PDF Origine du nom de famille LE BOULCH (Oeuvres courtes) (French Edition)

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The samples of Zircaloy -2 and -4 tubes, used as fuel cladding in Indian boiling water reactors BWR and pressurized heavy water reactors PHWR , respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. Concentrations were also determined in a smaller size Zircaloy -4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory.

Values of sensitivities and detection limits have been evaluated for the elements analyzed. A Zr-Ti-Cu-Fe quaternary eutectic alloy was employed as a new Be-free brazing filler metal for Zircaloy -4 to supersede physically vapor-deposited Be coatings used conventionally with several disadvantages.

The quaternary eutectic composition of Zr58Ti16Cu10Fe16 at. The influence of strain rate and hydrogen on the plane-strain ductility of Zircaloy cladding. The authors studied the ductility of unirradiated Zircaloy -4 cladding under loading conditions prototypical of those found in reactivity-initiated accidents RIA , i. To conduct these studies, they developed a specimen configuration in which near plane-strain deformation is achieved in the gage section, and a testing methodology that allows one to determine both the limit strain at the onset of localized necking and the fracture strain.

Preliminary experiments on cladding containing ppm hydrogen show only a small loss of fracture strain but no clear effect on limit strain. The experiments also indicate that there is a significant loss of Zircaloy ductility when surface flaws are present in the form of thickness imperfections. Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI. Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures.

The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion SCC and not delayed hydride cracking DHC. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy.

This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures. Implications of Zircaloy creep and growth to light water reactor performance. Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors LWRs most of the core structural materials have been fabricated from either Zircaloy -2 or Zircaloy Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished.

Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design.

Although irradiation creep and growth have resulted in only one significant performance problem creep collapse of fuel cladding, which has been eliminated , deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure.

The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements.

Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic. Cyclic softening in annealed Zircaloy Role of edge dislocation dipoles and vacancies. The mechanism of cyclic softening in annealed Zircaloy -2 at low strain amplitudes under strain controlled fatigue at room temperature is rationalized. The unusual softening due to continuous decrease in the phenomenological friction stress is found to be associated with decrease in the resistance against movement of dislocations because of the formation and easy glide of pure edge dislocation dipoles and consequent decrease in friction stress from reduction in the shear modulus.

Positron annihilation spectroscopy data strongly support the increase in edge dislocation density containing jogs, from increased positron trapping and increase in annihilation lifetime. In-reactor oxidation of zircaloy -4 under low water vapor pressures. Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions.

Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation. In situ high temperature oxidation analysis of Zircaloy -4 using acoustic emission coupled with thermogravimetry. Two different atmospheres were used to study the oxidation of Zircaloy a helium and pure oxygen, b helium and oxygen combined with slight addition of air. Under a mixture of oxygen and air in helium, an acceleration of the corrosion was observed due to the detrimental effect of nitrogen.

The kinetic rate increased significantly after a kinetic transition breakaway. This acceleration was accompanied by an acoustic emission activity. Most of the acoustic emission bursts were recorded after the kinetic transition post-transition or during the cooling of the sample.


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Acoustic events were recorded during the isothermal dwell time at high temperature under air. They were associated with large cracks in the zirconia porous layer. Acoustic events were also recorded during cooling after oxidation tests both under air or oxygen. For the latter, cracks were observed in the oxygen enriched zirconium metal phase and not in the dense zirconia layer after 5 h of oxidation. This report presents an accelerated kinetic Monte Carlo KMC method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling.

The acceleration is achieved by replacing regular KMC jumps in trapping energy basins formed by neighboring tetrahedral interstitial sites, with analytical solutions for basin exiting time and probability. The diffusivity along is found to be slightly higher than that along , with the anisotropy saturated at about 1. PubMed Central. Room temperature mechanical properties of electron beam welded zircaloy -4 sheet. Room temperature mechanical properties of electron beam welded and plain Zircaloy -4 sheet 1.

Various welding parameters were utilized to join sheet material. Electron beam welded specimens and as-received sheet specimens show comparable mechanical properties. Zr-4 sheet displays anisotropy; tensile properties measured for transverse display higher elastic modulus, yield strength, reduction of area and slightly lower ductility than for the longitudinal rolling direction.

Hardness at heat-affected-zone is slightly higher. Electron microscopic examination shows distinct microstructure morphology and grain size at the weld zone, HAZ and parent metal. A correlation between welding parameters, mechanical properties and microstructural features was established for electron beam welded Zircaloy -4 sheet material.

Here, this report presents an accelerated kinetic Monte Carlo KMC method to compute the diffusivity of hydrogen in hcp metals and alloys, considering both thermally activated hopping and quantum tunneling. The properties studied were electrical resistivity and X-ray line sharpening. Recrystallization kinetics were described with sigmoidal curves derived from X-ray intensity and microhardness data. Light, replica, and transmission electron microscopy and selected-area electron diffraction were used to postulate recovery and recrystallization mechanisms.

Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes.

Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations Reference a. Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction Reference c. Once above the solubility limit in Zircaloy , the hydride precipitates as zirconium hydride References c through j.


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  6. Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal Reference k. Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface References l through n , in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer.

    References o through q investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference o , reported that the applied tensile stress increased corrosion rates above a critical stress level in C and C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed.

    A novel procedure employing laser ultrasound technique and simplex algorism for the characterization of mechanical and geometrical properties in Zircaloy tubes with different levels of hydrogen charging. In this research, a procedure employing a laser ultrasound technique LUT and an inversion algorism is reported for nondestructive characterization of mechanical and geometrical properties in Zircaloy tubes with different levels of hydrogen charging.

    With the LUT, guided acoustic waves are generated to propagate in the Zircaloy tubes and are detected remotely by optical means. By measuring the dispersive wavespeeds followed by the inversion algorism, mechanical properties such as elastic moduli and geometrical property such as wall-thickness of Zircaloy tubes are characterized for different levels of hydrogen charging. The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory INL was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design unfueled in a simulated PWR water environment under irradiation conditions.

    The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability LWRS program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. Intergrannular strain evolution in a zircaloy -4 alloy with Widmanstatten microstructure.

    A Zircaloy -4 alloy with Widmanstatten-Basketweave microstructure and random texture has been used to study the deformation systems responsible for the polycrystalline plasticity at the grain level. The evolution of internal strain and bulk texture is investigated using neutron diffraction and an elasto-plastic self-consistent EPSC modeling scheme. The macroscopic stress-strain behavior and intergranular hkil-specific strain development, parallel and perpendicular to the loading direction, were measured in-situ during uniaxial tensile loading.

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    Then, the EPSC model was employed to simulate the experimental results. The agreement between the experiment and the model will be discussed as well as the critical resolved shear stresses CRSS and the hardening coefficients obtained from the model. Transition joints between Zircaloy -2 and stainless steel by diffusion bonding. Bonding was carried out in a vacuum hot press, under compressive loading. Electron probe microanalysis and metallographic analysis showed a good metallurgical compatibility and also indicated the absence of discontunities, micropores and intermetallic compounds at various interfaces.

    The dimple structure on the fractured surface is indicative of the ductile mode of failure of the bonded assembly. Effect of He implantation on the microstructure of zircaloy -4 studied using in situ TEM. Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation.

    Characterization of zircaloy -4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2. Ordered arrays of He bubbles were observed at and K at a fluence of 1. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated. In situ synchrotron X-ray diffraction study of hydrides in Zircaloy -4 during thermomechanical cycling. The d-spacing evolution of both in-plane and out-of-plane hydrides has been studied using in situ synchrotron radiation X-ray diffraction during thermo-mechanical cycling of cold-worked stress-relieved Zircaloy In contrast, the d-spacing of the plane aligned with the hydride plate face initially contracts upon heating.

    On-site, dry cask storage was originally by the intended to be a short-term solution for holding spent nuclear fuel. Due to the lack of a permanent storage facility, the nuclear power industry seeks to assess the effective lifetime of the casks. One issue which could compromise cask integrity is Hydrogen embrittlement.

    This phenomenon occurs in the Zircaloy -4 fuel-rod cladding and is caused by the formation of Zirconium hydrides. Over time, thermal stresses caused by the heat from reactions of the stored nuclear fuel could result in significant breaches of the cladding. We will irradiate samples of the alloy with protons of energies up to keV using a Van de Graaff particle accelerator.

    Once irradiated, their properties will be characterized using scanning electron microscopy and Vickers hardness tests. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. For Zircaloy -4, the ductile to brittle transition occurs at an equivalent cladding reacted ECR of No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours. Post-quench ductility evaluation of Zircaloy -4 and select iron alloys under design basis and extended LOCA conditions.

    The purpose was to determine the oxidation behavior and post-quench ductility under postulated and extended LOCA conditions. No ductility decrease was observed for the post-quenched APMT samples oxidized up to 4 h. Texture and hydride orientation relationship of Zircaloy -4 fuel clad tube during its fabrication for pressurized heavy water reactors. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties.

    Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy -4 picks-up hydrogen. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation F n with the texture in the tube during its fabrication has been developed using a second order polynomial.

    The present work is aimed at quantification and correlation of texture evolved in Zircaloy -4 cladding tube using Kearn's f-parameter during its fabrication process. Des ballons pour demain. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys.

    The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the s and s involving the U.

    Federal government and U. National Laboratory employees involving the use of Be. Materion formerly, Brush Wellman also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and. Effects of anisotropy and irradiation on the deformation behavior of Zircaloy 2. Final report. An experimental program investigated the effects of texture anisotropy and irradiation on the mechanical behavior of Zircaloy Short time and time dependent mechanical behavior were considered. Irradiation effects were simulated through the use of 4.

    Both cold worked-stress relieved and annealed material were used in this experimental program. Monotonic flow loci were constructed for each texture. The tensile creep strength is proportional to the resolved fraction of basal poles in the test direction. In variable stress and temperature tests, the time-hardening rule was found to be inapplicable. In the irradiation creep tests irradiation hardening and enhanced irradiation creep were observed. Radiation hardening effects were significant in annealed material but were attenuated in cold worked-stress relieved material.

    Litigation involving DES. Focus is on the diethylstilbestrol DES litigation which has resulted from the discovery that this synthetic estrogen can cause cancer in the daughters of women who used the drug during pregnancy in an effort to prevent threatened abortion.

    Possibly suits are pending at this time in which DES daughters claim injuries. In most of these vaginal or cervical cancer has appeared -- with or without a hysterectomy having been performed. Several women died from cancer. The fact that the use of DES occurred many years ago is the legal hurdle most troublesome to lawyers.

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    The average women coming to a lawyer's office today has a mother who used some form of DES , perhaps in Few drugstores have records today of the prescriptions which they filled 20 years ago. It has been estimated that over the period more than different companies manufactured or "tabletized" under their own name DES plus a variety of similar synthetic estrogens promoted for the prevention of threatened abortion.


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    A further hurdle caused by the passage of time is that even the records of the physicians are frequently lost. A final problem created by the age of the cases is statute of limitations. If the actual manufacturer of the DES cannot be identified, this is generally the end of the lawyer's interest in the case. The chance of the plaintiff winning may be increased if the action against all the manufacturers is a class action. Most of the pending DES suits are against the manufacturer and not against the doctor.

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    Thus far no DES case has been tried to completion. Several have been settled by the manufacturers on the eve of the trial, generally for less than the full sum that a cancer victim would expect to receive. Crack growth through the thickness of thin-sheet Hydrided Zircaloy In recent years, the limits on fuel burnup have been increased to allow an increase in the amount of energy produced by a nuclear fuel assembly thus reducing waste volume and allowing greater capacity factors.

    As a result, it is paramount to ensure safety after longer reactor exposure times in the case of design-basis accidents, such as reactivity-initiated accidents RIA. Previously proposed failure criteria do not directly address the particular cladding failure mechanism during a RIA, in which crack initiation in brittle outer-layers is immediately followed by crack growth through the thickness of the thin-wall tubing. In such a case, the fracture toughness of hydrided thin-wall cladding material must be known for the conditions of through-thickness crack growth in order to predict the failure of high-burnup cladding.

    To achieve this goal, an experimental procedure was developed in which a linear hydride blister formed across the width of a four-point bend specimen was used to inject a sharp crack that was subsequently extended by fatigue pre-cracking. The electrical potential drop method was used to monitor the crack length during fracture toughness testing, thus allowing for correlation of the load-displacement record with the crack length. Elastic-plastic fracture mechanics were used to interpret the experimental test results in terms of fracture toughness, and J-R crack growth resistance curves were generated.

    Finite element modeling was performed to adapt the classic theories of fracture mechanics applicable to thick-plate specimens to the case of through-thickness crack growth in thin-sheet materials, and to account for non-uniform crack fronts. Finally, the hydride microstructure was characterized in the vicinity of the crack tip by. From considerations of the results of experimental simulations of non-axial placement of fuel elements in process tubes and in-reactor experience, it is concluded that the ultimate outcome of a charging error which results in operation with one or more unsupported fuel elements in a K Zircaloy -2 process tube would be multiple fuel failure and failure of the process tube.

    The outcome of the accident is determined by the speed with which the fuel failure is detected and the reactor is shut down. The highest probability for fission product release occurs during the discharge of failed fuel elements, when a small fraction of the exposed uranium of the fuel element may be oxidized when exposed to air before the element falls into the water-filled discharge chute. The confinement and fog spray facilities were installed to reduce the amount of fission products which might escape from the reactor building after such an event.

    Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature HT for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. The atomic structure of the interface reveals the presence of Zr Fe, Cr 2 Laves phase, displaying both C14 and C15 structure.

    After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation. Fabrication and testing of U—7Mo monolithic plate fuel with Zircaloy cladding. The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Phenomenological analysis revealed that in the steam starvation condition, i. Microstructural changes of oxide demonstrate that when the weight gain exceeds a certain critical value, massive hydriding takes place due to the significant localized crack development within the oxide, which possibly simulates the secondary hydriding failure in a defective fuel operation.

    In other words, in a real defective fuel operation incident, the secondary failure is initiated only when both steam starvation and oxide degradation conditions are simultaneously met. Therefore, it is concluded that the indispensable time for the critical oxide growth primarily determines the triggering time of massive hydriding failure.

    Estimation of ring tensile properties of steam oxidized Zircaloy -4 fuel cladding under simulated LOCA condition. Shriwastaw, R. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test RTT. The microstructure was correlated with the mechanical properties. The yield strength YS and ultimate tensile strength UTS increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation.

    Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0. Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance.

    An appropriately defined minimum margin of safety was estimated for each of these criteria. Results show that the coating deposited at V shows better high temperature oxidation resistance behavior, at the same time, Al elements contained in the coating is of the highest amount, meanwhile, the adhesion strength of the coating to the substrate is the highest, which is 33N. As the bias increases, high temperature oxidation resistance behavior of the coating weakens first and then increases, the amount of large particles on the surface of the coating increases first and then decreases whereas the density of the coating decreases first and then increases, and adhesion strength of the coating to the substrate increases first and then weakens.

    The coating's quality is relatively poor when the bias is V. The effect of plastic strain on the evolution of crystallographic texture in Zircaloy The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy -2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at K and K.

    Incremental transverse platic strain ratios R were mesured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis.

    The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. Maladie des vibrations. Developpement de mesures non destructives, par ondes ultrasonores, d'epaisseurs de fronts de solidification dans les reacteurs metallurgiques. Dans les cuves d'electrolyse d'aluminium, le milieu de reaction tres corrosif attaque les parois de la cuve, ce qui diminue leur duree de vie et augmente les couts de production.

    Le talus, qui se forme sous l'effet des pertes de chaleur qui maintiennent un equilibre thermique dans la cuve, sert de protection naturelle a la cuve. Son epaisseur doit etre controlee pour maximiser cet effet. Advenant la resorption non voulue de ce talus, les degats generes peuvent s'evaluer a plusieurs centaines de milliers de dollars par cuve.

    Aussi, l'objectif est de developper une mesure ultrasonore de l'epaisseur du talus, car elle serait non intrusive et non destructive. La precision attendue est de l'ordre du centimetre pour des mesures d'epaisseurs comprenant 2 materiaux, allant de 5 a 20 cm. Cette precision est le facteur cle permettant aux industriels de controler l'epaisseur du talus de maniere efficace maximiser la protection des parois tout en maximisant l'efficacite energetique du procede , par l'ajout d'un flux thermique.

    Cependant, l'efficacite d'une mesure ultrasonore dans cet environnement hostile reste a demontrer. Les travaux preliminaires ont permis de selectionner un transducteur ultrasonore a contact ayant la capacite a resister aux conditions de mesure hautes temperatures, materiaux non caracterises Differentes mesures a froid traite par analyse temps-frequence ont permis d'evaluer la vitesse de propagation des ondes dans le materiau de la cuve en graphite et de la cryolite, demontrant la possibilite d'extraire l'information pertinente d'epaisseur du talus in fine.

    Fort de cette phase de caracterisation des materiaux sur la reponse acoustique des materiaux, les travaux a venir ont ete realises sur un modele reduit de la cuve. Robinson Nuclear Reactor. This study investigates the hydride rim microstructure at the metal-oxide interface of Zircaloy -4 cladding segment removed from H. A complex stacking and orientation of hydride platelets has been observed below the sub-oxide layer.

    Furthermore, radial hydride platelets have been observed. Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding. Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. Because Zry-4 and U- Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded.

    On the corrosion behavior of zircaloy -4 in spent fuel pools under accidental conditions. After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically.

    This was the case in Fukushima Daiichi in March Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. The critical parameter is the presence of chloride ions. SCE , leaving a shorter stable passive range for the Zr-4 cladding tubes. Baris, A.

    A high burn-up Zircaloy -2 cladding is characterised in order to correlate its microstructure and composition to the change of oxidation and hydrogen uptake behaviour during long term service in the reactor. After 9 cycle of service, the chemical analysis of the cladding segment shows that most secondary phase particles SPPs have dissolved into the matrix.

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