Redistribution of alloying elements in Zircaloy-2 after in-reactor exposure

被引：57

作者：

Sundell, G. ^{[1
]}

Thuvander, M. ^{[1
]}

Tejland, P. ^{[2
]}

Dahlback, M. ^{[3
]}

Hallstadius, L. ^{[3
]}

Andren, H. -O. ^{[1
]}

机构：

[1] Chalmers, Dept Appl Phys, SE-41296 Gothenburg, Sweden

[2] Studsvik Nucl AB, SE-61182 Nykoping, Sweden

[3] Westinghouse Elect Sweden, SE-721 Vasteras, Sweden

来源：

JOURNAL OF NUCLEAR MATERIALS | 2014年 / 454卷 / 1-3期

基金：

瑞典研究理事会;

关键词：

ATOM-PROBE TOMOGRAPHY; ZIRCONIUM ALLOYS; ZR ALLOYS; IRRADIATION; DIFFUSION; OXIDATION; EVOLUTION; METAL;

D O I：

10.1016/j.jnucmat.2014.07.072

中图分类号：

T [工业技术];

学科分类号：

08 ;

摘要：

An atom probe tomography study of the microstructure of a Zircaloy-2 material subjected to 9 annual cycles of BWR exposure has been conducted. Upon dissolution of secondary phase particles, Fe and Cr are seen to reprecipitate in large numbers of clusters and particles of 1-5 nm sizes throughout the Zr metal matrix. Fe and Sn were observed to segregate to ring-shaped features in the metal that are interpreted to be < c >-component vacancy loops. This implies that these two elements play a major role in the irradiation growth phenomenon in Zr alloys, which is believed to be caused by the formation of < c >-loops. Similarly to autoclave-corroded Zr alloys, the formation of a sub-oxide layer of approximate composition ZrO was observed. On the other hand, no oxygen saturated metal phase was detected underneath the oxide scale. (C) 2014 Elsevier B.V. All rights reserved.

引用

页码：178 / 185

页数：8

共 50 条

[21] ZIRCALOY-2 CORROSION IN STEAM-WATER MIXTURE UNDER REACTOR RADIATIONS
CERRAI, E
GADDA, F
SCARONI, A
ENERGIA NUCLEARE, 1971, 18 (11): : 607 - &
[22] PHASE INSTABILITY, DECOMPOSITION AND REDISTRIBUTION OF INTERMETALLIC PRECIPITATES IN ZIRCALOY-2 AND ZIRCALOY-4 DURING NEUTRON-IRRADIATION
GRIFFITHS, M
GILBERT, RW
CARPENTER, GJC
JOURNAL OF NUCLEAR MATERIALS, 1987, 150 (01) : 53 - 66
[23] ZIRCALOY-2 CORROSION IN STEAM-WATER MIXTURE UNDER REACTOR RADIATIONS
CERRAI, E
SCARONI, A
GADDA, F
CANADIAN METALLURGICAL QUARTERLY, 1972, 11 (01) : 21 - &
[24] METALLURGICAL PROPERTIES OF ZIRCALOY-2 RELEVANT TO ITS USE IN REACTOR PRESSURE TUBES
HURST, P
NICHOLS, RW
PICKLES, B
WATKINS, B
WOOD, DS
ATOMIC ENERGY REVIEW, 1976, 14 (03) : 439 - 520
[25] FABRICATION OF ZIRCALOY-2 CHANNELS FOR VALLECITOS BOILING WATER REACTOR FUEL ASSEMBLIES
SPALARIS, CN
NUCLEAR SCIENCE AND ENGINEERING, 1959, 6 (01) : 37 - 43
[26] STRESS-DEPENDENCE OF IN-REACTOR CREEP RATE OF HEAT-TREATED ZR-2.5 WT PERCENT NB AND COLD-WORKED ZIRCALOY-2
FIDLERIS, V
JOURNAL OF NUCLEAR MATERIALS, 1970, 36 (03) : 343 - &
[27] Hydrogen Uptake In Zircaloy-2 Reactor Fuel Claddings Studied With Elastic Recoil Detection
Rajasekhara, S.
Doyle, B. L.
Enos, D. G.
Clark, B. G.
APPLICATION OF ACCELERATORS IN RESEARCH AND INDUSTRY, 2013, 1525 : 270 - 275
[28] IN-REACTOR CREEP OF ZIRCALOY-2, ZIRCALOY-4 AND ZR-1.15 WT PERCENT CR-0.1 WT PERCENT FE AT 568 K DERIVED FROM THEIR STRESS-RELAXATION BEHAVIOR
CAUSEY, AR
JOURNAL OF NUCLEAR MATERIALS, 1974, 54 (01) : 64 - 72
[29] CREEP LIMIT APPROACH TO DESIGN OF ZIRCALOY-2 REACTOR PRESSURE TUBES AT 275 DEGREES C
WOOD, DS
WATKINS, B
JOURNAL OF NUCLEAR MATERIALS, 1971, 41 (03) : 327 - &
[30] OUT-OF-REACTOR STUDY ON EXTERNAL-PRESSURE CREEP OF ZIRCALOY-2 FUEL CLADDING TUBES
MAKI, H
HARA, T
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 1975, 12 (01) : 43 - 52

← 1 2 3 4 5 →