National Key Laboratory For Nuclear Fuel And Materials

Chengdu, China

National Key Laboratory For Nuclear Fuel And Materials

Chengdu, China
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Zhang L.,Shanghai JiaoTong University | Bao Y.,Shanghai Nuclear Engineering Research and Design Institute | Tang R.,National Key Laboratory For Nuclear Fuel And Materials
Nuclear Engineering and Design | Year: 2012

Supercritical water cooled reactor (SCWR) is a promising Gen IV high performance reactor which can be developed for future large capacity electric power plants. However, the material selection for fuel cladding still remains one of the key issues. For a typical prototype SCWR design with outlet coolant temperature of 510°C and pressure of 25 MPa, the hot spot temperature on fuel cladding exceeds 600°C at normal operation conditions, and will be much higher during transients. Materials for fuel cladding should have good mechanical properties to meet the harsh working conditions in order to keep the integrity of fuel rod under normal and abnormal operational conditions, while corrosion in supercritical water and neutron irradiation damage will not lead to significant loss of strength and ductility, or lead to unacceptable deformation during service life. Materials for ultra-supercritical fossil fire plants, fast breeder reactor and jet air engines etc., are proposed as the candidate materials for SCWR fuel cladding. This paper reports the results based on corrosion screening tests of candidate materials exposed in supercritical water up to 650°C. Results show that their corrosion rates increase significantly with the increase of temperature, and the protective oxide films of most candidate materials turn to be unstable above temperature of 600°C. According to the present knowledge available, austenitic stainless steels with high Cr concentration show better performance and are more potential to be the references for developing the SCWR fuel cladding material. © 2012 Elsevier B.V.


Xiao H.-X.,National Key Laboratory For Nuclear Fuel And Materials | Long C.-S.,National Key Laboratory For Nuclear Fuel And Materials
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2011

A model for the swelling behavior of fission gas in the range of low temperature of irradiated uranium dioxide fuel is established. In this paper, the finite difference method is adopted to compile a computational program and quantitatively calculated the fractions of fission gas as solid solution in UO 2, the density and average radius of intragranular bubbles and their contributions to the swelling of uranium dioxide fuel. in different burn-ups and temperatures of uranium dioxide fuel. The calculation shows that this model can be used to predict the fission gas swelling of uranium dioxide fuel versus burn-up in the range of low temperature.


Liu Y.,Southwest Petroleum University | Huang H.,National Key Laboratory For Nuclear Fuel And Materials | Pan Y.,State Key Laboratory of Advanced Technologies for Comprehensive Utilization of Platinum Metals | Zhao G.,Southwest Petroleum University | Liang Z.,Southwest Petroleum University
Journal of Alloys and Compounds | Year: 2014

The phase transition, formation enthalpies, elastic properties and electronic structure of Pt3Al alloys are studied using first-principle approach. The calculated results show that the pressure leads to phase transition from tetragonal structure to cubic structure at 60 GPa. With increasing pressure, the elastic constants, bulk modulus and shear modulus of these Pt3Al alloys increase linearly and the bond lengths of Pt-Al metallic bonds and the peak at EF decrease. The cubic Pt 3Al alloy has excellent resistance to volume deformation under high pressure. We suggest that the phase transition is derived from the hybridization between Pt and Al atoms for cubic structure is stronger than that of tetragonal structure and forms the strong Pt-Al metallic bonds under high pressure. © 2014 Elsevier B.V. All rights reserved.


Tian X.-F.,Chengdu University of Technology | Xiao H.-X.,National Key Laboratory For Nuclear Fuel And Materials | Tang R.,National Key Laboratory For Nuclear Fuel And Materials | Lu C.-H.,Chengdu University of Technology
Nuclear Instruments and Methods in Physics Research, Section B: Beam Interactions with Materials and Atoms | Year: 2014

Molecular dynamics simulations are employed to investigate the displacement cascades in U-Mo alloys. The cascade process is analyzed in detail. The effects of initial directions of primary knock-on atom (PKA), Mo content and PKA energies on the final damage state are evaluated. The results suggest that the direction of the PKA has no effect on the final primary damage state. A high content of Mo will raise the number of defects and the probability of Mo replacement. Most of the sizes of defects cluster are no larger than three and the probabilities of producing larger interstitial and vacancy clusters are increased with higher PKA energy. The fractions of Mo interstitial in clusters with size larger than three and isolated Mo interstitials is low, while more than half the total Mo interstitials are contained in dumb-bells. Finally, it is found that the number of U-U dumb-bells is the highest and the number of Mo-Mo dumb-bells is the lowest in both alloys. The number of Mo-Mo dumb-bells seems to be independent of Mo content but the numbers of U-U and U-Mo dumb-bells decline with the increase of Mo content in alloys. © 2013 Elsevier B.V. All rights reserved.


Tian X.-F.,Chengdu University of Technology | Wang H.,National Key Laboratory For Nuclear Fuel And Materials | Xiao H.-X.,National Key Laboratory For Nuclear Fuel And Materials | Gao T.,University of Sichuan
Computational Materials Science | Year: 2014

The interaction between the surface of UO2 and molecular water is a serious concern in the range of nuclear waste management. We present a first-principle investigation of the interaction between water and UO 2 (1 1 1) surface based on density functional (DFT) approach. The approximations of GGA and GGA + U were employed with the projector-augmented- wave method. Both stoichiometric and reduced surfaces were considered in our simulations. We study the atomic structures and adsorption energies of various configurations of water adsorption on UO2 (1 1 1) with water coverage of 0.25 ML. The mechanism of the interaction between water (molecular water and dissociated water) and the surface is discussed in detail. Comparing the adsorption energies for various configurations, both our GGA and GGA + U calculations suggested that molecular adsorption is more favored than dissociative adsorption of water on defect-free surface, while oxygen vacancy on the surface could make the adsorption of dissociated water more favorable. Our calculated results are in good agreement with reported experimental study and help comprehensive understanding of interactions between water and the stable UO2 (1 1 1) surface. © 2014 Published by Elsevier B.V. All rights reserved.


Wu X.-Y.,National Key Laboratory For Nuclear Fuel And Materials | Wang F.,National Key Laboratory For Nuclear Fuel And Materials | Wen B.,National Key Laboratory For Nuclear Fuel And Materials
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2012

The blister annealing has been performed on UO 2 dispersed fuel plate under various temperature conditions. The microstructure of UO 2 particles changing with the rise of temperature is studied by using the microscope. Fission gas releases at high temperature and the bubbles burst the coat to generate cracks and result in the local fracture and fall off of UO 2 particles. A lot of holes occur in the reaction layer outside of UO 2 coat because of the diffusion reaction intensified by the high temperature.


Pan Y.,Southwest Petroleum University | Lin Y.,Southwest Petroleum University | Wang H.,National Key Laboratory For Nuclear Fuel And Materials | Zhang C.,Southwest Petroleum University
Materials and Design | Year: 2015

The effect of vacancy on mechanical properties of α-Nb5Si3 is systematically investigated by first-principles calculations. Four different mono vacancies in this alloy are considered in detail. The vacancy formation energy, formation enthalpy, elastic modulus, hardness, B/G ratio, thermodynamic properties and electronic structure of α-Nb5Si3 with different vacancies are discussed. The calculated vacancy formation energies show that Nb vacancies are more stable than that of Si vacancies, and α-Nb5Si3 prefers to form the Nb-va2 vacancy. Those vacancies weaken the volume, shear deformation resistances and reduce the elastic stiffness. However, those vacancies result in brittle-to-ductile transition and α-Nb5Si3 with Si-va1 mono vacancy exhibits ductile behavior. The calculations of electronic structure reveal that these vacancies change the localized hybridization between Nb-Si and Nb-Nb atoms, which are the origin of brittle-to-ductile transition. Finally, we conclude that vacancy is beneficial for improving the ductility of Nb5Si3. © 2015 Elsevier Ltd.


Luo Q.,National Key Laboratory For Nuclear Fuel And Materials | Chen Y.,National Key Laboratory For Nuclear Fuel And Materials | Liu S.,National Key Laboratory For Nuclear Fuel And Materials
Procedia Engineering | Year: 2012

The strength and corrosion properties of austenitic steels could be improved through adding appropriate amount of nitrogen and reducing the amount of carbon, while the plasticity and toughness are not influenced. The corrosion performance of 316NG and 304NG nitrogen-containing stainless steels and 321 stainless steels which used in nuclear plants have been studied, including the uniform corrosion performance, electrochemistry corrosion performance, stress corrosion performance, pitting corrosion performance and salt spray corrosion performance. The results indicated that the corrosion resistance of 316NG and 304NG nitrogen-containing stainless steels was more excellent than the 321 stainless steel, and N and Mo in alloys improved the corrosion behaviors of stainless steels. © 2011 Published by Elsevier Ltd.


Tian X.,University of Sichuan | Gao T.,University of Sichuan | Jiang G.,University of Sichuan | He D.,University of Sichuan | Xiao H.,National Key Laboratory For Nuclear Fuel And Materials
Computational Materials Science | Year: 2012

Ab initio calculations based on density functional theory have been carried to investigate the incorporation and solution of krypton in uranium dioxide. The GGA and GGA + U approximations were used with the projector-augmented-wave method. Several defects that are likely to accommodate the incorporation of krypton in UO 2, such as oxygen and uranium vacancy, divacancy and Schottky defects were considered in this work. Both our SP-GGA and SP-GGA + U calculations suggested that the lowest incorporation energy corresponds to the divacancy. With SP-GGA method, the solution energies are positive whatever the trapping site considered, which confirms that Kr atoms are insoluble in UO 2, but notable discrepancy exits between the results calculated by SP-GGA + U and SP-GGA. We highlight that the use of SP + GGA + U significantly increases the number of the energy minima of the system. Furthermore, the concentrations of the point defects and the solution energy of Kr for the different incorporation sites as a function of the stoichiometry were also obtained when the deviation from stoichiometry is small. © 2011 Elsevier B.V. All rights reserved.


Tian X.,Chengdu University of Technology | Gao T.,University of Sichuan | Lu C.,Chengdu University of Technology | Shang J.,National Key Laboratory For Nuclear Fuel And Materials | Xiao H.,National Key Laboratory For Nuclear Fuel And Materials
European Physical Journal B | Year: 2013

The incorporation and solution of helium in plutonium dioxide have been investigated based on density functional theory. The GGA and GGA + U approximations were used with the projector-augmented-wave method. Several defects that are likely to accommodate the incorporation of helium in PuO 2, such as oxygen vacancy, plutonium vacancy, divacancy and Schotty defects were considered in this work. With GGA approach, the lowest incorporation energy corresponds to neutral trivacancy, followed by divacancy and plutonium vacancy, while the GGA + U scheme gave us that oxygen vacancy is the most favorable incorporation site for He. Both SP-GGA and SP-GGA + U methods obtained a same conclusion that the most favorable solution site for He is oxygen vacancy, interstitial site and plutonium vacancy for under-, perfect and over-stoichiometry, respectively. Additionally, the concentrations of the point defects and the solution energy of He for the different incorporation sites as a function of the stoichiometry were also obtained based on the point-defect model. © 2013 EDP Sciences, SIF, Springer-Verlag Berlin Heidelberg.

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