High hydrogen pickup (HPU) and large scatter on HPU at high burnup/high exposure times is observed for Zircaloy-2 (Zr-2), which is used as fuel cladding, fuel channels, water rods, and spacer materials in boiling water reactors (BWRs). However, Zircaloy-4 (Zr-4) does not show this behavior in similar BWR environments. Because the main difference between Zr-2 and Zr-4 is that Zr-2 contains nickel as an alloying element, an investigation was pursued to elucidate the role of nickel. In this study, several Zr-2 and Zr-4 BWR fuel channel materials with low and high HPU exposed to known operating conditions, as well as a set of Zr-2 water rod materials irradiated in commercial BWRs in various exposures presenting higher HPU later in life, were examined using scanning electron microscopy, transmission electron microscopy (TEM), and 3D atom probe (3DAP) to determine the microstructural differences in the oxide layers and underlying metal that may lead to increased HPU, specifically focusing on where the nickel and iron are in the metal and in the oxide layers to explain the role of nickel on enhanced HPU in Zr-2 at high residence times. TEM analyses showed that nickel-containing secondary-phase particles (SPPs) dissolve as a function of fast fluence and that nickel dissolves slower than iron in irradiated Zr-2. The 3DAP analyses showed increased nickel content in solid solution in the alloy matrix, but there was no difference in the measured nickel content between low and high HPU fraction materials. Alloying element segregation was observed at the metal grain boundaries in TEM and 3DAP. 3DAP results showed some nickel and iron clustering in the irradiated metal and occasionally in the oxide layer. However, nickel segregation was not observed unequivocally in the oxide layers either using TEM or 3DAP in this study.

Nuclear fuel has made great strides from the advent of the nuclear industry at Shippingport to the on-going development of Accident Tolerant Fuel (ATF) for potential implementation in commercial nuclear plants. The development of Zircaloy-2 fuel with UO2 pellets at Shippingport was revolutionary. Incremental advancements have been made from Zircaloy-2 cladding to Zircaloy-4, and more advanced cladding materials for PWRs and BWRs to significantly improve fuel rod performance over the years. Today, significant steps are being made to further improve nuclear safety by developing cladding materials that could possibly survive severe accidents similar to those that occurred at Three Mile Island Unit 2 and the Fukushima Daiichi Nuclear Power Plant, where release of fission products and hydrogen can be significantly reduced, if not eliminated. This could make existing plants more passively safe in the event of a severe accident. The implementation of ATF cladding, together with advanced pellet materials, will improve safety and can also produce significant economic benefits for the fuel cycle and plants.

Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350 degrees C and Zircaloy-2 up to 14.7 x 10(25) n. m(-2), approximately 24 dpa, in a boiling water reactor at approximately 300 degrees C. Baseline transmission electron microscopy showed that the Zr3Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr(Fe,Cr)(2) and Zr2(Fe,Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze(3)Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher < a >-loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nanoclusters. A delay in the onset of < c >-loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of < a > and < c > loops is described. The implication this may have on irradiation induced growth of zirconium fuel cladding is also discussed.

Irradiation-induced structural changes of alpha-zirconium alloys and in particular the effect of iron were investigated by molecular dynamics simulations using embedded atom potentials derived from first-principles calculations. The simulations revealed that at temperatures between 500 and 600 K self interstitial atoms (SlAs) diffuse rapidly in a cooperative movement, preferably parallel to basal planes (a directions; < a >), forming nanoclusters with an extension in < a > and < c >. Vacancies diffuse more slowly than SIAs and remain isolated for a longer period of time. Nanoclusters associated with SIAs cause a pronounced overall expansion in a directions, as well as local strains. Under compressive strain in the c direction, vacancy diffusivity increases in the c direction. In contrast, the diffusivity of SIAs increases in the c direction under a tensile strain in the c direction. SIA nanoclusters are highly mobile within basal planes. Vacancy clusters grow by merging, leading to a contraction in the a direction, compensating for the expansion caused by SIA nanoclusters and possibly contributing to the plateau in growth after the initial rapid expansion. At the onset of breakaway growth, possibly due to stress buildup, the vacancy nanoclusters can condense into c loops, thereby diminishing the compensation effect. The alloying elements iron, nickel, chromium, and niobium liberated from secondary phase particles under irradiation or already in solution are attracted to vacancies and SIAs and are found inside vacancy and SIA loops. The interaction of alloying elements with defect clusters is discussed, with a particular focus on iron. Iron has been found to promote cluster formation in zirconium, and the structures of zirconium-iron clusters have been analyzed. Tin is repelled by SIA clusters and only weakly attracted by vacancies. Niobium impedes the diffusion of SIAs (and therefore may increase annihilation rates with nearby vacancies) and does not destabilize vacancy or SIA clusters. Ab initio calculations of the dimensional and elastic coefficients of the intermetallic phases occurring in secondary phase particles, such as Zr2Fe and Zr3Fe, are presented, allowing an assessment of local strains in a zirconium matrix. Thus, novel results from extended molecular dynamics simulations provide new insights and contribute to a deeper understanding of the complex mechanisms causing irradiation-induced dimensional changes and the breakaway growth of zirconium alloys.

In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components especially at high exposures cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multiyear (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2 x 10(26) m(-2) (E > 1 MeV) equivalent to maximum of approximately 37 dpa exposure followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320 +/- 10 degrees C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM). All irradiation samples (typically flat rectangular coupons or curvilinear cutouts of cladding tubes sized 35 by 6.5 by 0.8 mm) were prefilmed to avoid the uptake of impurity hydrogen from sodium-cooled BOR-60. A wide variety of samples representing standard LWR cladding alloys with and without prehydriding (approximately 116 to approximately 718 ppm) as well as special compositions with iron contents (100 to 4,000 ppm) were irradiated. The irradiation in BOR-60 was done in five different stages (eight microcycles) and lasted approximately 18 months with interim and final growth measurements made using a high precision-length measurement device. Results of the extensive investigation include: significant effects of Fe, Nb, and hydrogen additions; quantification of growth rates from low to very high fluences (dpas); measurement of volume changes; and correlation of growth with < c >-component dislocation densities.

This paper provides the results of investigations by transmission electron microscopy (TEM) on the selected materials from in-reactor oxidation tests in the Halden test reactor (Reference No. IFA-638) from 1998 to 2006. The objective of the IFA-638 test was to study the corrosion behavior of modern zirconium-based claddings to high burnup in pressurized water reactor water chemistry and thermal hydraulic conditions. The aim of this paper is to report on the microstructure of selected materials (ZIRLO (R), E635, and Alloy A) after the irradiation to different burnup levels to determine the modifications induced by irradiation and to correlate results to their oxidation behavior. The TEM examinations revealed the nature of secondary phase particles (SPPs) and their modification under irradiation. Four types of SPPs were observed, namely beta-niobium precipitates, Zr0.5Nb0.3Fe0.(2) (mainly in the ZIRLO alloy), Zr(Fe,Nb)(2) (in E635), and (Cr,Fe)(2)Zr,Nb with varying niobium content (present in Alloy A: Zr-0.58Sn-0.31Nb-0.36Fe-0.26Cr). TEM observations showed that all three materials contained still several precipitates after irradiation and in the case of the ZIRLO alloy even after high burnups. Furthermore, the analysis of the metal side of the interface and its comparison with the oxide side led to the conclusion that all types of precipitates dissolved to some extent under irradiation and that their alloying element content decreased. The dissolution was intensified in the oxide. However, a more detailed examination showed that the beta-niobium precipitates dissolved at a slower rate, or knowing that their composition was much richer in niobium, the time needed for the precipitates to become fully depleted from niobium was longer. Regarding the amorphization under irradiation, the beta-niobium- and chromium-containing precipitates did not amorphize in the metal part of the interface. This was not the case for the other types of precipitates. Furthermore, these two types of SPP both showed delayed oxidation and due to this behavior the typical crack above the SPP in the oxide was also observed. These results are discussed to gain an improved understanding of the oxidation behavior of materials studied as a function of irradiation and residence time.

We investigate nano-scale irradiation-induced precipitation in a Zr-Sn-Fe-Cr-Ni-alloy (Zircaloy-2) by combining atom probe tomography (APT) for chemical detail with scanning transmission electron microscopy (STEM) and high resolution energy dispersive X-ray (EDX) spectroscopy for wider context and complimentary and correlative TEM diffraction techniques for crystallographic relationships. We find that Fe- and Cr-rich nano-rods precipitate in Zircaloy-2 following proton irradiation at 350 degrees C to a low dose of similar to 2 dpa. The long-axis of the nano-rods are aligned in a direction 12-15 degrees from the Zr matrix < 0001 >, align in the basal plane and are of width 1.5-5 nm. Smaller rods are of APT-determined composition Zr-4(Fe0.67Cr0.33), tending towards Zr-3(Fe0.69Cr0.31) as the rod volume increases to > similar to 400 nm(3), in agreement with STEM-EDX determination of composition resembling that of Zr3Fe with Cr replacing some of the Fe. The Fe/Cr ratio has been shown to increase with distance from the nearest partially-dissolved Zr(Fe,Cr)(2) phase particle. The nucleation of nano rods has implications for macroscopic irradiation-induced deformation phenomena, irradiation-induced hardening and the evolution of dislocation loops and other defects. (C) 2018 Published by Elsevier B.V.

Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr)2 and Zr2(Fe,Ni). This is accomplished through ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr)2, predominantly from the edge region, and homogeneously in the case of Zr2(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr2(Fe,Ni) SPP with respect to the Zr(Fe,Cr)2. As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed.

Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4–15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys.

Advancements in transmission electron microscopy allow us to draw correlations between evolving matrix chemistry environments and the resulting dislocation structures that form. Such phenomena are essential in predicting the lifetime of neutron reactor components, but are not well understood at the fundamental level. We investigate the effect of nano-scale matrix chemical evolution in Zircaloy-2 on dislocation formation after emulating commercial reactor irradiation conditions on a proton beamline. Similarity in the dislocation type, morphology, density and evolution between the different irradiation types establishes proton irradiation in this regard. For the first time, we observe chemical segregation of Fe, Ni and Cr to a-loop positions in basal traces and the segregation of Sn in alternate rows, anticorrelated to the positions of the light transition elements. The resulting layered structure with a periodicity of similar to 50 nm creates an even greater anisotropy than that usually associated with HCP materials. Concurrent analysis of chemical effects and dislocation spatial relationships provides evidence that may explain the delayed onset of c-loop nucleation and accelerated dimensional instability regimes in its dependence on the alignment of a-loops parallel to the trace of the basal plane. This demonstrates the applicability of chemical-structural correlations towards key research questions regarding deformation behaviour. (C) 2017 Acta Materialia Inc. Published by Elsevier Ltd.

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

This chapter discusses the Westinghouse Electric Company LLC's accident tolerant fuel (ATF) program that utilizes Cr coated zirconium alloy (CZA) cladding with U3Si2 high density or high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. Westinghouse has performed corrosion testing using the autoclave facility at the Churchill, PA site to screen various coatings and SiC preparation methods for corrosion resistance. Initial autoclave and reactor testing indicated relatively high levels of SiC corrosion. Autoclave testing with hydrogen peroxide was used to simulate more aggressive oxidation conditions of the reactor and to explore coolant conditions that would minimize SiC corrosion rates. U3Si2 was tested for air and steam oxidation as compared with UO2 using digital scanning calorimeters at both the Westinghouse Columbia facility and at Los Alamos National Laboratory (LANL).

A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel.In this work, the tomographic device has been experimentally benchmarked for non-destructive rod wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of Kr-85 in the gas plenum to Cs-137 in the fuel stack. The rod-wise ratio of Kr-85/Cs-137 was, in turn, used to calculate the rod wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method.Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO(2), and four rods were from an assembly with a burnup of 26 MWd/kgUO(2). At the time of the measurements, the nine rods had cooled for approximately 22 years. All fuel rods had operated at high linear heat rates (around 70 kW/m), thus leading to relatively high FGR fractions. Here, the FGR fraction was determined to be similar to 24% in the high-burnup rods, and similar to 17% in the low-burnup rods. The tomography measurement results were in good agreement with the results from individual rod scanning, demonstrating the feasibility of tomography for this application. The capability of tomography to assess individual fuel rods without the need to dismantle the assembly can be particularly valuable in cases of fuels that do not allow disassembly, such as experimental HBWR fuel fitted with extensive instrumentation. (C) 2016 Elsevier B.V. All rights reserved.

The desire to improve the corrosion resistance of Zr cladding material to allow high burnup has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. Whereas the detrimental effect of Sn on high-temperature aqueous corrosion performance is widely accepted, the reason for it remains unclear. High-energy synchrotron x-ray diffraction was used to characterise the oxides formed by autoclave exposure on Zr-Sn-Nb alloys with tin concentrations ranging from 0.01 to 0.92 wt. %. The alloys studied included the commercial alloy ZIRLO and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy during autoclave testing at 360 degrees C was investigated by cross-sectional scanning and transmission electron microscopy (SEM and TEM). Non-destructive synchrotron x-ray diffraction analysis on the oxides revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on both phase and alloy. Additional in situ synchrotron x-ray diffraction experiments during oxidation at 550 degrees C provided further confirmation of the trends seen for autoclave-tested samples and demonstrated the presence of elevated levels of tetragonal phase in the initial stages of oxidation. In situ and ex situ measurements demonstrate unambiguously that the amount of tetragonal phase present and, more importantly, the degree of transformation from tetragonal to monoclinic oxide both decrease with decreasing tin levels, suggesting that tin stabilises the tetragonal phase. It is proposed that in Zr-Nb-Sn alloys with low Sn, the tetragonal phase is mainly stabilised by very small grain size and, therefore, remains stable throughout the corrosion process. By contrast, in alloys with higher tin levels, larger, stress stabilised, tetragonal grains can form initially, but then transform as the corrosion front progresses inward and stresses in the existing oxide relax.

Hydrogen-assisted irradiation growth may result in significant channel bow in addition to regular fluence gradient-induced bow in boiling-water reactor (BWR) fuel channels, especially at high exposures through "shadow corrosion," if hydrogen is picked up early in channel sides facing a control rod. This phenomenon may be responsible for recent high channel bow observations. To develop a better understanding of the effect of hydrogen on dimensional changes of channel materials, first-principles calculations combined with embedded-atom molecular dynamics simulations have been performed under EPRI's BWR channel distortion program. The simulations reveal that: (1) H dissolved in zirconium expands the lattice; (2) the volume effect of H in solution and as hydride is similar; (3) regions under tensile strain attract hydrogen; (4) near Ni atoms the binding of H is increased, and reduced near Sn and Nb; (5) 1 % Zr vacancies decrease the volume by 0.44 % and 1 % Zr self-interstitial atoms (SIAs) expand the volume by up to 1.2 %; (6) the bulk modulus of hydrides rises with increasing H concentration, the shear modulus of hydrides is similar to that of pure Zr, while Young's modulus decreases; (7) coalescence of isolated vacancies into dislocation loops releases up to 80kJ/mol; (8) vacancy dislocation loops larger than 10-15 angstrom in diameter tend to collapse thereby shrinking the lattice; (9) interstitial hydrogen is attracted to isolated vacancies and vacancy loops and can retard or prevent their collapse; (10) the diffusivity of interstitial Zr is higher than that of interstitial H atoms, diffusion of vacancies is slower; (11) substitutional Fe and Cr atoms spontaneously swap with interstitial Zr atoms and diffuse rapidly in the c direction; and (12) Nb impedes the diffusion of Zr self interstitials, thus reducing the buildup of a-loops. These simulations confirm some trends observed in material test reactors followed by advanced transmission electron microscopy (TEM). Simulations can be used to help optimize the materials properties for development of future channel alloys to minimize their in-service distortion up to very high fluence.

Highly-ordered and hexagonally close-packed nanoporous zirconium oxide layer is formed on the surface of zirconium alloy by anodization, and the anti-oxidation behavior of the zirconium alloy with the nanoporous oxide layer has been investigated. The oxidation experiments were carried out in both air and steam environments at 1000 °C. Interestingly, zirconium alloy with the nanoporous oxide layer exhibits dramatic improvement in the oxidation resistance compared to bare zirconium alloy without the nanoporous oxide layer. Analysis using several characterization tools reveals that large single-crystalline columnar zirconium oxide grains are formed beneath the nanoporous oxide layer and these grains prevent further oxidation.

The ability of a zirconium alloy to resist corrosion relies on a compromise between two opposing strategies. Minimizing the hydrogen pickup fraction (HPUF) by invoking metallic electron conduction in the barrier oxide results in rapid parabolic oxide growth. On the other hand, slow sub-parabolic barrier oxide growth, as reflected in rate limiting electron transport, may result in a high HPUF. The objective of the present study is to offer mechanistic insights as to how low concentrations of different alloying elements become decisive for the overall corrosion behavior. Combining atomistic microanalysis with first principles modeling by means of density functional theory, the speciation and redox properties of Fe and Ni towards hydrogen evolution are firstly explored. Complementary atom probe microanalysis at the metal–oxide interface provides evidence for Fe and Ni segregation to grain boundaries in Zircaloy-2 that propagates into the ZrO2 scale. Descriptors for how alloying elements in ZrO2 control electron transport as well as catalytic electron-proton recombination in grain boundaries to form H2 are determined by means of theory. The findings are generalized by further atomistic modeling, and are thus put in the context of early reports from autoclave experiments on HPUFs of zirconium with the alloying elements Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, and Nb. A shunting mechanism which combines inner and outer hydrogen evolution mechanisms is proposed. Properties of the transient zirconium sub-oxide are discussed. A plausible atomistic overall understanding emerges.

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.