Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors PDF Author:
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ISBN:
Category :
Languages : en
Pages :

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Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are "deterministic" because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday's law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the ...

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Book Description
Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are "deterministic" because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday's law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the ...

Hydrogen Entry in Zircaloy-4 Fuel Cladding

Hydrogen Entry in Zircaloy-4 Fuel Cladding PDF Author: Jennifer Anne Jarvis
Publisher:
ISBN:
Category :
Languages : en
Pages : 318

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Book Description
Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.

Zircaloy Cladding Behavior During Irradiation Tests Under Power-Cooling-Mismatch Conditions

Zircaloy Cladding Behavior During Irradiation Tests Under Power-Cooling-Mismatch Conditions PDF Author: GW. Gibson
Publisher:
ISBN:
Category : Cladding
Languages : en
Pages : 27

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Book Description
The behavior of Zircaloy-clad fuel rods for light water reactors under off-normal operating conditions is being studied in the Power Burst Facility. Fuel rods from three irradiation experiments conducted under power-cooling-mismatch conditions have been examined destructively in remote handling facilities. The postirradiation examinations of these pressurized water reactor-type rods have included visual and photographic examination, physical dimensioning, and metallography.

Thermal-Hydraulics of Water Cooled Nuclear Reactors

Thermal-Hydraulics of Water Cooled Nuclear Reactors PDF Author: Francesco D'Auria
Publisher: Woodhead Publishing
ISBN: 0081006799
Category : Technology & Engineering
Languages : en
Pages : 1200

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Book Description
Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of current approaches to reactor thermal hydraulic analysis, also considering the importance of this discipline for the design and operation of safe and efficient water-cooled and moderated reactors. Part One presents the background to nuclear thermal hydraulics, starting with a historical perspective, defining key terms, and considering thermal hydraulics requirements in nuclear technology. Part Two addresses the principles of thermodynamics and relevant target phenomena in nuclear systems. Next, the book focuses on nuclear thermal hydraulics modeling, covering the key areas of heat transfer and pressure drops, then moving on to an introduction to SYS TH and computational fluid dynamics codes. The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis. This book is a valuable resource for academics, graduate students, and professionals studying the thermal hydraulic analysis of nuclear power plants and using SYS TH to demonstrate their safety and acceptability. Contains a systematic and comprehensive review of current approaches to the thermal-hydraulic analysis of water-cooled and moderated nuclear reactors Clearly presents the relationship between system level (top-down analysis) and component level phenomenology (bottom-up analysis) Provides a strong focus on nuclear system thermal hydraulic (SYS TH) codes Presents detailed coverage of the applications of thermal-hydraulics to demonstrate the safety and acceptability of nuclear power plants

Oxidation of Zirconium and Zirconium Alloys

Oxidation of Zirconium and Zirconium Alloys PDF Author:
Publisher:
ISBN:
Category : Oxidation
Languages : en
Pages : 48

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Book Description
The oxidation rate was found to be relatively insensitive to various types of surface preparations in the temperature range 400 to 700 deg C. No dependence of reaction rate on oxygen pressure was observed. The cubic rate law also was obeyed by foil specimens at 700 deg C; however, the rate constants were slightly larger than values obtained from parallelepiped samples.

Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors

Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors PDF Author: Francesco D'Auria
Publisher: Elsevier
ISBN: 0323856071
Category : Business & Economics
Languages : en
Pages : 932

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Book Description
Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 1, Foundations and Principles includes all new chapters which delve deeper into the topic, adding context and practical examples to help readers apply learnings to their own setting. Topics covered include experimental thermal-hydraulics and instrumentation, numerics, scaling and containment in thermal-hydraulics, as well as a title dedicated to good practices in verification and validation. This book will be a valuable reference for graduate and undergraduate students of nuclear or thermal engineering, as well as researchers in nuclear thermal-hydraulics and reactor technology, engineers working in simulation and modeling of nuclear reactors, and more. In addition, nuclear operators, code developers and safety engineers will also benefit from the practical guidance provided. Presents a comprehensive analysis on the connection between nuclear power and thermal hydraulics Includes end-of-chapter questions, quizzes and exercises to confirm understanding and provides solutions in an appendix Covers applicable nuclear reactor safety considerations and design technology throughout

Oxidation of Zirconium and Zirconium Alloys in Liquid Sodium

Oxidation of Zirconium and Zirconium Alloys in Liquid Sodium PDF Author: T. L. Mackay
Publisher:
ISBN:
Category : Zirconium
Languages : en
Pages : 26

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Zirconium in the Nuclear Industry

Zirconium in the Nuclear Industry PDF Author: Leo F. P. Van Swam
Publisher: ASTM International
ISBN: 0803111991
Category : Nuclear fuel claddings
Languages : en
Pages : 781

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Book Description


Scientific and Technical Aerospace Reports

Scientific and Technical Aerospace Reports PDF Author:
Publisher:
ISBN:
Category : Aeronautics
Languages : en
Pages : 1330

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Book Description


Zirconium in the Nuclear Industry

Zirconium in the Nuclear Industry PDF Author: R. B. Adamson
Publisher: ASTM International
ISBN: 0803109350
Category : Creep
Languages : en
Pages : 832

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