Refinement of Irradiation and Analysis Techniques for Radiation-Induced Segregation

1996 ◽  
Vol 439 ◽  
Author(s):  
T. R. Allen ◽  
J. M. Cookson ◽  
D. L. Damcott ◽  
G. S. Was
Keyword(s):  
Grain Boundary ◽  
Reactor Core ◽  
Beam Current ◽  
304 Stainless Steel ◽  
Boundary Concentration ◽  
Ultra High Purity ◽  
Potential Contributor ◽  
Core Components ◽  
Radiation Induced ◽  
Boundary Chemistry

AbstractRadiation-induced segregation (RIS) has been implicated as a potential contributor to irradiation assisted stress corrosion cracking in light water reactor core components. To better understand changes to grain boundary chemistry during irradiation, RIS was measured in ultra-high purity (UHP) 304 stainless steel using Auger electron spectroscopy (AES). Variations in measured grain boundary concentration, both within a sample and between samples, are reduced by refinements in both the radiation and the AES techniques. These refinements include improvements in temperature control, uniformity of sample-to-sample dose, grain boundary acceptance criteria, amount of intergranular fracture, and amount of beam current used in analysis. AES measurements on samples irradiated at 400°C to 1.0 dpa show how implementing the technique refinements reduces the variability in the measured concentrations. Additionally, measurements from regions of ductile tearing in samples irradiated to 0.1 and 1.0 dpa at 400°C, to 1.0 dpa at 200°C, and from unirradiated samples show that sensitivity factors must be determined to obtain the most accurate measurement of grain boundary composition.

Proton Irradiation Induced Grain Boundary Segregation in Austenitic Stainless Steels

1994 ◽  
Vol 373 ◽  
Author(s):  
D.L. Damcott ◽  
G.S. Was ◽  
S.M. Bruemmer
Keyword(s):  
Grain Boundary ◽  
Proton Irradiation ◽  
Boundary Segregation ◽  
Reactor Core ◽  
Austenitic Stainless Steels ◽  
Sharp Decrease ◽  
Diffusion Parameters ◽  
Core Components ◽  
Radiation Induced ◽  
Higher Temperature

AbstractRadiation induced segregation (RIS) has been implicated as a mechanism for irradiationassisted stress corrosion cracking (IASCC) in reactor core components. Proton irradiation has been shown to be useful in creating grain boundary chemistries similar to those found in neutron and charged particle irradiated materials for accelerated testing of IASCC susceptibility. This work quantifies grain boundary RIS as a function of proton irradiation dose (0.1-3.0 dpa), temperature (200°−600°C), and alloy composition (20Cr-9Ni, 24Cr-19Ni, and xCr-24Ni, x=16, 20,24). Auger electron spectroscopy revealed Cr depletion and Ni enrichment under all irradiation conditions. As a function of dose, the degree of segregation increased rapidly to near saturation prior to 1 dpa, with a boundary composition of 12.1 at.% Cr and 36.0 at.% Ni at 1 dpa. Segregation peaked at approximately 500°C with 13.0 at.% Cr and 38.6 at.% Ni at the grain boundary at 0.5 dpa; very little segregation was observed at or below 300°C or at 600°C. The trends in segregation as a function of dose agreed well with the Perks' model predictions with the exception of the measurement at 600°C, which showed the sharp decrease in segregation predicted for a higher temperature (700°C-800°C). For alloys containing constant bulk Cr but varying Ni, the Perks' model agreed well with the observed segregation trend; however, for alloys containing constant bulk Ni and varying Cr, agreement was achieved only through the use of composition dependent diffusion parameters.

The Measurement of Light Element Segregation Using EDS and EELS

1998 ◽  
Vol 4 (S2) ◽  
pp. 772-773
Author(s):  
J.T. Busby ◽  
E.A. Kenik ◽  
G.S. Was
Keyword(s):  
Grain Boundaries ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Light Element ◽  
Free Surfaces ◽  
Spatial Redistribution ◽  
Core Components ◽  
Radiation Induced ◽  
Potential Interactions ◽  
Impurity Elements

Radiation-induced segregation (RIS) is the spatial redistribution of elements at defect sinks such as grain boundaries and free surfaces during irradiation. This phenomenon has been studied in a wide variety of alloys and has been linked to irradiation-assisted stress corrosion cracking (IASCC) of nuclear reactor core components. However, several recent studies have shown that Cr and Mo can be enriched to significant levels at grain boundaries prior to irradiation as a result of heat treatment. Segregation of this type may delay the onset of radiation-induced Cr depletion at the grain boundary, thus reducing IASCC susceptibility. Unfortunately, existing models of segregation phenomena do not correctly describe the physical processes and therefore are grossly inaccurate in predicting pre-existing segregation and subsequent redistribution during irradiation. Disagreement between existing models and measurement has been linked to potential interactions between the major alloying elements and lighter impurity elements such as S, P, and B.

scholarly journals The Effect of Bulk Composition on Swelling and Radiation-Induced Segregation in Austenitic Alloys

2000 ◽  
Vol 650 ◽  
Author(s):  
T. R. Allen ◽  
J. I. Cole ◽  
N. L. Dietz ◽  
Y. Wang ◽  
G. S. Was ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Grain Boundary ◽  
Bulk Composition ◽  
High Energy ◽  
Lattice Parameter ◽  
304 Stainless Steel ◽  
Microstructural Development ◽  
Nickel Concentration ◽  
Void Swelling ◽  
Radiation Induced

ABSTRACTChanges in bulk composition are known to affect both radiation-induced segregation and microstructural development, including void swelling in austenitic stainless steel. In this work, three alloys (designations corresponding to wt%) have been studied: Fe-18Cr-8Ni alloy (bulk composition corresponding to 304 stainless steel), Fe-18Cr-40Ni (bulk composition corresponding to 330 stainless steel), and Fe-16Cr-13Ni (bulk composition corresponding to 316 stainless steel). Following irradiation with high-energy protons, the change in hardness and microstructure (void size distribution and grain boundary composition) due to irradiation was investigated. Increasing the bulk nickel concentration decreases void swelling, increases matrix hardening, and increases grain boundary chromium depletion and nickel enrichment. The analysis shows that decreases in lattice parameter and shear modulus due to radiation- induced segregation (RIS) correlate with decreased void swelling and a decreased susceptibility to irradiation assisted stress corrosion cracking (IASCC). Traditional thinking on IASCC assumed RIS was a contributing factor to cracking. It may, however, be that properly controlled RIS can be used to mitigating cracking.

Radiation-Induced Segregation in 316 and 304 Stainless Steels Irradiated at Low Dose Rate

2000 ◽  
Vol 650 ◽  
Author(s):  
T. R. Allen ◽  
J. I. Cole ◽  
J. Ohta ◽  
K. Dohi ◽  
H. Kusanagi ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Grain Boundary ◽  
Dose Rate ◽  
Stainless Steels ◽  
Low Dose ◽  
Bulk Composition ◽  
304 Stainless Steel ◽  
316 Stainless Steel ◽  
Low Dose Rate ◽  
Radiation Induced

ABSTRACTAs part of the shutdown of the EBR-II reactor, structural materials were retrieved to analyze the effects of long-term irradiation on mechanical properties and microstructure. In this work, the effect of low dose rate irradiation (10−7 to 10−8 dpa/s) on grain boundary composition in 316 and 304 stainless steels was analyzed. Samples were taken from surveillance specimens and subassemblies irradiated in the reflector region of EBR-II at temperatures from 371-390°C to maximum doses of 30 dpa. The effects of dose, dose rate, and bulk composition on radiation- induced segregation are analyzed. In 316 stainless steel, changes in grain boundary chromium and nickel concentrations occur faster than changes in iron and molybdenum concentrations. In 304 stainless steel, decreasing the dose rate increases the amount of grain boundary segregation. For a dose of 20 dpa, chromium depletion and nickel enrichment are greater in 304 stainless steel than in 316 stainless steel, the difference most likely due to dose rate. In both 304 and 316 stainless steels, the presence of a grain boundary precipitate significantly changes the composition of the adjacent grain boundary.

Influence of Initial Grain Boundary Composition on the Evolution of Radiation-Induced Segregation Profiles

1998 ◽  
Vol 540 ◽  
Author(s):  
J.T. Busby ◽  
G.S. Was ◽  
S.M. Bruemmer ◽  
D. J. Edwards ◽  
E.A. Kenik
Keyword(s):  
Grain Boundary ◽  
Stainless Steels ◽  
Boundary Segregation ◽  
Reactor Core ◽  
Austenitic Stainless Steels ◽  
Central Peak ◽  
Boundary Composition ◽  
Radiation Induced ◽  
The Impact ◽  
Profile Development

AbstractRadiation-induced segregation (RIS) has been identified as a potential contributor to irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in reactor core components. The occurrence of grain boundary segregation prior to irradiation influences both the shape and magnitude of RIS profile development during subsequent irradiation. In an effort to better understand the impact of this pre-irradiation enrichment on RIS profile development, the evolution of grain boundary Cr segregation profiles with irradiation dose has been characterized. Commercial purity and high-purity austenitic stainless steels with different initial levels of grain boundary Cr have been irradiated with neutrons (at 275°C) or protons (at 360-400°C) to doses up to ∼5 dpa. Grain boundary composition profiles were measured before and after irradiation using scanning transmission electron microscopy with energy dispersive xray spectroscopy (STEM-EDS). The initial enrichment of Cr is shown to delay radiation-induced Cr depletion and produce a “W-shaped” profile at low irradiation doses. Further irradiation causes the central peak of the W to decrease, eventually resulting in the classical “V-shaped” depletion profile. Possible mechanisms for the pre-irradiation enrichment and its evolution into a “W-shaped” profile will be discussed.

The Effect of Low Dose Rate Radiation on the Microchemistry of 304 Stainless Steel

1998 ◽  
Vol 540 ◽  
Author(s):  
T. R. Allen ◽  
J. I. Cole ◽  
E. A. Kenik
Keyword(s):  
Stainless Steel ◽  
Grain Boundary ◽  
Dose Rate ◽  
Low Dose ◽  
304 Stainless Steel ◽  
Pressurized Water ◽  
Dose Rates ◽  
Low Dose Rate ◽  
Radiation Induced ◽  
Water Reactors

AbstractAs part of the shutdown of the EBR-II reactor, structural materials were retrieved to analyze the effect of long term, low dose rate irradiation. In this work, the effect of low dose rate (10 to 10−9 dpa/s) irradiation on grain boundary and void surface chemistry is analyzed. These dose rates are comparable to those in light water reactor structural components. The components were irradiated at 375-379°C, temperatures near the highest temperatures experienced in pressurized water reactors. Radiation-induced segregation (RIS) was measured on samples taken from 304 stainless steel hex ducts irradiated to doses between 10 and 12 dpa. Radiation-induced segregation is shown to vary with dose rate, with measured grain boundary chromium concentrations reaching as low as 5 at. % and nickel concentrations reaching as high as 33 at. %. For some radiation conditions, significant grain boundary precipitation occurs, possibly leaving components susceptible to environmental attack.

Phase Identification in Neutron Irradiated Stainless Steels

1980 ◽  
Vol 38 ◽  
pp. 382-383
Author(s):  
Eal H. Lee ◽  
A. F. Rowcliffe
Keyword(s):  
Reactor Core ◽  
Nucleation And Growth ◽  
Phase Identification ◽  
Void Swelling ◽  
Structure Sensitive ◽  
The Matrix ◽  
Aisi 316 Stainless Steel ◽  
Core Components ◽  
Radiation Induced ◽  
Aisi 316

Currently, AISI 316 stainless steel is widely used as a structural material for fast reactor core components. This material is susceptible to void swelling during irradiation at temperatures in the range 400 to 650°C. The nucleation and growth of radiation-induced voids are structure sensitive phenomena and are strongly affected by the nature and morphology of phases which develop during irradiation, the chemical composition of the matrix and the dislocation density. To understand the mechanisms of void swelling and to eventually develop materials with greater resistance to these phenomena, it is important to correctly determine the structure and composition of the phases which develop during irradiation of these complex alloys.

Point Defect Absorption and Dislocation Loop Formation at Grain Boundaries in Hvem-Irradiated Zr and Its Alloys:1. Influence of Solute Addition

1985 ◽  
Vol 43 ◽  
pp. 350-351
Author(s):  
R.A. Herring ◽  
M. Griffiths ◽  
M.H Loretto ◽  
R.E. Smallman
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Point Defect ◽  
Reactor Core ◽  
Solute Concentration ◽  
Nuclear Industry ◽  
Dislocation Loops ◽  
Loop Formation ◽  
Component Material ◽  
Impurity Interaction

Because Zr is used in the nuclear industry to sheath fuel and as structural component material within the reactor core, it is important to understand Zr's point defect properties. In the present work point defect-impurity interaction has been assessed by measuring the influence of grain boundaries on the width of the zone denuded of dislocation loops in a series of irradiated Zr alloys. Electropolished Zr and its alloys have been irradiated using an AEI EM7 HVEM at 1 MeV, ∼675 K and ∼10-6 torr vacuum pressure. During some HVEM irradiations it has been seen that there is a difference in the loop nucleation and growth behaviour adjacent to the grain boundary as compared with the mid-grain region. The width of the region influenced by the presence of the grain boundary should be a function of the irradiation temperature, dose rate, solute concentration and crystallographic orientation.

Effect of thermal annealing on grain boundary chemistry of polycrystalline TiBa2Ca2Cu3Oy films

1994 ◽  
Vol 52 ◽  
pp. 968-969
Author(s):  
Z. L. Wang ◽  
C. L. Briant ◽  
J. DeLuca ◽  
A. Goyal ◽  
D. M. Kroeger ◽  
...  
Keyword(s):  
Grain Boundary ◽  
Weak Link ◽  
Zero Field ◽  
Stabilized Zirconia ◽  
Local Texture ◽  
Transmission Electron ◽  
Axis Alignment ◽  
Boundary Chemistry ◽  
Small Misorientation ◽  
Misorientation Angles

Recent studies have shown that spray-pyrolyzed films of the Tl-1223 compound (TlxBa2Ca2Cu3Oy, with 0.7 < × < 0.95) on polycrystalline yttrium stabilized zirconia substrates can be prepared which have critical current density Jc near 105 A/cm2 at 77 K, in zero field. The films are polycrystalline, have excellent c-axis alignment, and show little evidence of weak-link behavior. Transmission electron microscopy (TEM) studies have shown that most grain boundaries have small misorientation angles. It has been found that the films have a nigh degree of local texture indicative of colonies of similarly oriented grains. It is believed that inter-colony conduction is enhanced by a percolative network of small angle boundaries at colony interfaces. It has also been found that Jc is increased by a factor of 4 - 5 after the films were annealed at 600 °C in oxygen. This study is thus carried out to determine the effect on grain boundary chemistry of the heat treatment.

Sign in / Sign up

Export Citation Format

Share Document