using molecular dynamics simulation with a polarizable force field. In a molten salt of pure UCl3, there exists a network of [UCln] (n = 6, 7, 8, …) clusters, dominated by corner-sharing, Cl-linked [UCl8]5−. The networks became sparser with the UCl3 concentration decreasing. The local structures of U3+ complexes in LiClKCl salts were also compared with that of typical fission products Sc3+, Y3+, La3+ and Tb3+. It is found that the local structure of the U3+ complexes in LiClKCl were very similar to that of La3+, while the UCl coordination bonds were less stable than that of YCl, ScCl, and TbCl complexes. Finally, two basic transport properties, diffusion coefficient and shear viscosity of UCl3LiClKCl were predicted to a broader range than reported before. These results help to understand the underlying mechanisms relative to the pyrochemical processing for separation of trivalent uranium ions from other fission product ions in LiClKCl salts.

    更新日期:2018-09-03
  • Lattice thermodynamic behavior in nuclear fuel ThO2 from first principles
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-09-01
    Jianye Liu, Zhenhong Dai, Xiuxian Yang, Yinchang Zhao, Sheng Meng
    更新日期:2018-09-03
  • In-situ TEM study of the ion irradiation behavior of U3Si2 and U3Si5
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-31
    Tiankai Yao, Bowen Gong, Lingfeng He, Yinbin Miao, Jason Harp, Michael Tonks, Jie Lian

    U3Si2 and U3Si5 are two important uranium silicide phases currently under extensive investigation as potential fuel forms or components for light water reactors (LWRs) to enhance accident tolerance. In this paper, their irradiation behaviors are studied by ion beam irradiations with various ion mass and energies, and their microstructure evolution are investigated by in-situ transmission electron microscopy (TEM). U3Si2 can easily be amorphized by ion beam irradiations (by 1 MeV Ar2+ or Kr2+) at room temperature with the critical amorphization dose less than 1 dpa. The critical amorphization temperatures of U3Si2 irradiated by 1 MeV Kr2+ and 1 MeV Ar2+ ion are determined as 580 ± 10 K and 540 ± 5 K, respectively. In contrast, U3Si5 remains crystalline up to 8 dpa at room temperature and is stable against ion irradiation-induced amorphization up to ∼50 dpa by either 1 MeV Kr2+ or 150 KeV Kr+ at 623 K. These results provide valuable experimental data to guide future irradiation experiments, support the relevant post irradiation examination, and serve as the experimental basis for modeling validation.

    更新日期:2018-09-01
  • Defect structures and statistics in overlapping cascade damage in fusion-relevant bcc metals
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-31
    A.E. Sand, J. Byggmästar, A. Zitting, K. Nordlund
    更新日期:2018-08-31
  • Ab initio study of the stability of intrinsic and extrinsic Ag point defects in 3CSiC
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-29
    Nanjun Chen, Qing Peng, Zhijie Jiao, Isabella van Rooyen, William F. Skerjanc, Fei Gao

    We have systematically investigated the energetics and stability of Ag atom in 3CSiC with various charge states using first-principles calculations within large supercells. Up to 18 Ag-defect configurations have been examined, including substitutionals, interstitials, and vacancy-based complexes. A general trend is that the formation energy of Ag-defect complexes is generally lower than interstitial typed defects. With the lowest formation energy, the configuration with Ag_TSi-VC3+ turns out to be the most stable one. It has also been found a neutral Ag is more likely to substitute a silicon lattice site with a nearest carbon vacancy, thus forming an AgSi-VC pair. All these data are important inputs in the next coarser-level modeling to understand the Ag migration in and release from 3CSiC under both thermal and radiation conditions.

    更新日期:2018-08-30
  • Ab-initio calculation on electronic and optical properties of ThO2, UO2 and PuO2
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-29
    Shilpa Singh, Sanjeev K. Gupta, Yogesh Sonvane, K.A. Nekrasov, A. Ya Kupryazhkin, P.N. Gajjar

    We have investigated the structural and electronic properties of oxides of Th, U and Pu using GGA + U method. Structure of these oxides is of cubic nature and they have indirect band gaps of 4.34 eV along M→R (ThO2), 2.30 eV along Γ→R (UO2) and 2.27 eV along M→R (PuO2). The density of states (DOS) of these oxides shows main contribution of 2p orbital in valence band maxima of ThO2 and PuO2 while in UO2 5f orbital contributes mainly in VBM. We also investigated the optical properties of these oxides and found that static dielectric function increases from ThO2 to PuO2.

    更新日期:2018-08-30
  • Sintering behavior of UO2Er2O3 mixed fuel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-29
    Michelangelo Durazzo, Artur C. Freitas, Alberto E.S. Sansone, Nildemar A.M. Ferreira, Elita F. Urano de Carvalho, Humberto G. Riella, Ricardo M. Leal Neto
    更新日期:2018-08-29
  • Synthesis and characterization of oxyapatite [Ca2Nd8(SiO4)6O2] and mixed-alkaline-earth powellite [(Ca,Sr,Ba)MoO4] for a glass-ceramic waste form
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-28
    Jacob A. Peterson, Jarrod V. Crum, Brian J. Riley, R. Matthew Asmussen, James J. Neeway

    This paper discusses the synthesis, characterization, and chemical durability assessment of oxyapatite [Ca2Nd8(SiO4)6O2] and mixed-alkaline-earth powellite [(Ca,Sr,Ba)MoO4]. These are the major crystalline phases that precipitate from the melt during cooling of a glass ceramic waste form currently being evaluated for immobilizing wastes produced during the aqueous reprocessing of used nuclear fuel. The oxyapatite was made at 99.7% purity using a solution-based process followed by heat treatments. The powellite, made by melting carbonates and slow-cooling the melt, formed two different phases, one rich in Ca (74.7%) and Sr (19.4%) (balance is Ba) and the other rich in Sr (25.6%) and Ba (65.6%) (balance is Ca). Following static dissolution tests after 5 days, the oxyapatite phase had a maximum normalized loss of 0.024 g m−2 for Ca and the powellite phases showed a higher normalized loss from the Ba-powellite (0.22 g m−2) compared with the Sr-powellite (0.06 g m−2) and Ca-powellite (0.02 g m−2). Additionally, crystal structure data were measured using X-ray diffraction and are compared in detail with literature data of powellites and oxyapatites of similar chemistries.

    更新日期:2018-08-29
  • Oxidation behavior of (U1-yCey)O2.00; (y = 0.21, 0.28 and 0.44) solid solutions under different oxygen potentials. Thermogravimetric and in situ X-ray diffraction studies
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-24
    S.K. Sali, Meera Keskar, Rohan Phatak, K. Krishnan, Geeta P. Shelke, P.P. Muhammed Shafeeq, S. Kannan

    For the first time, oxidation behavior of (U1-yCey)O2.00; (y = 0.21, 0.28 and 0.44) was examined under various oxygen potentials ranging from 0 to −58 kJ/mol. Thermogravimetry (TG) and powder X-ray powder diffraction (XRD) were used as prime techniques for determination of oxygen stoichiometry and identification of different compounds formed. Rietveld analysis of XRD data of the oxidized products, formed under different oxygen potentials, showed the formation of different proportions of orthorhombic M3O8 (M = U + Ce) and face centered cubic (FCC) MO2+x phases for y = 0.21 and 0.28 and a single FCC phase for y = 0.44. A novel method based on TG and XRD analysis was used to get the quantitative information on the distribution of uranium, cerium and oxygen between the product phases. It has been observed that lowering of oxygen potential resists the formation of M3O8 in the oxidized products and potential lower than −70 kJ/mol is required for maintaining single FCC phase during the storage of fast reactor fuel. The oxidation kinetics of (U1-yCey)O2.00; (y = 0.21, 0.28 and 0.44) was studied using model free iso-conversional method. High temperature X-ray diffraction (HT–XRD) studies in vacuum and oxygen atmospheres were used to separate out the oxidation effect from the combined effect of expansion and oxidation. For the first time, HT–XRD and TG studies were coupled together to correlate the change in lattice parameter of FCC phase (Δa) obtained in vacuum and oxygen atmosphere, with change in oxygen to metal ratio (ΔO/M). The influence of phase separation on the correlation of lattice parameter with O/M was also discussed.

    更新日期:2018-08-26
  • 更新日期:2018-08-24
  • Alkali-activated materials for radionuclide immobilisation and the effect of precursor composition on Cs/Sr retention
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-23
    Niels Vandevenne, Remus Ion Iacobescu, Robert Carleer, Pieter Samyn, Jan D'Haen, Yiannis Pontikes, Sonja Schreurs, Wouter Schroeyers
    更新日期:2018-08-23
  • Effect of pre-existing dislocations on the formation of dislocation loops: Pure magnesium under electron irradiation
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-23
    Qingshan Dong, Zhongwen Yao, Peyman Saidi, Mark R. Daymond
    更新日期:2018-08-23
  • A parametric study of operating carbon anodes in the oxide reduction process
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-23
    A. Merwin, P. Motsegood, J. Willit, M.A. Williamson

    A parametric study of the electrolytic reduction of uranium dioxide in molten lithium chloride – lithium oxide using carbon anodes was conducted to determine the operational parameter values necessary for high yields. Operational parameters evaluated in this study include anode and cathode current densities, anodic polarization, UO2 batch size, and method of electrochemical control. Seven oxide reduction experiments were conducted with between 25 g and 100 g UO2. The current density on the cathode was the critical parameter, which indicates the reduction process is kinetically controlled by cathodic reactions. High cell currents were achieved without the application of high anodic potentials by using a large anode surface area. This approach facilitates efficient reduction at high throughput without production of chlorine or other corrosive gasses.

    更新日期:2018-08-23
  • Pore pressure estimation in irradiated UMo
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-22
    D. Salvato, A. Leenaers, S. Van den Berghe, C. Detavernier

    Image analysis was performed on SEM micrographs of recently irradiated UMo dispersion fuel plates. Detailed information on the fission gas inter-granular bubbles accompanying recrystallization, including average diameter, density and size distribution were extracted and compared with previous results on UMo and UO2. A pore density drop was notice at high fission densities and attributed mainly to a pore coarsening dominated by irradiation induced phenomena. Based on the image analysis data and theoretical considerations, a model was developed to estimate the pressure inside the pores as a function of fission density, temperature and pore radius. The developed pressure can give indications of the mechanical stability of the fuel towards the progressive building-up of fission gases. Finally, the proposed methodology was applied to the nanobubble lattice decorating the fuel grains at low fission densities in order to infer the physical state of the contained fission gases. The estimated values suggest the presence of solid xenon precipitates.

    更新日期:2018-08-22
  • Effect of irradiation and irradiation defects on the mobility of Σ5 symmetric tilt grain boundaries in iron: An atomistic study
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-22
    X.Y. Wang, N. Gao, B. Xu, Y.N. Wang, G.G. Shu, C.L. Li, W. Liu

    Body center cubic iron materials are commonly used in nuclear power plants. In bcc-iron, symmetric tilt grain boundaries (STGBs), which are believed to play important roles on self-healing, are susceptible to configuration and structural change under irradiation. The effect of such changes on mechanical properties of such GBs is still unknown. In this work, using molecular dynamics simulations, we find that the critical shear stress τc required for Σ5 STGB migration is greatly reduced by either displacement cascade nearby or absorption of defect clusters. Moreover, we find that the trapping of radiation-induced interstitial-type dislocation loops could also decrease τc, indicating that displacement cascades could also exert a long-range effect on GBs due to the fast diffusion of interstitial clusters.

    更新日期:2018-08-22
  • Effect of neon on the hydrogen behaviors in tungsten: A first-principles study
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-22
    Guangdong Liu, Shifang Xiao, Hong-Bo Zhou, Zhixiao Liu, Wangyu Hu, Fei Gao, Huiqiu Deng

    Seeding the impurity elements in W can affect the H behaviors. The effect of Ne on H behaviors in W bulk and W (110) surface has been investigated with the first-principles density functional theory (DFT) calculations in the present paper. It is found that the interactions between the interstitial Ne and H are attractive in bulk W. Ne atom can reduce the binding energies for H in the vacancy and divacancy, and the binding energies decrease linearly with the ratio of Ne and vacancy. In the W (110) surface without vacancy, the threefold hollow site is the most preferable site for H adsorption; Ne atom cannot be adsorbed on the W surface and in the subsurface because of the weak interaction. The substitutional Ne in subsurface can impede H absorption, and increase the diffusion barrier for H from the surface into subsurface.

    更新日期:2018-08-22
  • Permeability of observed three dimensional fracture networks in spent fuel pins
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-21
    Robin N. Thomas, Adriana Paluszny, David Hambley, Frazer M. Hawthorne, Robert W. Zimmerman

    The three-dimensional fracture network within a spent fuel pin is characterised using sequential grinding, and its permeability is numerically estimated. Advanced Gas-cooled Reactors (AGRs) produce spent fuel pins consisting of an outer steel cladding enclosing ceramic uranium dioxide (UO2) pellets. During irradiation, fuel pellets may undergo fracturing due to thermal, densification and swelling effects. Fracture patterns are usually observed on the surface of the pellet or through a cross section or longitudinal plane along the pellet. In this work, the three-dimensional fracture pattern within the pellet is characterised using an optical microscope. The pellet is progressively ground and polished, providing sequential cross sections, which together yield a three-dimensional discrete fracture pattern. Multiple large fractures grow to connect the cladding to the internal region of the pellet. Multiple surface fractures are observed that do not penetrate into the matrix of the pellet. The porosity of the UO2 and apertures of the fractures are estimated by sampling microscopic images. Darcy flow is numerically solved using the finite element method, computing flow through the matrix and fractures simultaneously. The equivalent tensorial permeability of the system is estimated for various approximate fracture apertures. The fracture network raises the permeability of the pellet by an order of magnitude.

    更新日期:2018-08-21
  • Co-dependent microstructural evolution pathways in metastable δ-ferrite in cast austenitic stainless steels during thermal aging
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-21
    Timothy G. Lach, Arun Devaraj, Keith J. Leonard, Thak Sang Byun
    更新日期:2018-08-21
  • Investigation of electrochemical kinetics for La(III)/La reaction in molten LiClKCl eutectic salt using potentiometric polarization
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    Shaoqiang Guo, Evan Wu, Jinsuo Zhang

    Although the activity coefficients and diffusion coefficients of LaCl3 in molten LiClKCl salt have been extensively studied, the data on the electrochemical properties of La(III)/La reaction in the melt are very scarce. Among the few measurements, up to five orders of magnitude discrepancies for the reaction rate constant k 0 have been reported. In the present study, potentiodynamic polarization measurements at different LaCl3 concentrations (1–6 wt%) and temperatures (723–823 K) are conducted to obtain accurate values of the exchange current density i 0 and charge transfer coefficient α . The data are analyzed using non-simplified electrode kinetic equations through the optimization fitting procedures, and conventional Tafel and linear polarization (LP) methods. The concentration dependence of i 0 obtained from the optimization fitting method is examined, showing agreement with the theoretical correlation i 0 = n F k 0 c ( 1 − α ) . The corresponding values of k 0 are 1.76–3.22 × 10−5 cm s−1 in 723–823 K.

    更新日期:2018-08-18
  • Understanding irradiation-induced nanoprecipitation in zirconium alloys using parallel TEM and APT
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    A. Harte, R. Prasath Babu, C.A. Hirst, T.L. Martin, P.A.J. Bagot, M.P. Moody, P. Frankel, J. Romero, L. Hallstadius, E.C. Darby, M. Preuss
    更新日期:2018-08-18
  • Surface morphology of F82H steel exposed to low-energy D plasma at elevated temperatures
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-18
    V. Kh. Alimov, Y. Hatano, N. Yoshida, N.P. Bobyr, M. Oyaidzu, M. Tokitani, T. Hayashi

    Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 1022 D/m2s) deuterium (D) plasma at 623–773 K to various D fluencies in the range from 1 × 1025 to 2.5 × 1026 D/m2. The surface morphology of the plasma-exposed targetswas examined with a field-emission scanning electron microscope. Cross-sectional observations of nano-structures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiber-like layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure.

    更新日期:2018-08-18
  • Effect of deformation twinning on dissolution corrosion of 316L stainless steels in contact with static liquid lead-bismuth eutectic (LBE) at 500 °C
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-18
    Oksana Klok, Konstantina Lambrinou, Serguei Gavrilov, Erich Stergar, Jun Lim, Tom Van der Donck, Wouter Van Renterghem, Iris De Graeve
    更新日期:2018-08-18
  • Radiation tolerance of commercial and advanced alloys for core internals: A comprehensive microstructure characterization
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    Miao Song, Calvin R. Lear, Chad M. Parish, Mi Wang, Gary S. Was

    Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.

    更新日期:2018-08-17
  • 更新日期:2018-08-17
  • Model of nuclear fuel pellets densification under irradiation and isothermal conditions: Application to UO2 fuels
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    Mauricio E. Cazado, Alicia C. Denis

    The dimensional changes of a nuclear fuel in operation are strongly determined by two opposite effects. One of them is due to contraction of the as-fabricated pores, giving place to densification which is evident during the first stages of irradiation. This effect is counteracted by the swelling phenomenon provoked by the fission products that progressively accumulate in the fuel material. A model to evaluate the changes in fuel pellets porosity due to radiation and thermal effects taking into account the point defects flow to and from the pores is presented. A simplification of the model to assess the progress of porosity in isothermal re-sintering tests is also given. Simulations are compared with experimental data measured on UO2 fuel pellets with a variety of microstructures at different temperatures and radiation conditions, attaining a good agreement.

    更新日期:2018-08-17
  • Irradiation effects in high entropy alloys and 316H stainless steels at 300 °C
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    Wei-Ying Chen, Xiang Liu, Yiren Chen, Jien-Wei Yeh, Ko-Kai Tseng, Krishnamurti Natesan

    High entropy alloys (HEAs) have been considered for applications in nuclear reactors due to their promising mechanical properties, corrosion and radiation resistance. It has been suggested that sluggish diffusion kinetics and lattice distortion of HEAs can enhance the annihilation of irradiation-induced defects, giving rise to a higher degree of tolerance to irradiation damage. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages over conventional austenitic stainless steels (SS), we performed in-situ ion irradiation experiments with 1 MeV krypton at 300 °C on two HEAs and a 316H SS under an identical irradiation condition. The irradiation introduced a high density of dislocation loops in all materials, and the microstructural evolution as a function of dose was similar for HEAs and 316H SS. Nanoindentation tests showed that the degree of irradiation hardening was also comparable between them. The similar microstructural evolution and irradiation hardening behavior between the HEAs and 316H indicate that, at low temperatures (≤300 °C), the irradiation damage of fcc alloys is not sensitive to compositional variation and configurational entropy.

    更新日期:2018-08-17
  • Effect of tempering time on fatigue crack growth behavior of CLAM steel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-17
    Mengtian Liang, Yanyun Zhao, Mingjie Zheng, Xiaodong Mao

    Effect of tempering time on fatigue crack behavior of China Low Activation Martensitic (CLAM) steel was studied at room temperature in air. The fatigue crack growth rates subjected to the tempering time range of 90–9000 min were investigated. The results showed that the fatigue crack growth decelerated in the near-threshold region as the tempering time increased. The fatigue crack growth rate after tempered for 9000 min was approximately one-third of that of the specimen tempered for 90 min with ΔK of 14 MPa m0.5. The improved fatigue crack growth resistance could be explained by the influence of variation in microstructure, i.e., martensitic laths and precipitates. This study suggested that the effective driving force for fatigue crack growth was reduced in the near-threshold region by the activated roughness-induced crack closure mechanism as a result of rougher fracture surface caused by coarsened martensitic laths and precipitates.

    更新日期:2018-08-17
  • 更新日期:2018-08-17
  • Stability of U5Si4 phase in U-Si system: Crystal structure prediction and phonon properties using first-principles calculations
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-14
    D.A. Lopes, V. Kocevski, T.L. Wilson, E.E. Moore, T.M. Besmann

    U-Si systems have recently received considerable attention due to the potential application of U3Si2 as a high-density fuel under an accident tolerant fuel initiative. However, the thermodynamic stability of the more recently reported adjacent U5Si4 phase is uncertain and could play a significant role in fuel performance. In this work, the enthalpy of formation of the phase predicted by density functional theory (DFT) using the DFT + U formalism is used with an evolutionary algorithm (USPEX) to evaluate stability and possible atomic structures for U5Si4. The structure of U-Si convex hull phases and the confirmed U3Si2 structure were predicted providing confidence in the reliability of the evolutionary algorithm, as well as to obtain a convex hull with comparable enthalpies of formation. Subsequently, the code was applied to the U5Si4 composition for cells with 18 and 36 atoms, predicting a 36-atom hexagonal symmetrized unit cell with space group P6/mmm as the lowest-energy configuration, agreeing with that experimentally reported for U5Si4. Yet, phonon calculations using the density functional perturbation theory formalism, demonstrated that the predicted structure is dynamically unstable, exhibiting negative vibrational modes for the uranium. These indicated that generated shears are directed toward the formation of potential uranium octahedral sites, analogous to those occupied by carbon atoms in U20Si16C3. It was thus concluded that omitting the U5Si4 phase from assessed U-Si phase equilibria is currently justified.

    更新日期:2018-08-15
  • Electronic structure and thermophysical properties of U3Si2: A systematic first principle study
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-15
    K. Srinivasu, Brindaban Modak, Tapan K. Ghanty
    更新日期:2018-08-15
  • Elevated temperature tensile tests on DU–10Mo rolled foils
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-15
    Jason Schulthess, Randy Lloyd, Barry Rabin, Michael Heighes, Tammy Trowbridge, Emmanuel Perez

    Studies were completed to obtain tensile mechanical properties for uranium-10 wt.% molybdenum (U10Mo) foils which were subjected to four different thermomechanical processing conditions. UMo alloy foils are being investigated to support fuel conversion of high power research reactors from their current high enriched fuel form to a low enriched fuel form. Mechanical properties of the fuel foil have an effect on irradiation performance and fuel fabrication and therefore are required to support modeling and qualification of new low-enriched uranium monolithic fuel plate designs. The data contained in this document contributes to fuel qualification by fulfilling the requirement that physical properties related to fuel meat be established. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low-enriched U10Mo to be used in actual fuel plates; therefore DU–10Mo was studied to simplify material processing, handling, and testing requirements. In this report, the different thermomechanical treatments included variations of wrought hot and cold rolling reduction and post rolling annealing. Each of the four foils was hot rolled. After hot rolling reduction, three of the four foils were further reduced by cold rolling. One of the three was reduced a further 20% by cold rolling, and the remaining two were reduced 50% by cold rolling. Following cold rolling reduction, one of the two foils which had been reduced 50% by cold rolling was annealed at 650 °C. Performing this analysis allows assessment of the impact of foil fabrication history on the resultant tensile properties DU–10Mo fuel foils. Tensile properties of DU–10Mo at room temperature through approximately 400 °C determined from the tests conducted herein suggest the material is stronger and has lower ductility than what has been reported previously in the literature. The explanation for these differences has yet to be determined, but is likely related to differences in grain size and/or impurity content, and variation in fabrication history. At the highest temperatures tested (550 °C) better agreement between the values reported here and available literature was found. As expected, yield and ultimate tensile strength decreased with increasing test temperature. Generally, the yield stress for all foil processing conditions was found to be in the range of 1100 MPa for room temperature tests, and in the range of 200 MPa for tests conducted at 550 °C. Ultimate tensile stress was in the range of 1175 MPa at room temperature, decreasing to approximately 225 MPa at 550 °C. Elongation increased significantly, from 0 to 2% at room temperature to 50% or more for the tests at 550 °C.

    更新日期:2018-08-15
Some contents have been Reproduced with permission of the American Chemical Society.
Some contents have been Reproduced by permission of The Royal Society of Chemistry.
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