Oxide Surface Peeling of Advanced Zirconium Alloy Cladding After High Burnup Irradiation in Pressurized Water Reactors PDF Download
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Author: A. J. Mueller
Publisher:
ISBN:
Category : Ductility
Languages : en
Pages : 20
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Book Description
Microscopic examinations of advanced zirconium alloy cladding irradiated to burnups over 70 GWd/MTU have revealed de-lamination of surface layers of the thin oxide. The new observation is termed "oxide surface peeling" or OSP. Examinations have revealed the layered structure of the oxide. Metallographic examination revealed that the waterside oxide on different irradiated Zr alloy cladding had a layered structure similar to the autoclaved corrosion specimens examined earlier. However, the OSP observations discussed here apply only to irradiated cladding. OSP is not observed in autoclave corrosion. A featureless intact oxide sub-layer was present in the interior oxide at the metal/oxide interface for all alloys. On top of this featureless layer, there were additional sub-layers with fine circumferential fissures believed to be associated with the cyclic corrosion rate transitions. The number of sub-layers depended on the corrosion resistance of the alloy. Intermediate corrosion resistance alloy had many sub-layers forming an overall intermediate thickness oxide. Higher corrosion resistance alloys had fewer sub-layers on top of the barrier oxide layer. In some cases, small patches of the top surface layers of the thin oxide were peeled off because of radial cracks generated in the oxide layer caused by tensile stresses created by a hard pellet contact. Metallography of the underlying cladding showed that hydride localization was not associated with the oxide surface peeling; in contrast to previous experience on the low corrosion resistance older claddings, such as Zircaloy-4, where a major thickness fraction of the thick oxide extending to the underlying metal was removed ("spalled"). The oxide surface layer peeling does not lead to irradiated alloy ductility reduction or enhanced embrittlement. The impact of oxide surface peeling observations on fuel performance is discussed in the paper. For fuel designs with low margin against grid-to-rod fretting wear, OSP may reduce this margin further.
Author: A. J. Mueller
Publisher:
ISBN:
Category : Ductility
Languages : en
Pages : 20
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Book Description
Microscopic examinations of advanced zirconium alloy cladding irradiated to burnups over 70 GWd/MTU have revealed de-lamination of surface layers of the thin oxide. The new observation is termed "oxide surface peeling" or OSP. Examinations have revealed the layered structure of the oxide. Metallographic examination revealed that the waterside oxide on different irradiated Zr alloy cladding had a layered structure similar to the autoclaved corrosion specimens examined earlier. However, the OSP observations discussed here apply only to irradiated cladding. OSP is not observed in autoclave corrosion. A featureless intact oxide sub-layer was present in the interior oxide at the metal/oxide interface for all alloys. On top of this featureless layer, there were additional sub-layers with fine circumferential fissures believed to be associated with the cyclic corrosion rate transitions. The number of sub-layers depended on the corrosion resistance of the alloy. Intermediate corrosion resistance alloy had many sub-layers forming an overall intermediate thickness oxide. Higher corrosion resistance alloys had fewer sub-layers on top of the barrier oxide layer. In some cases, small patches of the top surface layers of the thin oxide were peeled off because of radial cracks generated in the oxide layer caused by tensile stresses created by a hard pellet contact. Metallography of the underlying cladding showed that hydride localization was not associated with the oxide surface peeling; in contrast to previous experience on the low corrosion resistance older claddings, such as Zircaloy-4, where a major thickness fraction of the thick oxide extending to the underlying metal was removed ("spalled"). The oxide surface layer peeling does not lead to irradiated alloy ductility reduction or enhanced embrittlement. The impact of oxide surface peeling observations on fuel performance is discussed in the paper. For fuel designs with low margin against grid-to-rod fretting wear, OSP may reduce this margin further.
Author: Robert Odette
Publisher: Newnes
ISBN: 012397349X
Category : Technology & Engineering
Languages : en
Pages : 673
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Book Description
High-performance alloys that can withstand operation in hazardous nuclear environments are critical to presentday in-service reactor support and maintenance and are foundational for reactor concepts of the future. With commercial nuclear energy vendors and operators facing the retirement of staff during the coming decades, much of the scholarly knowledge of nuclear materials pursuant to appropriate, impactful, and safe usage is at risk. Led by the multi-award winning editorial team of G. Robert Odette (UCSB) and Steven J. Zinkle (UTK/ORNL) and with contributions from leaders of each alloy discipline, Structural Alloys for Nuclear Energy Applications aids the next generation of researchers and industry staff developing and maintaining steels, nickel-base alloys, zirconium alloys, and other structural alloys in nuclear energy applications. This authoritative reference is a critical acquisition for institutions and individuals seeking state-of-the-art knowledge aided by the editors’ unique personal insight from decades of frontline research, engineering and management. Focuses on in-service irradiation, thermal, mechanical, and chemical performance capabilities. Covers the use of steels and other structural alloys in current fission technology, leading edge Generation-IV fission reactors, and future fusion power reactors. Provides a critical and comprehensive review of the state-of-the-art experimental knowledge base of reactor materials, for applications ranging from engineering safety and lifetime assessments to supporting the development of advanced computational models.
Author: Adrienn Baris
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Book Description
In light water reactors the fuel is encapsulated in Zr-based claddings that withstand the harsh environment (neutron bombardment, high temperature and water under pressure); without absorbing too many neutrons to sustain the chain reaction in the reactor core. Relatively high corrosion resistance of Zr is achieved when alloyed (e.g. with Sn, Fe, Cr, Ni, or Nb). Some elements form second phase particles (SPPs) and provide protection against rapid corrosion. The cladding undergoes compositional and microstructural changes, such as irradiation induced SPP dissolution. Zr oxidizes at the metal-oxide interface by diffusion of the oxidizing species through the oxide layer. Therefore, a protective inner barrier oxide is essential to prevent the metal from fast reaction with different species. Hydrogen is released as a by-product of the oxidation, and by the radiolysis of the coolant. If H enters the metal it precipitates as brittle Zr-hydrides degrading the cladding's mechanical properties. The H-uptake is a critical safety issue. Although, extensive literature is available on this topic, there are some aspects that need better understanding. Increasing H-uptake of certain cladding types at high burnups was reported. The causes are not yet fully understood. To better understand the causes of increased H-uptake at high burnups, an extremely high burnup cladding (9 cycle LK3/L Zircaloy-2) from boiling water reactor provided the basis of the study. The same type of cladding after different service times was examined revealing the compositional and microstructural evolution. Two types of cladding from pressurized water reactor with medium burnup were studied to separate the reactor- and alloy-specific parameters from the generic ones. FIB tomography was used for the 3D reconstructions of the microstructure; EPMA and ChemiSTEM for the micro- and nanometric chemical analysis. It is revealed that regardless of alloy- and reactor-type, crack-free oxide and the absence of large hydrides in the vicinity of the metal-oxide interface; undulated interface; and presence of SPPs are among the essential factors for the cladding's high performance. It is demonstrated that the oxidation of the hydrides at the metal-oxide interface induces crack formation in the oxide, reducing its protectiveness. High level of SPP dissolution, large hydride phases in the metal and high level of porosity in the oxide at the interface, straight metal-oxide interface, stoichiometric oxide, increased Ni concentration in the inner oxide, segregation of Fe, Ni, Sn and slightly Cr in the metal grain boundaries, Sn segregation at the interface oxide are identified as the causes of increased H-uptake of the LK3/L cladding at high burnups. Although all of these factors are present after 9 cycles, the cladding does not show extremely fast oxidation and H-uptake even beyond the designed service time.
Author: D. Pêcheur
Publisher:
ISBN:
Category : Corrosion modeling
Languages : en
Pages : 25
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Book Description
The previous CEA corrosion code COCHISE provided satisfactory simulations of in-reactor corrosion of the fuel cladding when used in its validity range. In contrast, it could lead to hazardous predictions if applied out of this range due to the strongly linked parameters mainly based on the analysis of French pressurized water reactor (PWR) data. To predict the oxidation kinetics for new operating conditions or new materials, the CEA and EDF decided to develop a new model, named CORCY, which is based on a more phenomenological approach and uses separate parameters deduced from analytical experiments. The aim of this paper is to present the new model for Zircaloy-4 in PWR. The phenomenological approach is described. It is based on out- and in-pile data. Typically, since (1) the oxidation kinetics of zirconium alloys in autoclave are periodic, and (2) the oxide films formed in autoclave, in out-of-pile loop, and in-reactor all exhibit periodic lateral cracks with a period similar to the oxide thickness to transition, the oxidation kinetics in CORCY are characterized by a cyclic repetition of semi-parabolic law. Each model parameter is detailed. They are deduced separately from (a) oxidation tests performed in autoclave on fresh alloys to determine their kinetics; (b) oxidation tests performed in the out-of-pile corrosion loops Corail and Reggae to quantify the effects of thermo-hydraulic conditions; (c) data provided by Testing Material Reactors (OSIRIS and Halden reactor) during isothermal oxidation to determine the effects of irradiation; and (d) oxidation tests performed on pre-hydrided alloys to take into account the accelerated corrosion phase occurring, in reactor, for Zircaloy-4 at high burn-up. After describing both the phenomenological approach and the different model parameters, a comparison of measured and calculated corrosion data from French PWRs at a burn-up up to 65 GWd/tU is provided.
Author: D. Schrire
Publisher:
ISBN:
Category : Grid
Languages : en
Pages : 31
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Book Description
Good structural performance of the fuel assembly during irradiation is an indispensable requirement. Extension of licensed burnups demands continuous improvements, and more precisely on the design and processing of components made of zirconium alloys. Experience feedback on the assembly behaviour is necessary and continuous surveillance of the assemblies' performance is maintained through on site inspections and post irradiation examinations (PIE). For that purpose, two research programs have recently been performed which included PIE on selected pressurised water reactor (PWR) assembly components made of ZIRLO. In the first program, a 15 by 15 fuel assembly irradiated for four annual cycles in Ringhals 2 NPP was selected for PIE. Samples extracted from grid strap vanes, guide thimble, and guide thimble end to top nozzle joints were subjected to visual examinations and characterizations such as oxide layer thickness, orientation, and distribution of hydride precipitates and hydrogen content. In the second program, as extension of the irradiated material evaluation of 17 by 17 lead test assemblies (LTA) irradiated in Vandellós II NPP, outer grid strap vanes were removed from a normal operation three-cycle assembly and from a four-cycle LTA and sent to the hot cell laboratory for destructive examinations. One objective of this work was to analyse the behaviour of skeleton key parts at the end of their irradiation life. Special attention was paid to the performance of the guide thimble end to top nozzle joint. Another objective was to study the effect of an additional irradiation cycle on the oxide thickness, hydride precipitates distribution, hydrogen concentration, and hydrogen pickup fraction of ZIRLO grids. Furthermore, an analysis of the oxidation and hydrogen uptake contribution on ZIRLO grids growth was performed. The hot cell examination results are presented and evaluated in the paper.
Author: T. L. Mackay
Publisher:
ISBN:
Category : Zirconium
Languages : en
Pages : 26
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Book Description
Author: Kimberly Colas
Publisher:
ISBN:
Category : Zirconium oxide
Languages : en
Pages : 30
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Book Description
The corrosion process (oxidation and hydriding) of zirconium alloy fuel cladding is one of the limiting factors on fuel rod lifetime, particularly for Zircaloy-4. The corrosion rate of this alloy shows indeed a great acceleration at high burnup in light water reactors (LWRs). Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. In this study, Zircaloy-4 samples underwent helium and proton ion irradiation up to 0.3 dpa, forming a uniform defect distribution up to 1 ?m deep. Both as-received and precorroded samples were irradiated to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples were corroded to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray microdiffraction and microfluorescence were used to follow the evolution of oxide crystallographic phases, texture, and stoichiometry both in the metal and in the oxide. In particular, the tetragonal oxide phase fraction, which has been known to play an important role in corrosion behavior, was mapped in both unirradiated and irradiated metals at the submicron scale and appeared to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest, were combined to fully characterize changes caused by irradiation in metal and oxide phases of both alloys.
Author: J. Paul Pemsler
Publisher:
ISBN:
Category : Steam
Languages : en
Pages : 26
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Author: B. Cox
Publisher:
ISBN:
Category : Irradiation effects
Languages : en
Pages : 21
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Book Description
The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ?55°C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO2) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries.
Author: Marc Tupin
Publisher:
ISBN:
Category : Corrosion
Languages : en
Pages : 41
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Book Description
The irradiation damage in the fuel cladding material is mainly caused by the neutron flux resulting from the fission reactions occurring in the fuel. From an experimental point of view, the neutrons have the disadvantage to activate materials by neutron capture rendering them difficult to handle. To avoid these constraints inherent in the handling of radioactive material, the radiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. A new experimental approach using ion irradiation was performed in the Microscopy and Irradiation Damage Studies Laboratory of the CEA in Saclay, with the aim to study more specifically the influence of the irradiation damages in the oxide on the corrosion rate of the zirconium alloys. This study was, moreover, focused on a particular distribution of defects in the oxide layer, basically, localised close to the metal/oxide interface. From the results of the irradiation of the metal/oxide interface, it was clearly shown that, whatever the incident ion, the irradiation of the internal interface results in a significant increase of the oxygen diffusion flux ratios between the most irradiated Zircaloy-4 and the unirradiated one, whereas that of the oxide formed on M5TM induces a big decrease of the oxygen diffusion flux in the film. These effects are less marked with helium ions compared to protons (M5TM is a trademark of AREVA NP registered in the United States and in other countries). Finally, the oxide irradiation impact on the oxygen diffusion through the layer could explain the corrosion acceleration factor observed on Zy4 during the first cycles of irradiation, but cannot alone explain observed corrosion accelerations under high burn-up conditions. The discussion on the oxide irradiation effects puts forward the probable role of the residual charge left by ion implantation.
