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.

The Defect Microstructure and Element Composition in Denuded Zones of Stainless Steels Irradiated in BN-350 Nuclear Reactor

MRS Advances ◽  
2016 ◽  
Vol 2 (21-22) ◽  
pp. 1209-1215 ◽  
Author(s):  
Oleg V. Rofman ◽  
Kira V. Tsay ◽  
Oleg P. Maksimkin
Keyword(s):  
Grain Boundaries ◽  
Stainless Steels ◽  
Nuclear Reactor ◽  
Polycrystalline Materials ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Radiation Induced ◽  
High Degree ◽  
Induced Defects

ABSTRACTIt is known that microstructure of metallic polycrystalline materials irradiated with neutrons is often characterized by a high degree of heterogeneity in distribution of radiation-induced defects. Depleted zones are located along grain boundaries and their width is not only determined by irradiation temperature and damage dose, but also by migration of point defects and dislocations integrity, that makes it more difficult to interpret experimental results of this phenomenon. At present, denuded zones are still objects for investigation as they influence both operation characteristics of reactor materials and their safe long-term storage. In this work, denuded zones in hexagonal ducts of spent fuel assemblies constructed from 0.08C-16Cr-11Ni-3Mo and 0.12C-18Cr-10Ni-Ti stainless steels from BN-350 fast nuclear reactor were investigated by TEM. There were determined some irradiation parameters affecting the development of denuded zones and their width; void size distributions in near-grain boundary regions are presented. There was shown redistribution of alloying elements at grain boundaries using Energy-dispersive X-ray spectroscopy (EDS).

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.

High Spatial Resolution X-ray Microanalysis of Radiation-Induced Segregation in Proton-Irradiated Stainless Steels

1999 ◽  
Vol 589 ◽  
Author(s):  
E.A. Kenik ◽  
J.T. Busby ◽  
G.S. Was
Keyword(s):  
Grain Boundaries ◽  
Analytical Electron Microscopy ◽  
Scanning Transmission Electron Microscope ◽  
Transmission Electron ◽  
Analytical Electron ◽  
Scanning Transmission ◽  
Spatial Redistribution ◽  
Radiation Induced ◽  
Composition Profiles ◽  
Compositional Changes

AbstractThe spatial redistribution of alloying elements and impurities near grain boundaries in several stainless steel alloys arising from non-equilibrium processes have been measured by analytical electron microscopy (AEM) in a field emission scanning transmission electron microscope. Radiation-induced segregation (RIS) has been shown to result in significant compositional changes at point defects sinks, such as grain boundaries. The influence of irradiation dose and temperature, alloy composition, prior heat treatment, and post-irradiation annealing on the grain boundary composition profiles have been investigated. Understanding the importance of these microchemical changes relative to the radiation-induced microstructural change in irradiation-assisted stress corrosion cracking (IASCC) of the irradiated materials is the primary goal of this study.

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.

Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Metin Yetisir ◽  
Rui Xu ◽  
Michel Gaudet ◽  
Mohammad Movassat ◽  
Holly Hamilton ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Operating Conditions ◽  
Stress Distributions ◽  
Light Water Reactors ◽  
Core Components ◽  
Water Reactors ◽  
Supercritical Water Cooled Reactor

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

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.

scholarly journals Neutron noise modelling for nuclear reactor diagnostics

2019 ◽  
Vol 26 ◽  
pp. 139
Author(s):  
A. Mylonakis ◽  
P. Vinai ◽  
C. Demaziére
Keyword(s):  
Diffusion Equation ◽  
Neutron Flux ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Reactor Core ◽  
Fine Mesh ◽  
Neutron Noise ◽  
Core Components ◽  
Numerical Tool ◽  
Modelling Approach

This paper presents the development of a neutron noise simulator for fine-mesh applications.The neutron noise of power nuclear reactors deals with the fluctuations of the neutron flux that are induced by fluctuations or oscillations of the reactor properties, i.e. displacement of core components, temperature or density variations, etc. Since the appearance of these perturbations can be problematic for the operation of nuclear reactors, it is desirable to be able to analyze their possible effects. The comparison between the modelling of such perturbations and possible measurements also gives the possibility to determine the driving perturbation in an operating nuclear reactor. One modelling approach is to solve numerically the neutron noise diffusion equation. This paper presents CORE SIM+, an under-development numerical tool oriented to neutron noise problems that require the fine-mesh spatial discretization of the reactor core.

In situ TEM studies of irradiation effects for materials improvement

1991 ◽  
Vol 49 ◽  
pp. 452-453
Author(s):  
Charles W. Allen
Keyword(s):  
Nuclear Reactor ◽  
Austenitic Stainless Steels ◽  
High Energy ◽  
Point Of View ◽  
In Situ Tem ◽  
Irradiation Effects ◽  
Tem Analysis ◽  
Radiation Induced ◽  
Analytical Tools

Irradiation effects studies employing TEMs as analytical tools have been conducted for almost as many years as materials people have done TEM, motivated largely by materials needs for nuclear reactor development. Such studies have focussed on the behavior both of nuclear fuels and of materials for other reactor components which are subjected to radiation-induced degradation. Especially in the 1950s and 60s, post-irradiation TEM analysis may have been coupled to in situ (in reactor or in pile) experiments (e.g., irradiation-induced creep experiments of austenitic stainless steels). Although necessary from a technological point of view, such experiments are difficult to instrument (measure strain dynamically, e.g.) and control (temperature, e.g.) and require months or even years to perform in a nuclear reactor or in a spallation neutron source. Consequently, methods were sought for simulation of neutroninduced radiation damage of materials, the simulations employing other forms of radiation; in the case of metals and alloys, high energy electrons and high energy ions.

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