A finite-strain homogenization creep model for composite fuels under irradiation conditions is developed and verified,with the irradiation creep strains of the fuel particles and matrix correlated to the macroscale cr...
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A finite-strain homogenization creep model for composite fuels under irradiation conditions is developed and verified,with the irradiation creep strains of the fuel particles and matrix correlated to the macroscale creep responses,excluding the contributions of volumetric strain induced by the irradiation swelling deformations of fuel particles.A finite element(FE)modeling method for uniaxial tensile creep tests is established with the irradiation effects of nuclear materials taken into *** proposed models and simulation strategy are numerically implemented to a kind of composite nuclear fuel,and the predicted mesoscale creep behaviors and the macroscale creep responses are *** research results indicate that:(1)the macroscale creep responses and the mesoscale stress and strain fields are all greatly affected by the irradiation swelling of fuel particles,owing to the strengthened mechanical interactions between the fuel particles and the matrix.(2)The effective creep rates for a certain case are approximately two constants before and after the critical fission density,which results from the accelerated fission gas swelling after fuel grain recrystallization,and the effects of macroscale tensile stress will be more enhanced at higher temperatures.(3)The macroscale creep contributions from the fuel particles and matrix depend mainly on the current volume fractions varying with fission density.(4)As a function of the macroscale stress,temperature,initial particle volume fraction and particle fission rate,a multi-variable mathematical model for effective creep rates is fitted out for the considered composite fuels,which matches well with the FE *** study supplies important theoretical models and research methods for the multi-scale creep behaviors of various composite fuels and provides a basis for simulation of the thermal–mechanical behavior in related composite fuel elements and assemblies.
Two-phase flow water hammer events occur in the pipelines of the nuclear power systems and lead to transient and violent pressure shock to tube structures. For the sake of operation safety, the occurrence and severity...
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Models to describe the damage and fracture behaviors of the interface between the fuel foil and cladding in UMo/Zr monolithic fuel plates were established and numerically *** effects of the interfacial cohesive streng...
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Models to describe the damage and fracture behaviors of the interface between the fuel foil and cladding in UMo/Zr monolithic fuel plates were established and numerically *** effects of the interfacial cohesive strength and cohesive energy on the irradiationinduced thermal-mechanical behaviors of fuel plates were *** results indicated that for heterogeneously irradiated fuel plates:(1)interfacial damage and failure were predicted to be initiated near the fuel foil corner with higher fission densities,accompanied by the formation of a large gap after interface failure,which was consistent with some experimental observations;high tensile stresses in the fuel foil occurred near the edges of the failed interface,attributed to through-thickness cracking of the fuel foil,as found in some post-irradiation examinations;(2)the cohesive strength and cohesive energy of the interface both influenced the in-pile evolution behaviors of fuel plates;a lower cohesive strength or cohesive energy resulted in faster interfacial damage;(3)after interface fracture,the thickness of the whole plate increased to a greater degree(by~20%)than that of the samples without interfacial damage,which was attributed to the locally enhanced Mises stresses and the nearby creep deformations around the cracked *** study provided a theoretical basis for assessing failure in fuel elements.
Direct contact condensation (DCC) is widely appeared in the nuclear power plants and will lead to serious temperature and pressure fluctuations. For ocean nuclear power plants, the DCC is inevitably affected by the se...
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The control rod drive mechanism (CRDM) is a critical device in nuclear plants. This paper proposes a kind of CRDM applying the cylindrical linear induction motor (CLIM). First, the structure and the working principle ...
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To solve the problem of mobile robots needing to adjust their pose for accurate operation after reaching the target point in the indoor environment,a localization method based on scene modeling and recognition has bee...
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To solve the problem of mobile robots needing to adjust their pose for accurate operation after reaching the target point in the indoor environment,a localization method based on scene modeling and recognition has been ***,the offline scene model is created by both handcrafted feature and semantic ***,the scene recognition and location calculation are performed online based on the offline scene *** improve the accuracy of recognition and location calculation,this paper proposes a method that integrates both semantic features matching and handcrafted features *** on the results of scene recognition,the accurate location is obtained through metric calculation with 3D *** experimental results show that the accuracy of scene recognition is over 90%,and the average localization error is less than 1 *** results demonstrate that the localization has a better performance after using the proposed improved method.
The control rod drive mechanism (CRDM) is a kind of servo mechanism for the reactor control system of the nuclear power plant. This paper presents a cylindrical linear induction motor suitable for the CRDM. First, the...
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With the rapid advancement of artificial intelligence technology, the development of intelligent manufacturing has become an inevitable trend. Utilizing AI technology to ensure the safe operation of factories is a cru...
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The oxide dispersion strengthening(ODS)high entropy alloy(HEA)exhibits the high elevated temperature performance and radiation resistance due to severe atomic lattice distortion and oxide particles dispersed in matrix...
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The oxide dispersion strengthening(ODS)high entropy alloy(HEA)exhibits the high elevated temperature performance and radiation resistance due to severe atomic lattice distortion and oxide particles dispersed in matrix,which is expected to become the most promising structural material in the next generation of nuclear energy ***,microstructure and damage evolution of irradiated ODS HEA under loading remain elusive at submicron scale using the existing simulations owing to a lack of atomic-lattice-distortion information from a micromechanics ***,the random field theory informed discrete dislocation dynamics simulations based on the results of high-resolution transmission electron microscopy are developed to study the dislocation behavior and damage evolution in ODS HEA considering the influence of severe lattice distortion and nanoscale oxide ***,the damage behavior shows an unusual trend of the decreasing-to-increasing transition with the continuous loading *** are two main types of damage micromechanics generated in irradiated ODS HEA:the dislocation loop damage in which the damage is controlled by irradiation-induced dislocation loops and their evolution,the strain localization damage in which the damage comes from the dislocation multiplication in the local plastic *** oxide particle hinders the dislocation movement in the main slip plane,and the lattice distortion induces the dislocation sliding to the secondary slip plane,which promotes the dislocation cross-slip and dislocation loop annihilation,and thus reduces the material damage in the elastic damage *** findings can deeply understand atomic-scale damage mechanism and guide the design of ODS HEA with high radiation resistance.
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