Oxidation, Embrittlement, and Growth of TREAT Zircaloy-3 Cladding PDF Download
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Author: Charles W. Solbrig
Publisher:
ISBN:
Category : Science
Languages : en
Pages :
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Book Description
This chapter analyzes the effects of oxidation, embrittlement, and cladding growth on the Zircaloy-3 alloy used for 25 mil thick TREAT fuel assembly cladding. The fuel cladding is a protective shell which is used to prevent damage to the enclosed fuel. Therefore, its integrity is important to guarantee this protection. The above three factors which can affect the Zircaloy-3 cladding are considered in this chapter and investigated. Limits to operation are determined. The oxidation of Zircaloy-3 in air is of interest to air-cooled reactors and Zircaloy-2 and 4 for accidents in fuel storage pools. The temperature range of interest is from room temperature where the fuel is stored for long periods of time, through the temperature range encountered in normal operation (400 to 600°C) to the highest temperatures which are possible in extreme accident situations. This temperature range is considered in this chapter to be from room temperature to 1200°C.
Author: Charles W. Solbrig
Publisher:
ISBN:
Category : Science
Languages : en
Pages :
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Book Description
This chapter analyzes the effects of oxidation, embrittlement, and cladding growth on the Zircaloy-3 alloy used for 25 mil thick TREAT fuel assembly cladding. The fuel cladding is a protective shell which is used to prevent damage to the enclosed fuel. Therefore, its integrity is important to guarantee this protection. The above three factors which can affect the Zircaloy-3 cladding are considered in this chapter and investigated. Limits to operation are determined. The oxidation of Zircaloy-3 in air is of interest to air-cooled reactors and Zircaloy-2 and 4 for accidents in fuel storage pools. The temperature range of interest is from room temperature where the fuel is stored for long periods of time, through the temperature range encountered in normal operation (400 to 600°C) to the highest temperatures which are possible in extreme accident situations. This temperature range is considered in this chapter to be from room temperature to 1200°C.
Author: Rehab Abdel Rahman
Publisher: BoD – Books on Demand
ISBN: 9535124471
Category : Science
Languages : en
Pages : 174
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Book Description
Assessing and improving nuclear material performance is a crucial subject for the sustainability of the nuclear energy and radioactive isotope supplies. This book aims to present research efforts used to identify nuclear materials performances in different areas. The contributions of esteemed international experts have covered important research aspects in fission and fusion technologies and naturally occurring radioactive materials management. The authors introduced current and anticipated trends toward better performances and mitigating challenges for commercial application of innovative technologies, biological remediation of mine effluents, nuclear fuel performance in power and research fission reactors, gamma ray spectrometer calibration, and recent advances in understanding the performance of tungsten composite in fusion reactor environment.
Author: Stephen L. Seiffert
Publisher:
ISBN:
Category : Light water reactors
Languages : en
Pages : 76
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Author: HM. Chung
Publisher:
ISBN:
Category : Deformation
Languages : en
Pages : 28
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Book Description
To establish the mechanical response of Zircaloy cladding under thermal shock conditions typical of hypothetical loss-of-coolant accident (LOCA) situations in light-water reactors (LWRs), cladding specimens were ruptured in steam during transient heating (10 K/s), oxidized at maximum temperatures between 1140 and 1770 K for various times, and cooled from the isothermal oxidation temperature to ~1100 K at a rate of 5 K/s, and rapidly quenched by bottom flooding with water at a rate of ~0.05 m/s. Failure "maps" for fracture of the cladding by thermal shock were developed relative to the maximum oxidation temperature and various time-dependent oxidation parameters. In situ pendulum-load impact tests were conducted at room temperature on tubes that survived the thermal quench. Information on the total absorbed energy from these tests was correlated with more extensive results from instrumented drop-weight impact tests. The thermal shock results indicate that the present Zircaloy embrittlement criterion (that is, a total oxidation limit of 17 percent of the wall thickness and a maximum cladding temperature of 1477 K) is conservative and that a more quantitative criterion, based upon the mechanical behavior of the oxidized material, can be formulated with a specified degree of conservatism consistent with the mechanical loads imposed on the cladding during reflood and the maximum amount of oxidation set by the margin of performance of emergency core-cooling systems in LWRs.
Author: Toshiyuki Yamashita
Publisher:
ISBN:
Category : Fission products
Languages : en
Pages : 64
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Author:
Publisher:
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 ...
Author: A. Sawatzky
Publisher:
ISBN:
Category : Embrittlement
Languages : en
Pages : 18
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Book Description
Under the condition postulated for a loss-of-coolant accident (LOCA), Zircaloy-4 fuel cladding may experience a temperature transient during which it absorbs an appreciable amount of oxygen from the coolant. Theoretical models for predicting cladding behavior during loss-of-coolant (LOC) are being developed, but until the failure mechanisms can be clearly established, an empirical criterion must be employed.
Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 1224
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![Embrittlement of Zircaloy Cladding Due to Oxygen Uptake (CBRTTL). [BWR; PWR]. PDF](https://books.google.com/books/content/images/frontcover/CLzTwQEACAAJ?fife=w80-h100)
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Category :
Languages : en
Pages :
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Book Description
A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes.
Author: Jordan Lee Vandegrift
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 122
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"The effect of sample geometry, welding strategies, atmosphere, plastic deformation, and rapid heating on the oxidation behavior of zirconium alloys has been investigated in this work. The goal of this work was to determine which zirconium alloy would be best suited as nuclear fuel cladding material in the Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL), which has unique operating conditions compared to typical reactors. TREAT is air-cooled, operates at high temperatures (400-600°C), and produces rapid transients (less than or equal to 700°C/s). Additionally, TREAT's cladding geometry is unique in that it has chamfers and welds. Alloying elements such as Fe, Sn, Cr, and Nb are typically added to zirconium and can drastically alter the corrosion properties of the material. The effects of such fabrication on the oxidation behavior of zirconium alloys is not well documented in literature and no direct comparison is provided for the alloys of interest, thus it is unclear how these alloys will behave under TREAT's conditions. Isothermal and non-isothermal oxidation studies were completed on pure Zr, Zircaloy-3 (Zry-3), Zircaloy-4 (Zry-4), Zr-1Nb, and Zr-2.5Nb plate specimens in both Ar+20%O2 and N2+20%O2 to study the effect of nitrogen on the oxidation behavior. Electron beam welded (EBW) and tungsten inert gas (TIG) welded Zry-3, Zry-4, and Zr-1Nb tube samples were oxidized under rapid heating and isothermal conditions in dry and humid N2+20%O2. Through these studies, the effect of chamfering, welding, humidity, and rapid heating were characterized. Macroscopic images of the samples were taken after oxidation, the oxide thickness was measured, and mass gain data was used to determine the oxidation rate constants and activation energies. It was found that Zry-3, Zry-4, and Zr-1Nb experience faster oxidation in N2+20%O2 than Ar+20%O2 at 800°C, while Zr and Zr-2.5Nb were relatively unaffected. Zry-3, Zry-4, and Zr-1Nb were found to experience accelerated oxidation in the weld region. Additionally, Zry-3 and Zry-4 experienced accelerated oxidation at the chamfers, while the chamfered region of Zr-1Nb experienced less oxidation. In all oxidation experiments, Zr-1Nb had the most favorable oxidation behavior."--Boise State University ScholarWorks.
