Radiation-induced segregation on dislocation loops in austenitic Fe-Cr-Ni alloys

被引:9
作者
Dai, C. [1 ]
Saidi, P. [2 ]
Langelier, B. [3 ]
Wang, Q. [1 ]
Judge, C. D. [4 ]
Daymond, M. R. [2 ]
Mattucci, M. [1 ]
机构
[1] Canadian Nucl Labs, Chalk River, ON K0J 1J0, Canada
[2] Queens Univ, Dept Mech & Mat Engn, Kingston, ON K7L 3N6, Canada
[3] McMaster Univ, Canadian Ctr Electron Microscopy, 1280 Main St West, Hamilton, ON L8S 4L8, Canada
[4] Idaho Natl Lab, Idaho Falls, ID 83415 USA
基金
加拿大自然科学与工程研究理事会; 加拿大创新基金会;
关键词
GRAIN-BOUNDARY SEGREGATION; STAINLESS-STEELS; NEUTRON-IRRADIATION; MICROSTRUCTURAL EVOLUTION; ATOMISTIC SIMULATIONS; SOLUTE SEGREGATION; EDGE DISLOCATION; FRANK LOOPS; TEMPERATURE; ENERGY;
D O I
10.1103/PhysRevMaterials.6.053606
中图分类号
T [工业技术];
学科分类号
08 ;
摘要
Radiation-induced segregation of alloying elements to crystallographic defects is commonly observed in irradiated austenitic stainless steels. The interaction between solutes and radiation-induced defects changes the physical distribution of solutes and thus affects the formation and growth of defects. The change of the microstructure consequently affects the mechanical properties of the material. A qualitative and quantitative understanding of the interaction between solutes and defects is desirable to better predict the service lifetime of nuclear materials. We used atom probe tomography to measure the distribution of solutes at dislocation loops in 304L stainless steel, irradiated with 2 MeV protons up to 1.5 displacements per atom at 373 and 633 K. No segregation at dislocation loops was found in samples irradiated at 373 K, whereas Ni and Si enrichment and Cr depletion were detected at dislocation loops irradiated at 633 K. The experimentally observed perfect and faulted dislocation loops in vacancy and interstitial types were reproduced by molecular dynamics (MD). A hybrid MD/Monte Carlo method was used to predict the redistribution of alloying atoms at all possible types of dislocation loops in face-centered cubic Fe-Cr-Ni alloys at the same irradiated temperatures (373 and 633 K). The simulations show that, at both temperatures, Cr clusters were formed and distributed randomly, and Ni atoms enriched or depleted at interstitial or vacancy dislocation loops, respectively. The change of solute concentration reaches the highest at the edge of the loop. Ni profiles exhibit characteristic behavior in terms of the stress field of the loops: tension inside of vacancy loops showing depletion of Ni atoms compared with compression inside of interstitial loops showing enrichment of Ni atoms. In addition, the stress field is reduced after solute redistribution. The absence of alloying segregation observed in experiments at a lower temperature (373 K) is explained by a rate theory model: Low-temperature irradiation requires significantly longer irradiation time to see the same amount of segregation as at high temperatures because of the extremely low diffusion of vacancies at low temperatures.
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页数:15
相关论文
共 89 条
[71]   Thermodynamics of an austenitic stainless steel (AISI-348) under in situ TEM heavy ion irradiation [J].
Tunes, Matheus A. ;
Greaves, Graeme ;
Kremmer, Thomas M. ;
Vishnyakov, Vladimir M. ;
Edmondson, Philip D. ;
Donnelly, Stephen E. ;
Pogatscher, Stefan ;
Schon, Claudio G. .
ACTA MATERIALIA, 2019, 179 :360-371
[72]   Evolution of the radiation-induced defect structure in 316 type stainless steel after post-irradiation annealing [J].
Van Renterghem, W. ;
Konstantinovic, M. J. ;
Vankeerberghen, M. .
JOURNAL OF NUCLEAR MATERIALS, 2014, 452 (1-3) :158-165
[73]   Average-atom interatomic potential for random alloys [J].
Varvenne, Celine ;
Luque, Aitor ;
Noehring, Wolfram G. ;
Curtin, William A. .
PHYSICAL REVIEW B, 2016, 93 (10)
[74]   Vacancy clustering in zirconium: An atomic-scale study [J].
Varvenne, Celine ;
Mackain, Olivier ;
Clouet, Emmanuel .
ACTA MATERIALIA, 2014, 78 :65-77
[75]   Energetics analysis of interstitial loops in single-phase concentrated solid-solution alloys [J].
Wang, Xin-Xin ;
Niu, Liang-Liang ;
Wang, Shaoqing .
JOURNAL OF NUCLEAR MATERIALS, 2018, 501 :94-103
[76]   Radiation-induced segregation in a ceramic [J].
Wang, Xing ;
Zhang, Hongliang ;
Baba, Tomonori ;
Jiang, Hao ;
Liu, Cheng ;
Guan, Yingxin ;
Elleuch, Omar ;
Kuech, Thomas ;
Morgan, Dane ;
Idrobo, Juan-Carlos ;
Voyles, Paul M. ;
Szlufarska, Izabela .
NATURE MATERIALS, 2020, 19 (09) :992-+
[77]  
Was G.S., 2016, Fundamentals of Radiation Materials science: Metals and Alloys
[78]   Irradiation-assisted stress corrosion cracking [J].
Was, Gary S. ;
Ashida, Yugo ;
Andresen, Peter L. .
CORROSION REVIEWS, 2011, 29 (1-2) :7-49
[79]   Assessment of radiation-induced segregation mechanisms in austenitic and ferritic-martensitic alloys [J].
Was, Gary S. ;
Wharry, Janelle P. ;
Frisbie, Brian ;
Wirth, Brian D. ;
Morgan, Dane ;
Tucker, Julie D. ;
Allen, Todd R. .
JOURNAL OF NUCLEAR MATERIALS, 2011, 411 (1-3) :41-50
[80]   On the mechanism of radiation-induced segregation [J].
Watanabe, S ;
Sakaguchi, N ;
Kurome, K ;
Nakamura, M ;
Takahashi, H .
JOURNAL OF NUCLEAR MATERIALS, 1997, 240 (03) :251-253