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Interaction of edge dislocations with voids in tungsten

Tungsten 2024年第3期6卷 633-646页

作者： Arseny Kazakov Rita I.Babicheva Aleksandr Zinovev dmitry terentyev Kun Zhou Elena A.Korznikova Sergey V.Dmitriev Research Laboratory for Metals and Alloys under Extreme Impacts Ufa University of Science and Technology Institute for Metals Superplasticity Problems of RAS School of Mechanical and Aerospace Engineering Nanyang Technological University Institute for Nuclear Energy Technology Belgian Nuclear Research Center Technological Machines and Equipment Department Ufa State Petroleum Technological University Institute of Mechanical Science Problems of RAS Institute of Molecule and Crystal Physics UFRC of Russian Academy of Sciences

Tungsten is widely used as a material capable of withstanding extreme operating conditions and is one of the candidates for use in fusion and Generation Ⅳ fission reactors.Atomistic modelling implemented to solve various material science problems is an indispensable approach for improving material properties.This work analyzes the interaction of an edge dislocation with a void in single-crystal tungsten with the help of molecular dynamics simulation.Two different approaches for studying the dynamics of dislocations in tungsten that differ in boundary conditions are compared.Multiple interactions of the dislocation and void at different temperatures are investigated,and based on the obtained stress-strain curves,the dislocation obstacle strength for voids having various diameters is defined.It is shown that when applied to a crystal with small voids,the model with the fixed boundary conditions returns incorrect results.Moreover,the influence of the cell size for the model with periodic boundary conditions is analyzed,and the case of a non-central interaction of a dislocation with a void is considered.The obtained results improve our understanding of the mechanical response of irradiated materials.

关键词： Tungsten Molecular dynamics method Dislocations Voids Boundary conditions

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Theoretical models for irradiation hardening and embrittlement in nuclear structural materials:a review and perspective

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Acta Mechanica Sinica 2020年第2期36卷 397-411页

作者： Xiazi Xiao dmitry terentyev Haijian Chu Huiling Duan Department of Mechanics School of Civil EngineeringCentral South UniversityChangsha 410075China Structural Materials Group Institute of Nuclear Materials ScienceSCK-CEN2400 MolBelgium Shanghai Key Laboratory of Mechanics in Energy Engineering Shanghai Institute of Applied Mathematics and MechanicsSchool of Mechanics and Engineering ScienceShanghai UniversityShanghai 200444China State Key Laboratory for Turbulence and Complex System Department of Mechanics and Engineering Science College of EngineeringPeking UniversityBeijing 100871China CAPT HEDPS and IFSA CollaborativeInnovation Center of MoEBIC-ESATPeking UniversityBeijing 100871China

The study of irradiation hardening and embrittlement is critically important for the development of next-generation structural materials tolerant to neutron irradiation,and could dramatically affect the approach to the design of components for advanced nuclear reactors.In addition,a growing interest is observed in the field of research and development of irradiation-resistant materials.This review aims to provide an overview of the theoretical development related to irradiation hardening and embrittlement at moderate irradiation conditions achieved in recent years,which can help extend our fundamental knowledge on nuclear structural materials.After a general introduction to the irradiation effects on metallic materials,recent research progress covering theoretical modelling is summarized for different types of structural materials.The fundamental mechanisms are elucidated within a wide range of temporal and spatial scales.This review closes with the current understanding of irradiation hardening and embrittlement,and puts some perspectives deserving further study.

关键词： Irradiation hardening Irradiation embrittlement Theoretical models Fundamental mechanisms

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Neutron irradiation effects on mechanical properties of ITER specification tungsten

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Tungsten 2021年第4期3卷 415-433页

作者： dmitry terentyev Chih-Cheng Chang Chao Yin AZinovev Xin-Fu He Structural Materials Group Institute of Nuclear Materials ScienceBelgian Nuclear Research Center2400 MolBelgium Institute of Mechanics Materials and Civil EngineeringUCLouvain1348 Ottignies-Louvain-la-NeuveBelgium School of Nuclear Science and Technology University of Science and Technology of ChinaHefei 230026China China Institute of Atomic Energy Beijing 102413China

In this contribution,we present the results of recent neutron irradiation campaign performed in the material test reactor BR2(Belgium)on pure tungsten.We have applied various irradiation conditions and sample geometry to assess the effect of neu-tron irradiation on hardness,bending,tensile and fracture mechanical properties.The investigated material is a commercially pure tungsten plate fabricated according to the international thermonuclear experimental reactor(ITER)specification for the application in the divertor plasma-facing components.The neutron irradiation covers a large span of temperatures and damage doses,ranging from 600 to 1200°C and 0.1-1 dpa.The obtained mechanical properties were analyzed to deduce the shift of the ductile to brittle transition temperature(DBTT)applying bending,tensile and fracture toughness-testing procedures.Then,a correlation of the fracture toughness with the change of the hardness was established.The obtained results are compared with the already published results on another ITER specification grade produced in the form of a rod.The presented and discussed results show that the performance of the compared grades in terms of the irradiation-induced embrittlement is similar,and that the irradiation in the high-temperature region(600-800°C)causes a considerable DBTT shift already at 0.2-0.5 dpa.

关键词： Tungsten Neutron irradiation Embrittlement

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