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During the operation of a light water reactor, a fraction of the hydrogen produced by waterside corrosion is absorbed into the nuclear fuel cladding. When the hydrogen concentration reaches its solubility limit, a brittle zirconium hydride phase precipitates, leading to a loss of ductility of the cladding. To assess cladding integrity, an accurate simulation tool is needed to predict hydrogen distribution within the cladding and hydride precipitation. Recent studies have developed an improved understanding of the physical processes involved in hydrogen redistribution, and hydride precipitation and dissolution. This research led to the development of a new model, called Hydride Nucleation-Growth-Dissolution (HNGD). The present work describes the implementation of HNGD into the fuel performance code BISON, developed at Idaho National Laboratory. The main innovative feature of the HNGD model is that it accounts for hydride nucleation and growth as two distinct precipitation components, using the Johnson-Mehl-Avrami-Kolmogorov model to describe hydride growth kinetics. Each step of the model implementation into BISON was systematically verified, and simulations of experiments performed for validation, showing that the HNGD model provides improved predictions, and captures some experimentally observed physical phenomena related to hydride growth that the previous model could not.

Austenitic 21Cr32Ni model alloy thin foils, previously irradiated with 5 MeV Fe++ ions in bulk to create voids, were re-irradiated in-situ in the Intermediate Voltage Electron Microscope Facility (IVEM). The voids which had been formed under bulk-ion irradiation shrank and disappeared after in-situ Kr ion irradiation in the temperature range 50 K-713 K to an additional dose of 1 dpa. The voids were unaffected by eithersuccessive thermal annealing to 673 K and by prolonged exposure to the 200 keV electron beam at the irradiation temperature. The high void shrinkage rate observed did not change significantly for irradiation temperatures between 50 K and 713 K, suggesting that the void shrinkage process in thin foils during in-situ heavy-ion irradiation results from the interactions of displacement cascades with the voids. Possible void shrinkage mechanisms under thin foil irradiation are discussed in this study.

Introducing elastic energy in the phase field method has been shown to influence interfacial energy, depending on the elastic interpolation scheme. This study investigates the impact of the elastic energy when using a grand potential-based phase field method, comparing the result of Khachaturyan’s strain interpolation scheme (KHS) and Voight-Taylor’s elastic energy interpolation scheme (VTS). The KHS model leads to a decrease in the interfacial energy, while the VTS model leads to an increase. The change in interfacial energy is greater with the VTS model than the KHS model, which suggests that the KHS model is more appropriate to limit the artificial impact of the elastic energy on the interfacial energy. When the contribution at the interface is not negligible, it is shown that both the microstructure evolution kinetics and the equilibrium microstructure can be influenced by the choice of the elastic scheme being used. In addition, this paper shows that the grand potential model might not be appropriate when the system requires the introduction of a composition-dependent term in the elastic energy contribution. This limitation is due to the need for an explicit and invertible relation between the total potential and the composition.

An in situ ion-irradiation study, simultaneously examined using transmission electron microscopy, was performed to investigate irradiation-induced disordering and amorphization of Al3Ti-based intermetallic compounds. Thin foil samples of two crystalline structures: D022-structured Al3Ti and L12-structured (Al,Cr)3Ti were irradiated using 1.0 MeV Kr ions at a temperature range from 40 K to 573 K to doses up to 4.06 × 1015 ions/cm2. The results showed that both the compounds underwent an order-disorder transformation under irradiation, where both Al3Ti and (Al,Cr)3Ti ordered structures were fully transformed to the disordered face-centered cubic (FCC) structure except at the highest irradiation temperature of 573 K. A slightly higher irradiation dose was required for order-disorder transformation in case of Al3Ti as compared to (Al,Cr)3Ti at a given temperature. However, their amorphization resistances were different: while the disordered FCC (Al,Cr)3Ti amorphized at the irradiation dose of 6.25 × 1014 ions/cm2 (0.92 dpa) at 40 K and 100 K, the Al3Ti compound with the same disordered FCC structure maintained crystallinity up to 4.06 × 1015 ions/cm2 (5.62 dpa) at 40 K. The critical temperature for amorphization of (Al,Cr)3Ti under Kr ion irradiation is likely between 100 K and room temperature and the critical temperature for disordering between room temperature and 573 K.

In this comparative study, in situ investigations of the microstructure evolution in a Fee9Cr ferritic emartensitic steel, NF616, and a Fee9Cre0.1C-model alloy with a similar ferriticemartensitic microstructure have been performed. NF616 and Fee9Cre0.1C-model alloy were irradiated to high doses (up to ~10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fee9Cre0.1C-model alloy around ~1 dpa. At low temperatures (50e298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fee9Cre0.1C-model alloy around ~2e3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ~5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fee9Cre0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fee9Cre0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

Coating used for radioactive fuel or a structural component in radioactive fuel reactors, e.g., nuclear fuel cladding alloys, can include a ternary monolithic coating or multiple layers of one or more layers of TiAlN TiZrN, TiCrN, TiNbN and/or CrN, ZrN, NbN, TiN, TaN, HfN. TiHfN, TaHfN, TaNbN, or mixed combinations and/or CrN, ZrN, NbN, TiN, TaN, Si3N4, and/or HfN. In addition, one or more layers can be comprised of a nitride, oxide, or carbide or mixed combination (i.e., carbonides, oxynitrides, oxycarbides, etc.) from Ti, Al, Zr, Cr, Si, Nb, Hf, or mixed combination (i.e., TiAlC 1-x N x ). The multilayer coating can be doped with a dopant.

Zirconium carbide (ZrC) is a candidate material for use in one of the layers of TRISO coated fuel particles to be used in the Generation IV high-temperature, gas-cooled reactor, and thus it is necessary to study the effects of radiation damage on its structure. The microstructural evolution of ZrCx under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Samples of nominal stoichiometries ZrC0.8 and ZrC0.9 were irradiated in situ using 1 MeV Kr2+ ions at various irradiation temperatures (T = 20 K–1073 K). In situ experiments made it possible to continuously follow the evolution of the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were systematically recorded at selected dose points. After a threshold dose during irradiations conducted at room temperature and below, black-dot defects were observed which accumulated until saturation. Once created, the defect clusters did not move or get destroyed during irradiation so that at the final dose the low temperature microstructure consisted only of a saturation density of small defect clusters. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but some coarsening of the microstructure with the formation of dislocation loops was observed at higher temperatures. The irradiated microstructure was found to be only weakly dependent on the stoichiometry.

A first principles oxidation model has been developed which can rationalize the kinetics observed in various zirconium alloys using a unified theoretical approach, which predicts that space charges in the oxide have a major impact on oxidation kinetics. As a first development, the parabolic oxidation of Zr–0.4Nb alloy is fitted by the model assuming that space charges in the oxide are compensated by Nb ions. The model quantitatively reproduces the oxidation kinetics using physically significant parameters. XANES examinations show that there are enough aliovalent Nb ions in the oxide layer to verify oxide electroneutrality as predicted by the model.

A study of the oxide layers formed in 360 °C and 316 °C water on Zircaloy-4 samples has been performed in an attempt to help answer fundamental questions about oxide protectiveness, growth mechanisms, and the nature of oxide growth during autoclave corrosion. Two different oxide thicknesses – 12 and 39.5 μm – were investigated. Microbeam synchrotron radiation diffraction and fluorescence techniques with an X-ray beam size of 0.2 μm were used to characterize oxide in cross sections to determine the oxide phase content, grain size, texture, and orientation relationships as a function of through-thickness from the oxide–metal interface. The results confirm that the oxide is comprised primarily of monoclinic ZrO 2 , with tetragonal ZrO 2 present in small amounts. The observed diffraction peaks are consistent with monoclinic phases having a strong fiber texture with the 200 m plane aligned with the oxide–metal interface, and with the 011 m plane closely aligned with the transverse-normal ( T ) plane. The fraction of bulk tetragonal phase increased in the region located within one transition thickness near the oxide–metal interface. A strong periodicity was seen in oxide intensity from both the monoclinic and tetragonal phases corresponding to an oxide transition thickness of 1.8–1.9 μm. The grain size of the tetragonal phase was determined to be smaller than the monoclinic phase, and the grain size for the monoclinic phase decreased starting at a distance of approximately one transition layer from the oxide–metal interface. The relative amounts of monoclinic peak broadening due to strain and grain size were calculated, the former being approximately constant, while the latter decreased with increasing distance from the oxide–metal interface, corresponding to an increase in grain size. These findings are compared to previously performed microbeam diffraction experiments on Zircaloy-4 and other Zr-based alloys.

During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In zirconium alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup increases when electron transport decreases, requiring hydrogen ingress to close the reaction.

In an effort to understand the mechanisms resulting in the variations of corrosion rate observed throughout corrosion including at the first kinetic transition when the oxide temporarily loses its protective character, the oxide/metal interfaces of autoclave corroded Zircaloy-4 and ZIRLO™ before and after the transition are characterized using electron energy loss spectroscopy and atom probe tomography. The results reveal a complex structure of different phases at different stages of corrosion. The oxide/metal interface exhibits an intermediate layer, with an oxygen content between 45 and 55 O at.% and a suboxide layer corresponding to an oxygen-saturated solid solution in the metal matrix side (∼30 O at.%). Local variations are observed in the width of these characteristic structural features, especially near the transition. Good agreement on the layers present as well as their order, composition, and width was seen with the two techniques.

As a result of in-reactor corrosion during operation in nuclear reactors, hydrogen can enter the zirconium fuel cladding and precipitate as brittle hydride particles, which may reduce cladding ductility. Dissolved hydrogen responds to temperature gradients, resulting in transport and precipitation into cold spots so that the distribution of hydrides in the cladding is inhomogeneous. The hydrogen precipitation kinetics plays a strong role in the spatial distribution of the hydrides in the cladding. The precipitation rate is normally described as proportional to the supersaturation of hydrogen in solid solution. The proportionality constant, α2, for hydride precipitation in Zircaloy-4 is measured directly using in situ synchrotron X-Ray diffraction, at different temperatures and with three different initial hydrogen concentrations. The results validate the linear approximation of the phenomenological model and a near constant value of α2 = 4.5 × 10−4 s−1 was determined for the temperature range studied.

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.

Chloride induced stress corrosion cracking (SCC) of austenitic stainless steel is a potential issue in long term dry storage of spent nuclear fuel canisters. In order for SCC to occur there must be a corrosive environment, a susceptible material, and a driving force. Because it is likely that the material in the heat affected zone (HAZ) of welded stainless steel structures has been sensitized as a result of chromium depletion at the grain boundaries and a thermal residual stress driving force is likely present if solution annealing is not performed, two issues are critical. Is the environment corrosive, i.e., are chlorides present in solution on the surface? And then, are there cracks that could propagate? Remote detection of chlorides on the surface can be accomplished by laser induced breakdown spectroscopy (LIBS), while cracks can be detected by shear horizontal guided waves generated by electromagnetic acoustic transducers (EMATs). Both are noncontact methods that are amenable to robotic delivery systems and harsh environments. The sensitivity to chlorine on stainless steel of a LIBS system that employs optical fiber for pulse delivery is demonstrated. Likewise, the ability of the EMAT system to detect cracks of a prescribed size and orientation is shown. These results show the potential for remote detection of Cl and cracks in dry storage spent fuel canisters.Chloride induced stress corrosion cracking (SCC) of austenitic stainless steel is a potential issue in long term dry storage of spent nuclear fuel canisters. In order for SCC to occur there must be a corrosive environment, a susceptible material, and a driving force. Because it is likely that the material in the heat affected zone (HAZ) of welded stainless steel structures has been sensitized as a result of chromium depletion at the grain boundaries and a thermal residual stress driving force is likely present if solution annealing is not performed, two issues are critical. Is the environment corrosive, i.e., are chlorides present in solution on the surface? And then, are there cracks that could propagate? Remote detection of chlorides on the surface can be accomplished by laser induced breakdown spectroscopy (LIBS), while cracks can be detected by shear horizontal guided waves generated by electromagnetic acoustic transducers (EMATs). Both are noncontact methods that are amenable to robotic delivery syst...

Dysfunctional Uterine Bleeding (DUB) represents a particular type of Abnormal Uterine Bleeding (AUB), and is defined in the USA as an excessive, prolonged and irregular bleeding of the endometrium (frequency 7 days; daily use of sanitary towels/tampons >1/1–2 h), that does not cause pain and does not have any organic cause, so much so that it is frequently considered to be a symptom of anovulatory bleeding. The initial assessment of an adolescent with DUB requires an accurate summary of the patient’s medical history, both familiar (in order to exclude haemorrhagic diatheses) and personal (previous surgery, trauma with ecchymosis, epistaxis or gingival bleeding, previous or current pathologies, use of drugs, accurate menstrual and sexual history), in addition to an objective medical examination (body mass index, stage of puberty, inspection of external genitalia with an eventual gynaecological examination for sexually active patients, abdominal examination). It can sometimes be useful, although not strictly necessary, to perform a pelvic ultrasound (abdominal/transrectal or transvaginal in a sexually active patient). Among the possible causes of DUB, there are blood clotting disorders, renal and hepatic impairment, diabetes, enteric, rheumatic, cardiac and neurological diseases, as well as thyroid disorders and hyperandrogenic alterations. In this chapter, we will discuss how to manage all the possible causes of DUB and, finally, summarise a complete overview of the possible treatments.

Waterside corrosion of zirconium alloy nuclear fuel cladding varies markedly from one alloy to another. In addition, for a given alloy, the corrosion rate evolves during the corrosion process, most notably when the oxide loses its stability at the oxide transition. In an effort to understand the mechanism resulting in the variations of corrosion rate observed at the oxide transition, oxide layers formed on Zircaloy-4 and ZIRLO™ in high temperature water autoclave environments, and archived before and after the transition, are characterized using transmission electron microscopy. The study characterizes and compares the oxide morphology in both alloys at different times during the corrosion process, in an effort to understand the oxide growth mechanism for these alloys. Results show that the oxide is mainly composed of monoclinic ZrO 2 1 , with a preponderance of columnar oxide grains which extend to the oxide/metal interface. The oxide formed right after the transition has occurred, exhibits a 150 nm-wide layer of small equiaxed grains with high tetragonal oxide fraction. This layer has a similar morphology and structure as the first oxide layer formed (observed near the oxide/water interface). A study of the oxygen-rich region near the oxide/metal interface reveals a complex structure of different phases at different stages of corrosion. The interface exhibits an intermediate layer, identified as ZrO, a discontinuous layer of “blocky” Zr 3 O grains embedded in the ZrO layer, and a suboxide layer corresponding to an oxygen saturated solid solution in the metal matrix side. The thickness of this interfacial layer decreased markedly at the transition. Hydrides are also observed in that region, with a definite orientation relationship with the matrix. The observations of the oxide/metal interface are qualitatively similar for the two alloys but quantitatively different. The incorporation of intermetallic precipitates into the oxide layer is also studied, and compared between the two alloys. These results are discussed in terms of previous observations and of current models of oxide growth.

Abnormal uterine bleeding (AUB) is defined as any atypical genital bleeding originating from the uterine cavity, without the characteristics of normal menstrual period. AUB is an important symptom both for adolescents and their parents, and it usually leads to a state of anxiety. Although about 95% of AUB could be considered as a dysfunctional disorder, AUB requires well-defined diagnostic procedures in order to detect a physical cause, ruling out complex or systemic diseases, including oncological ones. Diagnostic procedures require the acquisition of a full and detailed history, and it is also crucial to obtain as much compliance from the patient as possible. A complete gynecological evaluation (whenever possible) and a full physical examination are useful to detect any kind of general disease which can compromise the hormonal reproductive system. Auxiliary tools such as gynecological ultrasonography for pelvic examination are allowed in sexually-active women, otherwise transrectal ultrasonography could be considered, if needed. Hematic β-hCG must be dosed in every fertile woman with AUB and laboratory tests must be tailored on each patient. The first-line treatment consists of combined oral contraceptives and, when they are contraindicated, progesterone alone, medicated intrauterine devices, GnRH-analogues, or desmopressin are the most common second-line treatments.

Genital malformations range from congenital absence of the vagina and uterus, to defects in lateral or vertical fusion of the Mullerian ducts, and to the formation of external genitalia that are ambiguous in sexual differentiation.

The optimization of zirconium-based alloys used for nuclear fuel cladding aims to reduce hydrogen pickup during operation, and the associated cladding degradation. The present study focuses on precisely and accurately measuring hydrogen pickup fraction for a set of alloys to specifically investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. To measure hydrogen concentrations in zirconium alloys two techniques have been used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results of both techniques show that hydrogen pickup fraction varies significantly with exposure time and between alloys. A possible interpretation of the results is that hydrogen pickup results from the need to balance charge. That is, the pickup of hydrogen shows an inverse relationship to oxidation kinetics, indicating that, if transport of charged species is rate limiting, oxide transport properties such as oxide electronic conductivity play a key role in the hydrogen pickup mechanism. Alloying elements (either in solid solution or in precipitates) would therefore impact the hydrogen pickup fraction by affecting charge transport.

Autoclave corrosion tests of Zircaloy-4 were performed at temperatures ranging from 274°C to 427 °C to determine how the characteristics of the oxide layer changed with corrosion temperature. Corroded samples were prepared in cross-section and examined at the Advanced Photon Source using x-ray diffraction and fluorescence with sub-micron spatial resolution. These measurements provide high fidelity data for characterization of the oxide layer grown on zirconium alloys over a range of temperatures to better understand the fundamental mechanisms of corrosion. It was found that the size of the monoclinic oxide grains increases with distance from the metal-oxide interface, while the average grain size of the tetragonal oxide phase decreases. This is consistent with the continuous nucleation of a mixture of equiaxed tetragonal and monoclinic grains at the metal-oxide interface, of which only the monoclinic oxide grains that are properly aligned to minimize stresses grow into long columnar oxide grains. The tetragonal phase oxide grains transform to monoclinic oxide once a critical size is reached. While the monoclinic oxide grain size increases with corrosion temperature, the tetragonal phase grain size remains below a maximum value. Evolution of the tetragonal fraction, calculated with the Garvie-Nicholson formula, was observed, with local maxima seen just before transition and local minima seen at oxide locations corresponding to the oxide kinetic transition. No difference was observed in the tetragonal phase fraction in the oxide layers formed at different corrosion temperatures. Detailed fluorescence and diffraction mapping of one sample showed the evolution of the diffraction intensity over a short range in the oxide. This work provides a systematic study revealing fundamental characteristics of the microstructure of Zircaloy-4 oxide layers to further understanding of corrosion resistance of these materials.

A detailed study of alloying element distributions in the metal and oxygen rich regions of corroded Zr alloys and of the phases formed ahead of the oxide front was conducted using atom probe tomography (APT). A consistent sequence of sub-oxide phases is observed ahead of the ZrO 2 oxide front, consisting of (i) a thin layer of equiatomic ZrO (occasionally slightly over and under stoichiometric) (ii) saturated solid solution Zr(O) sat , and (iii) a slowly decreasing oxygen profile into the metal. The results also show that the distribution of the alloying elements in the metal is more inhomogeneous than previously thought and that in the oxygen-rich phases enhanced segregation is observed, compared to the metal.

Corrosion tests were performed in steam and supercritical water at 500 °C for two ferritic–martensitic alloys: HCM12A and NF616. The corrosion kinetics for the two alloys are similar in both environments showing near cubic behavior, but the corrosion rate was significantly higher in supercritical water than in steam. Examinations of the oxide layers using scanning electron microscopy and microbeam synchrotron diffraction and fluorescence show that both alloys form two-layer oxide structures in either environment. The outer layer contains only Fe 3 O 4 , while the inner layer contains a mixture of Fe 3 O 4 and FeCr 2 O 4 . Additionally, marker experiments using a novel photolithographic deposition process show that the original water–metal interface corresponds with the outer–inner layer interface as expected. The results are discussed in light of known corrosion mechanisms.

The circumferential hydrides normally present in nuclear reactor fuel cladding after reactor exposure may dissolve during drying for dry storage and re-precipitate when cooled under load into a more radial orientation, which could embrittle the fuel cladding. It is necessary to study the rates and conditions under which hydride reorientation may happen in order to assess fuel integrity in dry storage. The objective of this work is to study the effect of applied stress and thermal cycling on the hydride morphology in cold-worked stress-relieved Zircaloy-4 by combining conventional metallography and in situ X-ray diffraction techniques. Metallography is used to study the evolution of hydride morphology after several thermo-mechanical cycles. In situ X-ray diffraction performed at the Advanced Photon Source synchrotron provides real-time information on the process of hydride dissolution and precipitation under stress during several thermal cycles. The detailed study of diffracted intensity, peak position and full-width at half-maximum provides information on precipitation kinetics, elastic strains and other characteristics of the hydride precipitation process. The results show that thermo-mechanical cycling significantly increases the radial hydride fraction as well as the hydride length and connectivity. The radial hydrides are observed to precipitate at a lower temperature than circumferential hydrides. Variations in the magnitude and range of hydride strains due to reorientation and cycling have also been observed. These results are discussed in light of existing models and experiments on hydride reorientation. The study of hydride elastic strains during precipitation shows marked differences between circumferential and radial hydrides, which can be used to investigate the reorientation process.

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser-induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. In addition, ultrasonic-guided wave detection of crack-like notches oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurements is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask's overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Background: The Aptima HCV Quant Dx assay (Aptima assay) is a fully automated quantitative assay on the Panther® system. This assay is intended for confirmation of diagnosis and monitoring of HCV RNA in plasma and serum specimens. The purpose of the testing described in this paper was to evaluate the performance of the Aptima assay. Methods: The analytical sensitivity, analytical specificity, precision, and linearity of the Aptima assay were assessed. The performance of the Aptima assay was compared to two commercially available HCV assays; the Abbott RealTime HCV assay (Abbott assay, Abbott Labs Illinois, USA) and the Roche COBAS Ampliprep/COBAS Taqman HCV Quantitative Test v2.0 (Roche Assay, Roche Molecular Systems, Pleasanton CA, USA). The 95% Lower Limit of Detection (LoD) of the assay was determined from dilutions of the 2nd HCV WHO International Standard (NIBSC 96/798 genotype 1) and HCV positive clinical specimens in HCV negative human plasma and serum. Probit analysis was performed to generate the 95% predicted detection limits. The Lower Limit of Quantitation (LLoQ) was established for each genotype by diluting clinical specimens and the 2nd HCV WHO International Standard (NIBSC 96/798 genotype 1) in HCV negative human plasma and serum. Specificity was determined using 200 fresh and 536 frozen HCV RNA negative clinical specimens including 370 plasma specimens and 366 serum specimens. Linearity for genotypes 1 to 6 was established by diluting armored RNA or HCV positive clinical specimens in HCV negative serum or plasma from 8.08 log IU/mL to below 1 log IU/mL. Precision was tested using a 10 member panel made by diluting HCV positive clinical specimens or spiking armored RNA into HCV negative plasma and serum. A method comparison was conducted against the Abbott assay using 1058 clinical specimens and against the Roche assay using 608 clinical specimens from HCV infected patients. In addition, agreement between the Roche assay and the Aptima assay using specimens with low HCV concentrations (

A new in situ sample environment has been designed and developed to study the interfacial interactions of nuclear cladding alloys with high temperature steam. The sample environment is particularly optimized for synchrotron X-ray diffraction studies for in situ structural analysis. The sample environment is highly corrosion resistant and can be readily adapted for steam environments. The in situ sample environment design complies with G2 ASTM standards for studying corrosion in zirconium and its alloys and offers remote temperature and pressure monitoring during the in situ data collection. The use of the in situ sample environment is exemplified by monitoring the oxidation of metallic zirconium during exposure to steam at 350 °C. The in situ sample environment provides a powerful tool for fundamental understanding of corrosion mechanisms by elucidating the substoichiometric oxide phases formed during the early stages of corrosion, which can provide a better understanding of the oxidation process.

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. Detection of a notch oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurement is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask’s overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are: to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Associated to the weld quality, the weld bead geometry is one of the most important parameters in welding processes. It is a significant requirement in a welding project, especially in automatic welding systems where a specific width, height, or penetration of weld bead is needed. This paper presents a novel technique for real-time measuring of the width and height of weld beads in gas metal arc welding (GMAW) using a single high-speed camera and a long-pass optical filter in a passive vision system. The measuring method is based on digital image processing techniques and the image calibration process is based on projective transformations. The measurement process takes less than 3 milliseconds per image, which allows a transfer rate of more than 300 frames per second. The proposed methodology can be used in any metal transfer mode of a gas metal arc welding process and does not have occlusion problems. The responses of the measurement system, presented here, are in a good agreement with off-line data collected by a common laser-based 3D scanner. Each measurement is compare using a statistical Welch’s t-test of the null hypothesis, which, in any case, does not exceed the threshold of significance level α = 0.01, validating the results and the performance of the proposed vision system.

This paper presents and discusses the results of an ongoing R&D project aiming to design and build a fully automated prototype of a specialized spherical robotic welding system for repairing hydraulic turbine surfaces eroded by cavitation pitting and/or cracks produced by cyclic loading. The system has an embedded vision sensor built to acquire range images by laser scanning over the blade's surface and produce 3D models to locate the damaged spots to be registered in a 3D coordinate system into the robot controller, enabling the robot to repair the flaws automatically by welding in layers. The paper is focused on the robot kinematic model and describes an iterative algorithm to process the inverse kinematics with only five degrees-of-freedom. The algorithm makes use of data collected from a vision sensor to ensure that the welding gun axis is perpendicular to the blade's surface. Besides this, it proposes a modelling and optimization mathematical routine for more efficient robot calibration, which can be used with any type of robot. This robot calibration optimization scheme finds the optimal error parameter vector based on the condition number of the manipulator transformation composed from the partial derivatives of the error parameters. Experimental results proved both the iterative algorithm to perform the inverse kinematics and the technique to optimize robot calibration to be very efficient.

Este trabalho apresenta uma metodologia para modelar empiricamente o modo de transferência goticular projetada em processos de soldagem GMAW utilizando algoritmos de processamento de imagens digitais e a metodologia de superfície de resposta. Modelos correspondentes ao tamanho e à frequência de destacamento das gotas são obtidos em relação às entradas de tensão e velocidade de alimentação do arame. Os dados experimentais de tamanho e frequência de destacamento das gotas, com os quais foram desenvolvidos os modelos, foram calculados a partir de imagens de transferência metálica adquiridas com uma câmera de alta velocidade e a técnica de iluminação conhecida como perfilografia. As técnicas utilizadas na coleta dos dados e os modelos encontrados validaram a metodologia proposta por apresentar uma correta representação do modo de transferência. O empescoçamento produzido pelo efeito pinch se apresentou como um parâmetro adequado para rastrear a formação das gotas e calcular a sua frequência de destacamento.This paper presents a methodology to get empirical models of projected spray transfer mode in GMAW processes. Digital image processing algorithms were used together with the surface response methodology. Droplet size and transfer rate models were obtained in relation to voltage and wire feed rate. To develop the models, experimental data collection of droplet size and transfer rate were calculated from metal transfer images acquired with a high-speed camera and backlighting illumination. The techniques for data collecting and the models constructed validated the proposed methodology presenting a correct transfer mode representation. The "neck-shrink" produced by the pinch effect proved to be an appropriate parameter to track and for calculating the droplets transfer rate.

The inner-diffusion layer interface of an HCM12A sample oxidized in 600 °C supercritical water (SCW) was analyzed using EFTEM and EELS. The EFTEM analysis showed the presence of chromium-rich zones linked with the porosity within the inner layer, as well as a nanometric iron–chromium separation, which may be linked with the presence of both Fe3O4 and FeCr2O4 in this layer. The diffusion layer was characterized by large chromium-rich oxides located at the tempered martensite lath boundaries, which suggested the preferential grain boundary diffusion of oxygen and the preferential oxidation of the chromium carbides present at these boundaries. The metal grains of the diffusion layer contained nanometric chromium-rich spinel oxides. The presence of large chromium-rich oxide precipitates in the diffusion layer appears to help improve the corrosion resistance of these alloys.

The ingress of hydrogen during corrosion in service can degrade the mechanical properties of zirconium alloy nuclear fuel cladding because of the formation of brittle hydrides. The formation of these hydrides is reviewed in light of recent synchrotron radiation experimental results and phase-field modeling computational results that provide new insight on the process.

We propose a novel use of a non-destructive technique to quantitatively assess hydrogen concentration in zirconium alloys. The technique, called Cold Neutron Prompt Gamma Activation Analysis (CNPGAA), is based on measuring prompt gamma rays following the absorption of cold neutrons, and comparing the rate of detection of characteristic hydrogen gamma rays to that of gamma rays from matrix atoms. Because the emission is prompt, this method has to be performed in close proximity to a neutron source such as the one at the National Institute of Technology (NIST) Center for Neutron Research. Determination shown here to be simple and accurate, matching the results given by usual destructive techniques such as Vacuum Hot Extraction (VHE), with a precision of ±2 mg kg−1 (or wt ppm). Very low levels of hydrogen (as low as 5 mg kg−1 (wt ppm)) can be detected. Also, it is demonstrated that CNPGAA can be applied sequentially on an individual corrosion coupon during autoclave testing, to measure a gradually increasing hydrogen concentration. Thus, this technique can replace destructive techniques performed on “sister” samples thereby reducing experimental uncertainties.

NF616 is a nominal 9Cr ferritic–martensitic steel that is amongst the primary candidates for cladding and duct applications in the Sodium-Cooled Fast Reactor, one of the Generation IV nuclear energy systems. In this study, an in situ investigation of the microstructure evolution in NF616 under heavy ion irradiation has been conducted. NF616 was irradiated to 8.4 dpa at 50 K and to 7.6 dpa at 473 K with 1 MeV Kr ions. Nano-sized defects first appeared as white dots in dark-field TEM images and their areal density increased until saturation (∼6 dpa). Dynamic observations at 50 K and 473 K showed appearance and disappearance of TEM-visible defect clusters under irradiation that continued above saturation dose. Quantitative analysis showed no significant change in the average size (∼3–4 nm) and distribution of defect clusters with increasing dose at 50 K and 473 K. These results indicate a cascade-driven process of microstructure evolution under irradiation in these alloys that involves both the formation of TEM-visible defect clusters by various degrees of cascade overlap and cascade induced defect cluster elimination. According to this mechanism, saturation of defect cluster density is reached when the rate of defect cluster formation by overlap is equal to the rate of cluster elimination during irradiation.

Ferritic–martensitic and ODS alloys are primary candidates for application as cladding and structural materials in the Generation IV Supercritical Water Reactor. One of the main in-service degradation mechanisms for these alloys is uniform corrosion. This article analyzes the oxide microstructure formed on these alloys to better understand their oxidation behavior. Corrosion tests were performed in both steam and supercritical water (SCW) at 500 and 600 °C. The oxide microstructure was analyzed using microbeam synchrotron radiation diffraction and fluorescence associated with electron microscopy. The oxide forms a three-layer structure with an outer layer containing only Fe 3 O 4 , an inner layer containing a non-uniform (Fe,Cr) 3 O 4 spinel structure, and a diffusion layer containing a mixture of metal grains and chromium-rich precipitates. A marker experiment located the original water–metal interface as the outer–inner layer interface implying a mechanism where iron migrates outwards to form the outer layer and oxygen diffuses inwards to form the inner layer.

In a reactivity-initiated accident, cladding failure may occur by crack initiation within a defect such as a hydride rim or blister and subsequent crack propagation through the thickness of the thin-wall cladding. In such a circumstance, determining the cladding resistance to crack propagation in the through-thickness direction is crucial to predicting cladding failure. To address this issue, through-thickness crack propagation in hydrided Zircaloy-4 sheet was analyzed at 25 °C, 300 °C, and 375 °C. At 25 °C, the fracture toughness decreased with increasing hydrogen content and with an increasing fraction of radial hydrides. Hydride particles fractured ahead of the crack tip, creating a path for crack growth. At both 300 °C and 375 °C, the resistance to crack-growth initiation was sufficiently high that crack extension was often caused by crack-tip blunting. There was no evidence of hydride particles fracturing near the crack tip, and no significant effect of hydrogen content on fracture toughness was observed at these elevated temperatures.

In Italy the use of oral contraceptives (OCs) is low in comparison with other countries,1 possibly because bimanual pelvic examination and laboratory tests are required before they can be prescribed. Such requirements may stop OCs being used, particularly by teenagers and young women, and so should be considered only in selected cases.2–4 We investigated the recall of these requirements by teenagers and young women in an online anonymous questionnaire on Facebook. Duplicate responses were ignored. A total of 842 women (mean age 21.5 (SD 1.7) years) completed the questionnaire; 236 smoked cigarettes, 177 had graduated from university, and 13 …

The influence of the alloy microstructure and microchemistry on uniform waterside corrosion of zirconium alloys is reviewed, with special attention to the various stages of corrosion, such as pre-transition, transition, and breakaway.

One of the greatest challenges in today's industrial robotics is the development of off-line programming systems that allow drastic reduction in robots' reprogramming time, improving productivity. The article purpose is to pave the way to the construction of generic calibration systems easily adapted to any type of robot, regardless their application, such as modular robots and robot controllers specifically designed for non-standard applications. A computer system was built for developing and implementing a calibration system that involves the joint work of computer and measurement systems. Each step of this system's development is presented together with its theoretical basis. With the development of a remote maneuvering system based on ABB S3 controller experimental tests have been carried out using an IRB2000 robot and a measurement arm (ITG ROMER) with 0.087 mm of position measurement accuracy. The robot model used by its controller was identified and the robot was calibrated and evaluated in different workspaces resulting in an average accuracy improvement from 1.5 mm to 0.3 mm.

The orientation and distribution of hydrides formed in zirconium alloy nuclear fuel cladding can strongly influence material behavior and in particular resistance to crack growth. The hydride microstructure and hydride platelet orientation (whether in-plane or radial relative to the cladding tubes) are crucial to determining cladding failure limits during mechanical testing. Hydride formation is normally studied by post-facto metallography, performed at room temperature and in the absence of applied stress. This study uses synchrotron radiation to observe in situ the kinetics of hydride dissolution and precipitation in previously hydrided Zircaloy samples. The experiments allow the direct observation of hydride dissolution, re-precipitation, and re-orientation, during heating and cooling under load. The solubility limits and the hydride-matrix orientation relationship determined from in situ experiments were in good agreement with previous post-facto examinations of bulk materials. The present measurements performed under stress and at temperature showed a characteristic diffraction signature of reoriented hydrides. The results suggest a threshold stress for hydride re-orientation between 75 and 80 MPa for the microstructure/texture studied. These results are discussed in light of existing knowledge.

Zirconium hydrides precipitate in fuel cladding alloys as a result of hydrogen uptake from the high-temperature corrosion environment of light water reactors. Synchrotron X-ray diffraction was performed at room temperature on stress-relieved Zircaloy-4 cladding with two distributions of hydrides – (1) uniformly distributed hydrides across the entire cladding wall and (2) hydride rim next to the outer surface. The d-hydride phase was found to be the predominant hydride phase to precipitate for hydrogen contents up to 1250 weight parts per million (wt ppm). At a higher content, about 3000 wt ppm, although dhydride is still the majority phase, a significant amount of c-hydride is also observed. At even higher hydrogen contents, in excess of approximately 6000 wt ppm, such as can occur in a highly dense hydride layer, peaks associated with the e-hydride phase are also observed in the diffraction pattern. The volume fraction of hydrides was estimated as a function of hydrogen content using the integrated intensities of select diffraction peaks corresponding to the a-Zr matrix and the hydride phases. These estimated values agree well with calculated values from the independently measured concentrations. The results of this study indicate that hydride precipitation in Zircaloy-4 is a complex process of evolving hydride phases with increasing local hydrogen content.