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Michael Preuss - One of the best experts on this subject based on the ideXlab platform.
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characterisation of deuterium distributions in corroded Zirconium Alloys using high resolution sims
Social Science Research Network, 2020Co-Authors: Junliang Liu, Alistair Garner, Michael Preuss, James Sayers, Thomas Aarholt, Helen Hulme, Heidi Nordin, Jonna M Partezana, Magnus Limback, Sergio LozanoperezAbstract:Hydrogen diffusion through the oxide grown on Zr Alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr Alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb Alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at SPPs in the oxides on any of the Alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of Zirconium Alloys in water-cooled reactors.
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in situ digital image correlation for fracture analysis of oxides formed on Zirconium Alloys
Corrosion Science, 2016Co-Authors: Philip Platt, David Lunt, Efthymios Polatidis, M R Wenman, Michael PreussAbstract:Abstract Repeated breakdown of the protective oxide layer can be a key factor in the oxidation of Zirconium Alloys. Hence, accurate measurement of the oxide fracture strength is crucial for simulating the oxidation of these Alloys. High resolution digital image correlation was applied to SEM images taken during in-situ tensile loading of oxidised ZIRLO TM . Bi-axial strain measurements obtained during crack formation in the oxide films were converted into stress, and fitted to Weibull distributions to predict the oxide failure strength. Analysis highlights the impact of surface roughness. Samples polished prior to oxidation gave a fracture strength of ~1.6 GPa.
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the effect of sn concentration on oxide texture and microstructure formation in Zirconium Alloys
Acta Materialia, 2015Co-Authors: Alistair Garner, Philipp Frankel, Allan Harte, C R M Grovenor, Sergio Lozanoperez, Michael PreussAbstract:Abstract The development of oxide texture and microstructure formed on two Zirconium Alloys with differing Sn contents (Zr–1Nb–1Sn–0.1Fe, i.e. ZIRLO™ and Zr–1.0Nb–0.1Fe) has been investigated using transmission Kikuchi diffraction (TKD) in the scanning electron microscope (SEM) and automated crystal orientation mapping in the transmission electron microscope (TEM). Bulk texture measurements were also performed using electron backscatter diffraction (EBSD) in order to quantify and compare the oxide macrotexture development. The Sn-free alloy showed significantly improved corrosion performance by delay of the transition region and reduced levels of hydrogen pickup. The macroscopic texture and grain misorientation analysis of the oxide films showed that the improved corrosion performance and reduced hydrogen pick up can be correlated with increased oxide texture strength, the improved oxide grain alignment resulting in longer, more protective columnar grain growth. A lower tetragonal phase fraction is also observed in the Sn-free alloy. This results in less transformation to the stable monoclinic phase during oxide growth, which leads to reduced cracking and interconnected porosity and also to the formation of larger, well-aligned monoclinic grains. It is concluded that the Zr–1.0Nb–0.1Fe alloy is more resistant to hydrogen pickup due the formation of a denser oxide with a larger columnar grain structure.
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a study into the impact of interface roughness development on mechanical degradation of oxides formed on Zirconium Alloys
Journal of Nuclear Materials, 2015Co-Authors: Philip Platt, Philipp Frankel, S Wedge, Mhairi Gass, R Howells, Michael PreussAbstract:Abstract As a cladding material used to encapsulate nuclear fuel pellets, Zirconium Alloys are the primary barrier separating the fuel and a pressurised steam or lithiated water environment. Degradation mechanisms such as oxidation can be the limiting factor in the life-time of the fuel assembly. Key to controlling oxidation, and therefore allowing increased burn-up of fuel, is the development of a mechanistic understanding of the corrosion process. In an autoclave, the oxidation kinetics for Zirconium Alloys are typically cyclical, with periods of accelerated kinetics being observed in steps of ∼2 μm oxide growth. These periods of accelerated oxidation are immediately preceded by the development of a layer of lateral cracks near the metal-oxide interface, which may be associated with the development of interface roughness. The present work uses scanning electron microscopy to carry out a statistical analysis of changes in the metal-oxide interface roughness between three different Alloys at different stages of autoclave oxidation. The first two Alloys are Zircaloy-4 and ZIRLO ™ for which analysis is carried out at stages before, during and after first transition. The third alloy is an experimental low tin alloy, which under the same oxidation conditions and during the same time period does not appear to go through transition. Assessment of the metal-oxide interface roughness is primarily carried out based on the root mean square of the interface slope known as the R dq parameter. Results show clear trends with relation to transition points in the corrosion kinetics. Discussion is given to how this relates to the existing mechanistic understanding of the corrosion process, and the components required for possible future modelling approaches.
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a study into stress relaxation in oxides formed on Zirconium Alloys
Journal of Nuclear Materials, 2015Co-Authors: Philip Platt, Philipp Frankel, Mhairi Gass, R Howells, Efthymios Polatidis, Manuela Klaus, Michael PreussAbstract:Abstract Pressurised and boiling water reactors contain Zirconium Alloys, which are used as nuclear fuel cladding. Oxidation of these Alloys, and the associated hydrogen pick-up, is a limiting factor in the lifetime of the fuel. To extend the burn-up of nuclear fuel requires control of the oxidation, and therefore development of a mechanistic understanding of the cladding corrosion process. Synchrotron X-ray diffraction (S-XRD) has been used to analyse oxide layers formed during in-situ air oxidation of Zircaloy-4 and ZIRLO™. Analysis shows that as the oxide thickness increases over time there is a relaxation of the stresses present in both the monoclinic and meta-stable tetragonal phases, and a reduction in the tetragonal phase fraction. To better understand the mechanisms behind stress relaxation in the oxide layer, finite element analysis has been used to simulate mechanical aspects of the oxidation process. This simulation was first developed based on stress relaxation in oxides formed in autoclave, and analysed ex-situ using S-XRD. Relaxation mechanisms include creep and hydrogen-induced lattice strain in the metal substrate and creep in the oxide layer. Subsequently the finite element analysis has been extended to stress relaxation observed by in-situ S-XRD oxidation experiments. Finite element analysis indicates that the impact of creep in the oxide is negligible, and the impact of both creep and hydrogen-induced lattice strain in the metal substrate metal is small. The implication is that stress relaxation must result from another source such as the development of roughness at the metal–oxide interface, or fracture in the oxide layer.
Sergio Lozanoperez - One of the best experts on this subject based on the ideXlab platform.
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characterisation of deuterium distributions in corroded Zirconium Alloys using high resolution sims
Social Science Research Network, 2020Co-Authors: Junliang Liu, Alistair Garner, Michael Preuss, James Sayers, Thomas Aarholt, Helen Hulme, Heidi Nordin, Jonna M Partezana, Magnus Limback, Sergio LozanoperezAbstract:Hydrogen diffusion through the oxide grown on Zr Alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr Alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb Alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at SPPs in the oxides on any of the Alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of Zirconium Alloys in water-cooled reactors.
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the effect of sn concentration on oxide texture and microstructure formation in Zirconium Alloys
Acta Materialia, 2015Co-Authors: Alistair Garner, Philipp Frankel, Allan Harte, C R M Grovenor, Sergio Lozanoperez, Michael PreussAbstract:Abstract The development of oxide texture and microstructure formed on two Zirconium Alloys with differing Sn contents (Zr–1Nb–1Sn–0.1Fe, i.e. ZIRLO™ and Zr–1.0Nb–0.1Fe) has been investigated using transmission Kikuchi diffraction (TKD) in the scanning electron microscope (SEM) and automated crystal orientation mapping in the transmission electron microscope (TEM). Bulk texture measurements were also performed using electron backscatter diffraction (EBSD) in order to quantify and compare the oxide macrotexture development. The Sn-free alloy showed significantly improved corrosion performance by delay of the transition region and reduced levels of hydrogen pickup. The macroscopic texture and grain misorientation analysis of the oxide films showed that the improved corrosion performance and reduced hydrogen pick up can be correlated with increased oxide texture strength, the improved oxide grain alignment resulting in longer, more protective columnar grain growth. A lower tetragonal phase fraction is also observed in the Sn-free alloy. This results in less transformation to the stable monoclinic phase during oxide growth, which leads to reduced cracking and interconnected porosity and also to the formation of larger, well-aligned monoclinic grains. It is concluded that the Zr–1.0Nb–0.1Fe alloy is more resistant to hydrogen pickup due the formation of a denser oxide with a larger columnar grain structure.
Robert J Comstock - One of the best experts on this subject based on the ideXlab platform.
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corrosion of Zirconium Alloys used for nuclear fuel cladding
Annual Review of Materials Research, 2015Co-Authors: A T Motta, Adrien Couet, Robert J ComstockAbstract:During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In Zirconium Alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The Alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup incr...
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hydrogen pickup measurements in Zirconium Alloys relation to oxidation kinetics
Journal of Nuclear Materials, 2014Co-Authors: Adrien Couet, A T Motta, Robert J ComstockAbstract:Abstract The optimization of Zirconium-based Alloys used for nuclear fuel cladding aims to reduce hydrogen pickup during operation, and the associated cladding degradation. The present study focuses on precisely and accurately measuring hydrogen pickup fraction for a set of Alloys to specifically investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. To measure hydrogen concentrations in Zirconium Alloys two techniques have been used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results of both techniques show that hydrogen pickup fraction varies significantly with exposure time and between Alloys. A possible interpretation of the results is that hydrogen pickup results from the need to balance charge. That is, the pickup of hydrogen shows an inverse relationship to oxidation kinetics, indicating that, if transport of charged species is rate limiting, oxide transport properties such as oxide electronic conductivity play a key role in the hydrogen pickup mechanism. Alloying elements (either in solid solution or in precipitates) would therefore impact the hydrogen pickup fraction by affecting charge transport.
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cold neutron prompt gamma activation analysis a non destructive technique for hydrogen level assessment in Zirconium Alloys
Journal of Nuclear Materials, 2012Co-Authors: Adrien Couet, A T Motta, Robert J Comstock, Rick L PaulAbstract:Abstract We propose a novel use of a non-destructive technique to quantitatively assess hydrogen concentration in Zirconium Alloys. The technique, called Cold Neutron Prompt Gamma Activation Analysis (CNPGAA), is based on measuring prompt gamma rays following the absorption of cold neutrons, and comparing the rate of detection of characteristic hydrogen gamma rays to that of gamma rays from matrix atoms. Because the emission is prompt, this method has to be performed in close proximity to a neutron source such as the one at the National Institute of Technology (NIST) Center for Neutron Research. Determination shown here to be simple and accurate, matching the results given by usual destructive techniques such as Vacuum Hot Extraction (VHE), with a precision of ±2 mg kg −1 (or wt ppm). Very low levels of hydrogen (as low as 5 mg kg −1 (wt ppm)) can be detected. Also, it is demonstrated that CNPGAA can be applied sequentially on an individual corrosion coupon during autoclave testing, to measure a gradually increasing hydrogen concentration. Thus, this technique can replace destructive techniques performed on “sister” samples thereby reducing experimental uncertainties.
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transmission electron microscopy examination of oxide layers formed on zr Alloys
Journal of Nuclear Materials, 2006Co-Authors: Aylin Yilmazbayhan, Arthur T. Motta, E Breval, Robert J ComstockAbstract:Abstract A transmission electron microscopy investigation was performed on oxides formed on three Zirconium Alloys (Zircaloy-4, ZIRLO and Zr–2.5Nb) in pure water and lithiated water environments. This research is part of a systematic study of oxide microstructures using various techniques to explain differences in corrosion rates of different Zirconium Alloys. In this work, cross-sectional transmission electron microscopy was used to determine the morphology of the oxide layers (grain size and shape, oxide phases, texture, cracks, and incorporation of precipitates). These characteristics were found to vary with the alloy chemistry, the corrosion environment, and the distance from the oxide/metal interface. These are discussed and used in conjunction with observations from other techniques to derive a mechanism of oxide growth in Zirconium Alloys.
Junliang Liu - One of the best experts on this subject based on the ideXlab platform.
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characterisation of deuterium distributions in corroded Zirconium Alloys using high resolution sims
Social Science Research Network, 2020Co-Authors: Junliang Liu, Alistair Garner, Michael Preuss, James Sayers, Thomas Aarholt, Helen Hulme, Heidi Nordin, Jonna M Partezana, Magnus Limback, Sergio LozanoperezAbstract:Hydrogen diffusion through the oxide grown on Zr Alloys by aqueous corrosion processes plays a critical role in determining the rate of hydrogen pickup which can result in embrittlement of fuel cladding and limit the burnup of the nuclear fuel it encapsulates. Mapping the hydrogen/deuterium distributions in these oxide layers, especially in the barrier layer close to the metal/oxide interface, is a powerful way to understand the mechanism of both oxidation and hydrogen pickup. Here we have characterised by high-resolution SIMS analysis the deuterium distribution in oxide layers on a series of Zr Alloys, including autoclave-oxidised Zircaloy-4, Zr-1Nb and Zr-2.5Nb Alloys, and in-flux and out-of-flux corroded Zr-2.5Nb samples. Pre-transition Zircaloy-4 samples show a high deuterium trapping ratio in the oxide and a higher diffusion coefficient than in oxides on the Nb-containing samples. Neutron irradiation increases the deuterium diffusion coefficient, the deuterium concentration in the oxide and the pickup fraction in Zr-2.5 Nb samples. Comparative NanoSIMS and EDX/SEM analysis demonstrates that the deuterium is not preferentially trapped at SPPs in the oxides on any of the Alloys studied, but there is direct evidence for trapping at the surfaces of small oxide cracks especially in Zircaloy-4 samples. The high resolution mapping of these hot-spots in 3D can provide unique information on the mechanisms of hydrogen uptake, and suggests that the development of interconnected porosity in the oxide may be the critical rate-determining mechanism that controls HPU in the aqueous corrosion of Zirconium Alloys in water-cooled reactors.
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High resolution structural analysis of irradiated Zirconium Alloys
2019Co-Authors: Junliang LiuAbstract:This thesis is part of the MUZIC-3 (Mechanistic Understanding of Zirconium Corrosion) project, with the overall goal to understand the in-reactor corrosion and hydrogen pick-up mechanisms of Zirconium Alloys. Zirconium Alloys are commonly used for fuel cladding and support structures in light water reactors but suffer from aqueous corrosion in service that can limit the operating lifetime and the effective burnup of the fuel. Understanding the mechanisms of corrosion, especially under irradiation, is thus of great importance for the development of accurate, physically based lifetime prediction models for cladding materials, and providing information for the design of new corrosion-resistant Alloys. Acceleration of the corrosion rate has been widely reported as a result of exposure to in-reactor conditions, but the mechanisms of this accelerating effect are still not well understood. It has been correlated with the specific in-reactor coolant chemistry, intense γ-radiation or irradiation-induced redistribution of alloying elements, but rather little attention has been paid to the irradiation-induced degradation of the corrosion protective oxide layers. In an effort to understand such mechanisms, I have used both in-situ radiation damage techniques and direct observations of ex-reactor materials to study radiation effects in Zirconium oxides formed on a range of Zr Alloys. The aqueous corrosion of Zirconium follows the pathway: Zrâh-ZrOâZrO2. The transformation from Zr to h-ZrO is found to follow a displacive mechanism, and the suboxide grains can show different morphologies, equiaxed, plate-like or sawtooth-like, depending on the underlying α-Zr grain orientation, although no specific orientation relationships between h-ZrO and m-ZrO2 or α-Zr and m-ZrO2 were identified. The monoclinic-ZrO2 oxides formed in an autoclave are observed to have a similar texture on different Alloys, with (10Ì4)m-ZrO2 parallel to the metal/oxide interface regardless of the underlying α-Zr or h-ZrO orientations. I report for the first time on the susceptibility to radiation damage of the suboxide phase which may influence the nucleation of new oxide grains and the transportation of oxidation species across the oxide/metal interface, and lead to enhanced corrosion rates. A monoclinic-to-cubic transformation of the bulk oxide is also observed by in-situ ion irradiation experiments, followed by irradiation-induced grain growth. The possibility of radiation-induced stabilisation of this cubic phase thus needs to be considered as a possible process that can occur at high burnups in reactors, and may further affect the corrosion rates. As a result of the in-reactor corrosion conditions, e.g. irradiation and water chemistries, the oxides formed in-flux are less well textured and with a more disrupted grain, which can contribute to enhanced corrosion rates in-flux. My results also indicate that oxide nano-porosity plays an important role in the transportation of oxidising species throughout the oxide layer. The density of nano-porosity in the oxide corrosion layers has been successfully quantified for the first time as function of both depth in the oxide and exposure time, and a clear correlation is observed between the measured increase in hydrogen pickup fraction and the characterised increase in interconnected porosity in the oxide. To complete the work in this thesis, a wide range of microstructural characterization techniques have been applied, like Focused Ion Beam, Transmission Electron Microscopy, Transmission Kikuchi Diffraction and high-resolution Secondary Ion Mass Spectrometry. The combination of these techniques has provided me with a wide range of information on microstructures and microchemistry, which is useful to the understanding of the degradation mechanisms in fuel cladding materials.
Adrien Couet - One of the best experts on this subject based on the ideXlab platform.
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the coupled current charge compensation model for Zirconium alloy fuel cladding oxidation i parabolic oxidation of Zirconium Alloys
Corrosion Science, 2015Co-Authors: Adrien Couet, Arthur T. Motta, Antoine AmbardAbstract:Abstract A first principles oxidation model has been developed which can rationalize the kinetics observed in various Zirconium Alloys using a unified theoretical approach, which predicts that space charges in the oxide have a major impact on oxidation kinetics. As a first development, the parabolic oxidation of Zr–0.4Nb alloy is fitted by the model assuming that space charges in the oxide are compensated by Nb ions. The model quantitatively reproduces the oxidation kinetics using physically significant parameters. XANES examinations show that there are enough aliovalent Nb ions in the oxide layer to verify oxide electroneutrality as predicted by the model.
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corrosion of Zirconium Alloys used for nuclear fuel cladding
Annual Review of Materials Research, 2015Co-Authors: A T Motta, Adrien Couet, Robert J ComstockAbstract:During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In Zirconium Alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The Alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup incr...
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hydrogen pickup measurements in Zirconium Alloys relation to oxidation kinetics
Journal of Nuclear Materials, 2014Co-Authors: Adrien Couet, A T Motta, Robert J ComstockAbstract:Abstract The optimization of Zirconium-based Alloys used for nuclear fuel cladding aims to reduce hydrogen pickup during operation, and the associated cladding degradation. The present study focuses on precisely and accurately measuring hydrogen pickup fraction for a set of Alloys to specifically investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. To measure hydrogen concentrations in Zirconium Alloys two techniques have been used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results of both techniques show that hydrogen pickup fraction varies significantly with exposure time and between Alloys. A possible interpretation of the results is that hydrogen pickup results from the need to balance charge. That is, the pickup of hydrogen shows an inverse relationship to oxidation kinetics, indicating that, if transport of charged species is rate limiting, oxide transport properties such as oxide electronic conductivity play a key role in the hydrogen pickup mechanism. Alloying elements (either in solid solution or in precipitates) would therefore impact the hydrogen pickup fraction by affecting charge transport.
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cold neutron prompt gamma activation analysis a non destructive technique for hydrogen level assessment in Zirconium Alloys
Journal of Nuclear Materials, 2012Co-Authors: Adrien Couet, A T Motta, Robert J Comstock, Rick L PaulAbstract:Abstract We propose a novel use of a non-destructive technique to quantitatively assess hydrogen concentration in Zirconium Alloys. The technique, called Cold Neutron Prompt Gamma Activation Analysis (CNPGAA), is based on measuring prompt gamma rays following the absorption of cold neutrons, and comparing the rate of detection of characteristic hydrogen gamma rays to that of gamma rays from matrix atoms. Because the emission is prompt, this method has to be performed in close proximity to a neutron source such as the one at the National Institute of Technology (NIST) Center for Neutron Research. Determination shown here to be simple and accurate, matching the results given by usual destructive techniques such as Vacuum Hot Extraction (VHE), with a precision of ±2 mg kg −1 (or wt ppm). Very low levels of hydrogen (as low as 5 mg kg −1 (wt ppm)) can be detected. Also, it is demonstrated that CNPGAA can be applied sequentially on an individual corrosion coupon during autoclave testing, to measure a gradually increasing hydrogen concentration. Thus, this technique can replace destructive techniques performed on “sister” samples thereby reducing experimental uncertainties.
