The Causes of Increased Hydrogen Uptake of Zirconium Based Fuel Claddings at High Burnup PDF Download
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Author: Jonathan E. D. Hawes
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Category :
Languages : en
Pages : 0
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Author: Jonathan E. D. Hawes
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Category :
Languages : en
Pages : 0
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Author: Adrienn Baris
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Category :
Languages : en
Pages : 0
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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: Holger Wiese
Publisher:
ISBN:
Category : High burnup
Languages : en
Pages : 34
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The objective of this paper is to summarize the results of the latest observations performed at Paul Scherrer Institut on irradiated fuel claddings, to characterize their corrosion and hydrogen-uptake behavior. Two categories of studies have been performed. (1) A series of destructive tests were achieved on the fuel rods irradiated in a boiling-water reactor (BWR), including hydrogen concentration by hot-gas extraction. These results provided the hydrogen content of the cladding at different stages of irradiation, at different elevations along the rod. (2) Another series of examinations using a correlative microscopy method, i.e., using different techniques, including transmission electron microscopy (TEM), electron probe microanalysis (EPMA), and secondary ion mass spectrometry (SIMS), on the same material and in the same region of the metal-oxide interface have provided useful data regarding the oxide layer combining the signals from oxides and from hydrides. Furthermore, the effect of the type of alloying element has been examined for in-reactor oxidation. These studies are subsequently combined with the findings from out-of-pile studies, using techniques, such as neutron radiography, to confirm the in-reactor observations. Results have shown that: (i) the hydrogen pickup fraction varies at different conditions and could even decrease as the oxide thickness increases; (ii) the distribution of hydrogen in the cladding is usually inhomogeneous; (iii) the most determining parameter for hydrogen uptake seems to be the microstructure of the oxide, and the nature of the alloying element will influence to a certain extent this parameter; (iv) furthermore, the stress in the oxide layer can modify the crack distribution in the latter, cracks will in turn shorten the route for the hydrogen to access the metal. These results will be discussed as a contribution to the available knowledge about hydrogen uptake and will provide a global support for the models of the uptake phenomenon.
Author: Pierre Yanis Bouhaddane
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Category :
Languages : en
Pages : 0
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Zirconium-based alloys have been used in nuclear reactors as fuel cladding and structural materials since the development of nuclear energy. Zircaloy-2, a Sn-Fe-Cr-Ni alloy was widely in service for years, and still is today in boiling water reactors (BWR). Among the many challenges the materials face during operation in the nuclear reactor, hydrogen pickup during corrosion of the metal components is of great concern due to the embrittlement properties of the zirconium hydrides. Zircaloy-2 materials show great corrosion resistance in the boiling environment but many in-pile fuel cladding and structural components, such as water rods and channel boxes, revealed accelerated ingress of hydrogen at high burnup when exposed for additional cycles in the reactor, while Zircaloy-4 components did not. The industry is driven toward increasing the fuel burnup in the reactors, as it reduces operation costs, and therefore it is necessary to prevent this effect from happening in modern alloys. Because the main difference between Zircaloy-2 and Zircaloy-4 is the removal of nickel replaced by additional iron in Zircaloy-4 - nickel was linked to increased hydrogen pickup as early as the 1960's - nickel was thought responsible for this acceleration of hydrogen pickup during the additional cycles in the reactors. In a previous study, metallic nickel was measured in the oxide layer near the metal interface of high hydrogen pickup Zry-2 water rods. In this work, additional materials were selected at low and high elevations in the Zircaloy-2 water rods corroded for 3 and 4 cycles in a BWR (Limerick-1) with low and high hydrogen pickup respectively; and were examined by microbeam X-ray absorption near-edge spectroscopy (XANES), microbeam X-ray diffraction (XRD), and scanning electron microscopy (SEM) in an effort to verify and understand further this observation. Cross-sectional samples were prepared from the two water rods and investigated at the Advance Photon Source (APS) at Argonne National Laboratory (ANL). In each material, the oxidation state of nickel atoms in the thick oxide layers was measured as a function of distance from the metal interface by XANES. The results confirm the presence of metallic nickel in the oxide layer of the high elevation material/high hydrogen pickup material (4 cycles) where 30-35% metallic nickel was seen in the near oxide (up to 10-12 [mu]m from the metal interface), as previously observed in two other high hydrogen pickup materials. At low elevation in the high hydrogen pickup water rod, the correlation was not directly verified (nickel atoms were fully oxidized in the oxide layer past 3-4 [mu]m from the metal/oxide interface) but we argue that the high hydrogen content observed at that location results from the diffusion down the water rod of hydrogen absorbed at higher elevation, driven by the concentration and temperature axial gradients. A detailed analysis of the XANES signal from the metallic nickel atoms in the oxide layer of the high hydrogen pickup material suggest that these nickel atoms are no longer bonded to zirconium atoms, which shows that the metallic nickel which can affect hydrogen pickup consists of atoms in solid solution or in small clusters in the oxide layer, rather than in second phase precipitates. This is in agreement with recent APT examinations of high burnup Zry-2 materials with high hydrogen pickup in which the nickel atoms were seen uniformly distributed in the oxide layer and only small clusters were observed. Additionally, metallic nickel in the outer oxide region close to the water interface was observed in most materials, with the highest metallic fraction (up to 75%) in the low hydrogen content samples. Nickel and iron high fluorescence counts near the oxide/water interface confirmed that the nickel atoms at that location corresponded to deposits from the corrosion of other reactor components on the water rod oxide surfaces. However, these metallic nickel atoms near the water interface of the thick oxide layers (>25 [mu]m) do not seem to affect the hydrogen uptake in the Zry-2 materials as they were mostly observed in the low hydrogen pickup samples. Many cracks (lateral and through thickness) were seen in the oxide layers of the materials with SEM imaging of the prepared samples, especially in the high hydrogen pickup water rod at high elevation. In all four materials investigated, the oxide layers were rather uniform, but extensive circumferential oxide thickness variations could be observed between different regions of the water rods. An increase in oxidation kinetics during the 4th cycle was seen at mid/high elevation, where the irradiation flux is the most intense, by comparing the oxide thicknesses of the 3-cycles to 4-cycles GNF water rods and was correlated to the presence of nickel in the oxide layer. As such, irradiation seems to play an important role in accelerating corrosion (as previously reported) and in stabilizing metallic nickel in the oxide layer (and in turn enhancing hydrogen pickup). Concurrently to the XANES examinations, X-ray diffraction patterns were collected in the oxide layers of the cross-section samples as a function of distance from the bulk metal in order to investigate the oxide microstructure (phase content, grain size, texture) of in-reactor Zry-2 materials at high burnup with low and high hydrogen pickup fraction. The oxide layers formed on the BWR Zry-2 water rod consisted of small and highly oriented monoclinic oxide grains, with a small fraction of tetragonal grains, maximum near the metal interface (3-6%). Grain growth was observed in all materials as the oxide thickens, especially at high elevation, with grain sizes at 17-20 nm near the bulk metal and 33-38 nm in the outer region. Additionally, small grains compose the oxide region near the metal/oxide interface of the high elevation/high hydrogen pickup material which is coherent with accelerated corrosion taking place during the 4th cycle. In all materials, an orientation relationship was apparent between the (111) m-ZrO_2 and the (101 ̅0) [alpha]-Zr crystal planes, and for a significant fraction of the oxide grains throughout the whole oxide layer, the (200) m-ZrO_2 direction is close to the oxide growth direction. This is coherent with previous XRD examinations of autoclave and in-reactor corroded Zr-alloys. After a thorough review of the presented results and of the literature available, the author proposed a mechanism for the enhancement of hydrogen uptake in Zry-2 materials in BWR at high burnup. A combination of a thick, porous oxide layer, of high fluence, of high irradiation flux and of low linear power -- especially for fuel rods -- are thought to be necessary conditions for the stability of metallic nickel in the near oxide layer of Zry-2 materials during additional cycles at high burnup. These metallic nickel atoms then catalyze the hydrogen absorption surface reaction at cracks and pores surfaces near the metal interface, as previously suggested, resulting in increased hydrogen pickup by the material. In turn, the results presented in this study support that the acceleration of hydrogen pickup observed in Zry-2 materials at high burnup in BWR is not likely to occur in the modern Ni-free Zr alloys.
Author: Aditya Shivprasad
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Hydrogen pick-up of zirconium-based fuel cladding and structural materials duringin-reactor corrosion can degrade fuel component performance in existing light waterreactors (LWRs) and advanced nuclear reactors, such as the LWR-like supercriticalwater reactors (SCWRs), as the ingress of corrosion hydrogen can lead to the formationof brittle hydrides. In the boiling water reactor (BWR) environment, Zircaloy-2 fuelcladding and reactor core components, such as water rods and channel boxes, canexperience accelerated hydrogen pick-up (higher pickup fraction) at high burnup whenexposed for one extra 24-month cycle, while Zircaloy-4 components under similarconditions do not. Because the principal difference between the two alloys is thatZircaloy-2 contains nickel, this accelerated hydrogen pick-up has been hypothesizedto result from the presence of nickel and its role in the corrosion process whenincorporated into the protective oxide layer.Zircaloy-2 and Zircaloy-4 sister samples were corroded in 360 _C water and anadditional set of Zircaloy-2 samples was corroded in 400 _C steam. Total weightgain, assumed to be due mostly to oxygen, and hydrogen content were measured asfunctions of exposure time. The results indicate that Zircaloy-2 samples absorbed morehydrogen than did Zircaloy-4 samples on the basis of total weight gain (hydrogen pickupfraction), though both exhibited similar corrosion kinetics parameters. Microbeamsynchrotron radiation X-ray absorption near-edge spectroscopy (XANES) of selectedZircaloy-2 samples at the Advanced Photon Source (APS) was used to probe theoxidation states of nickel and iron in these materials and understand the evolutionof the oxidation states of these alloying elements as functions of distance from theoxide/metal interface. Result showed that a significant fraction of nickel atomsremained metallic upon incorporation in the oxide layer. In contrast, iron atomsoxidized much earlier than did nickel atoms and, in most cases, fully oxidized withinseveral micrometers from the oxide/metal interface. A general hypothesis was madethat metallic nickel in contact with the coolant may catalyze the surface reactionsinvolved in the hydrogen pick-up mechanism.To understand accelerated hydrogen pick-up of certain Zircaloy-2 samples at highburn-up, additional XANES examinations were performed on Zircaloy-2 water rodsexposed in-reactor to high burn-up in commercial BWRs. The first set of samples wascorroded in the Limerick-1 reactor, while the second set was corroded in the Dresden-2reactor. Within each set of samples, fluences, oxide thicknesses, and sample elevationswere similar, but hydrogen pick-up fractions were vastly different. In the first setof samples, oxide thicknesses ranged from 28 - 35 m, but hydrogen pick-up rangedbetween 15 and 51%. In the second set of samples, oxide thicknesses ranged between3.5 m and 16 m, but hydrogen pick-up ranged from 28 - 69%. All samples wereirradiated to fluences between 9.4 and 13.1 1021 n/cm2 for neutron energies above1 MeV. Results of XANES examinations showed a similar correlation between thedelayed oxidation of nickel and higher hydrogen pick-up of Zircaloy-2 at high burn-up.A significant fraction (greater than 30%) of nickel atoms were found to be in themetallic state in the porous oxide layer. It was hypothesized that this metallic nickelis responsible for enhancing hydrogen pick-up by catalyzing the surface reactions thataffect the overall hydrogen pick-up reaction. This would allow for easier absorptionof hydrogen into the protective oxide layer from the coolant. Ab initio modeling ofXANES of selected iron- and nickel-containing compounds was also performed andcompared to experimental results to help understand how different populations ofalloying elements oxidized upon incorporation into the oxide layer.A concurrent study of the microstructure of oxide layers formed on these sameirradiated water rods was performed to understand if there was a characteristicmicrostructure associated with accelerated hydrogen pick-up. Microbeam X-raydiffraction (XRD) at the APS was performed on water rod samples to study oxidetexture, phase content, and grain size. A similar examination was performed onsteam-corroded Zircaloy-2 to serve as a comparison. Results showed that the oxidelayers formed on these samples consisted primarily of highly-oriented monoclinic phasezirconium oxide with a small fraction of tetragonal phase oxide. Monoclinic phasegrains were shown to grow as a function of distance from the oxide/metal interface,while tetragonal phase grains remained a constant size, indicating a tetragonal-to-monoclinic phase transformation above a critical grain size of approximately 10 nm.The tetragonal phase fraction was also calculated and observed to maximize nearthe oxide/metal interface, coinciding with the appearance of the (002)-tetragonalphase diffraction reflection, which appeared to be highly-oriented and strained, butdisappeared away from the oxide/metal interface. Findings were consistent withprevious microbeam XRD examinations of oxide layers formed on Zircaloy-4 underautoclave conditions. Transmission XRD examinations were also performed on aselected steam-corroded sample to serve as an additional comparison.The observations presented in this study helped to propose a mechanism foroxidation of different populations of iron and nickel upon incorporation into theZircaloy-2 oxide layer and the effect on the hydrogen pick-up mechanism.
Author: D. Schrire
Publisher:
ISBN:
Category : Grid
Languages : en
Pages : 31
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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:
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ISBN:
Category :
Languages : en
Pages : 12
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Author: Craig M. Eucken
Publisher: ASTM International
ISBN: 080311463X
Category : Nuclear fuel claddings
Languages : en
Pages : 794
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The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr
Author: Jennifer Anne Jarvis
Publisher:
ISBN:
Category :
Languages : en
Pages : 318
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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.
Author: SK. Yagnik
Publisher:
ISBN:
Category : Ductility
Languages : en
Pages : 28
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Zircaloy fuel cladding suffers progressive degradation of ductility as its neutron exposure and hydrogen uptake increase with burnup. The loss of ductility appears to be the key property governing the cladding integrity in service. We report ductility data of Zircaloy-4 fabricated in stress-relief annealed (SRA) and recrystallized (RXA) conditions, covering a range of fluence, hydrogen content, and irradiation and test temperature.
