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Sample records for zircaloy dosage du

  1. Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

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    Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

  2. Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4; Filage et etirage du zircaloy 2. Realisation des tubes de force pour EL-4

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    Thevenet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Buffet, J [Cefilac (France)

    1964-07-01

    The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the {phi} = 340 ingot into {phi} = 220 billets, cutting into lengths and hot drilling at {phi} = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes ({phi} =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [French] Les auteurs donnent un resume succint des proprietes physiques mecaniques et chimiques du zircaloy 2 en ce qui concerne la transformation de cet alliage. Filage: Apres quelques generalites sur le filage et le cofilage, dont une comparaison entre les resistances a la deformation des metaux de gainage et du zircaloy 2, on etudie successivement: - les

  3. Spectrographic determination of chlorine and fluorine; Dosage du chlore et du fluor par spectrographie d'emission en atmosphere inerte

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    Contamin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1965-04-01

    Experimental conditions have been investigated in order to obtain the highest sensitivity in spectrographic determination of chlorine and fluorine using the Fassel method of excitation in an inert atmosphere. The influence of the nature of the atmosphere, of the discharge conditions and of the matrix material has been investigated. The following results have been established: 1. chlorine determination is definitely possible: a working curve has been drawn between 10 {mu}g and 100 {mu}g, the detection limit being around 5 {mu}g; 2. fluorine determination is not satisfactory: the detection limit is still of the order of 80 {mu}g. The best operating conditions have been defined for both elements. (author) [French] Nous avons recherche quelles etaient les conditions permettant d'obtenir la meilleure sensibilite dans le dosage spectrographique du chlore et du fluor par la methode d'excitation en atmosphere inerte (methode de Fassel). Nous avons etudie l'influence de l'atmosphere gazeuse, des conditions de la decharge et du materiau de pastillage. Les points suivants ont ete etablis: 1. le dosage du chlore est possible: une courbe de dosage a ete tracee entre 10 {mu}g et 100 {mu}g et la limite de detection est de l'ordre de 5 {mu}g; 2. le dosage du fluor n'est pas satisfaisant: la limite de detection obtenue etant encore de l'ordre de 80 {mu}g. Les conditions operatoires ont ete precisees pour ces deux elements. (auteur)

  4. Dosage plasmatique et globulaire du magnesium dans l'exploration ...

    African Journals Online (AJOL)

    Objectives: The allergic rhinitis represents a real public health problem. The goal of this survey is to value the interest of the dosage plasmatical and globular of magnesium in the diagnosis of the allergic rhinitis. Materials and methods : Analytic and prospective survey of 80 files, on one period of 4 years and 5 months (from ...

  5. Determination of radium in urine; Dosage du radium dans l'urine

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    Fourniguet, H; Jeanmaire, L; Jammet, H [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    A procedure for the quantitative analysis of radium in urine is described. The radium is carried by a barium sulfate precipitate. The precipitate is mixed with zinc sulfide and the activity measured by scintillation counting. It is thus possible to detect an amount of radium less than 1 pico-curie in the sample. (author) [French] Cet article decrit une technique de dosage du radium dans l'urine. Le radium entraine par un precipite de sulfate de baryum est compte par scintillation apres melange du precipite avec du sulfure de zinc. Cette methode permet de deceler moins de 1 picocurie de radium dans l'echantillon. (auteur)

  6. Dosage of strontium 90 in human bone ashes; Dosage du strontium 90 sans les cendres d'os humain

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    Patti, F; Jeanmaire, L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    The determination of {sup 90}Sr in bones by dosage of its daughter product {sup 90}Y is a 4-step process: 1) elimination of the phosphate ions by precipitation of the Ca and Sr as oxalate in the presence of acid; 2) reduction in the calcium concentration to a suitable level by the addition of a known volume of nitric acid (a single precipitation is sufficient), the precipitation yield of the strontium nitrate is checked by the measurement of the amount of {sup 85}Sr added as tracer; 3) purification by a yttrium hydroxide precipitation; 4) extraction at equilibrium of the {sup 90}Y which is counted to give the concentration. By using 50 gm of ash it is possible to detect about 0.1 pCi of {sup 90}Sr per gram of calcium. The advantages of this technique: -) treatment of a large quantity of bone ash -) the use of a small volume of nitric acid (less than 2 ml/g of ash, and -) the various operations present no difficulty. (authors) [French] Determination du Sr dans les os par dosage de son produit de filiation {sup 90}Y. Principe du dosage: 1 - Eliminer les ions phosphates par precipitation du calcium et du strontium sous forme d'oxalate en milieu acide. 2 - Reduire la concentration en calcium a un niveau convenable par addition d'un volume determine d'acide nitrique (une seule precipitation est necessaire). Le rendement de precipitation du nitrate de strontium est controle par la mesure de {sup 85}Sr ajoute comme traceur. 3 - Purifier par une precipitation d'hydroxyde d'yttrium. 4 - Extraire a l'equilibre l'{sup 90}Y qui eat compte pour determiner le {sup 90}Sr. En traitant 50 g de cendre, il est possible de deceler de l'ordre de 0,1 pCi de {sup 90}Sr par gramme de calcium. Les 3 avantages de cette technique: 1 - traitement d'une quantite importante de cendres d'os, 2 - emploi d'un faible volume d'acide nitrique (moins de 2 ml/g de cendres), et 3 - les diverses operations ne presentent aucune difficulte.

  7. Determination of tritium by counting; Dosage du tritium par comptage

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    Schott, R; Froment, G; Pinson, J; Genty, C [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1968-07-01

    Ionisation chamber assay of tritium in any gaseous mixture is a simple, fast and accurate method. We used the method of relative determination by comparison to a standard rather than the method of absolute assay in which case the constants are known with too little accuracy. The efficiency of the chamber was studied in connection to the pressure inside the chamber and its total volume. The calibration is linear in the range we are taking into account (1 to 80 millicuries). The reproducibility of the method is good: 13 runs gave a coefficient of variation of 1.6 per cent. The relative accuracy was found equal to {+-} 1.3 per cent. To end the paper, we describe in detail the apparatus and the ways of proceedings. (authors) [French] Le comptage du tritium par chambre d'ionisation est une methode simple, rapide et precise pour determiner la teneur en tritium d'un melange gazeux quelconque. Nous avons prefere utiliser la methode de determination relative par rapport a un etalon car, dans le cas d'une determination absolue, les constantes sont connues avec une trop grande incertitude. L'efficacite de la chambre a ete etudiee en fonction de la variation de la pression d'argon a l'interieur de la chambre et du volume total, de cette derniere. L'etalonnage s'est revele lineaire dans le domaine de mesures qui nous interessaient (1 a 80 millicuries). La reproductibillte de la methode est tres bonne, le coefficient de variation pour une serie de 13 essais etant de 1,6 pour cent, quant a la precision relative, elle a ete evaluee a {+-} 1,3 pour cent. Pour terminer, nous donnons une description detaillee de l'appareillage utilise et du mode operatoire suivi. (auteurs)

  8. Oligo cyclic plastic fatigue of Zircaloy-4 under vacuum and in iodinated methanol; Fatigue plastique oligocyclique du Zircaloy-4 sous vide et dans le methanol iode

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    Beloucif, A.

    1995-01-01

    Our study was bound to the Zircaloy-4 fuel can damage in PWR type reactors. The topic was the damage mechanisms of Zircaloy-4 by oligo-cyclic plastic fatigue in inert atmosphere and in iodinated methanol. The oligo-cyclic plastic fatigue tests, under vacuum, were performed with steady plastic deformation and deformation speed. The corrosion fatigue tests in iodinated methanol put to the fore one obvious harmful part of iodine on Zircaloy-4 resistance to cyclic solicitations. The observations proved the existence of a very strong synergic effect between cyclic mechanical damage and corrosion. (MML). 84 refs., 117 figs., 3 tabs.

  9. Determination of strontium 90 in milk Ash; Dosage du strontium 90 dans les cendres de lait

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    Ballada, J; Jeanmaire, L [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

    1968-07-01

    This report describes a method of determination of {sup 90}Sr in milk ashes by extraction of {sup 90}Y in TBP. The tests which led to the choice of the operating process are presented together with tire result of an intercomparison. (author) [French] Le document decrit une methode de dosage du {sup 90}Sr dans les cendres de lait par extraction de {sup 90}Y dans le TBP. De plus, les auteurs rapportent les essais qui ont conduit au choix du mode operatoire presente, ainsi que les resultats d'une intercomparaison. (auteur)

  10. Influence of temperature on the Zircaloy-4 plastic anisotropy; Influence de la temperature sur l`anisotropie plastique du Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Limon, R.; Bechade, J.L.; Lehmann, S.; Maury, R.; Soniak, A. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Technologies Avancees; Mardon, J.P. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

    1995-12-31

    In order to improve the comportment modelling of PWR fuel pin, and more precisely their canning tubes, Framatome and the CEA have undertake an important study program of Zircaloy-4 mechanical properties. It includes in particular the study of the plasticity between 20 and 400 degree Celsius. This material being not isotropic because of the zirconium hexagonal crystal network and the texture presented by the canning tubes, its plastic anisotropy has been measured. The obtained results for the canning in *slack* and recrystallized before irradiation Zircaloy-4 are presented and the deformation systems able to explain the observed anisotropy is researched. (O.L.). 6 refs., 4 figs., 1 tab.

  11. Dosage of cesium 137 in radioactive wastes by the application of sodium tetraphenylborate; Dosage du cesium 137 dans les effluents radioactifs par le tetraphenylborate de sodium

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    Testemale, G; Girault, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    A simple technique of the dosage of {sup 137}Cs has been developed. The technique consists in the formation of cesium tetraphenyl borate, followed by a double extraction with isoamyl acetate, and washing of the organic phase. The counting of known parts of the cesium solution assaying of its purity by {gamma} spectrometry enable the determination of the {sup 137}Cs. The yield is about 98 per cent. (authors) [French] Une technique simple du dosage du {sup 137}Cs a ete mise au point. Elle consiste en une double extraction du tetraphenylborate de cesium forme par l'acetate d'isoamyle suivie d'un lavage de la phase organique. Des comptages sur des parties aliquotes de la solution de cesium et un controle de purete par spectrometrie {gamma} permettent la determination de cet element. Rendement: environ 98 pour cent. (auteurs)

  12. Thermomechanical treatment of {beta}-treated Zircaloy-4 within the upper {alpha}-range; Traitements thermomecaniques dans le haut domaine {alpha} du zircaloy-4 trempe-{beta}

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    Chauvy, C

    2004-09-15

    Zircaloy-4 is a Zr base alloy mainly used for nuclear applications. This study deals with its behaviour during forming, with a special interest for physical mechanisms acting in the upper {alpha}-range. The {beta}-treated Zircaloy-4 is first described in terms of microstructure and texture. The {alpha} plates are organised in colonies with alternating crystallographic orientations. The Burgers relationships have also been checked. The mechanical behaviour shows two distinct domains (with or without hardening). This could be linked to interactions between the solutes (Sn, O) and mobile dislocations. The evolution of crystallographic texture is characterised by X-ray diffraction. At 550 C, twinning is shown to be the main deformation mechanism under specific experimental conditions. Globularization of the initial lamellar structure is identified as a continuous dynamic recrystallization process, more efficient at higher temperature. High strains are necessary to achieve this phenomenon but meta-dynamic recrystallization can also be used to obtain an equiaxed microstructure after limited strains. (author)

  13. Radio-chemical dosage of {sup 90}Sr in large volumes of drinking water; Dosage radiochimique du {sup 90}Sr sur des volumes importants d'eaux potables

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    Jeanmaire, L; Patti, F; Bullier, D [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-07-01

    I. Principle of the method: 1. Fixing on a resin of all the cations present in the water. 2. Elution using 5 N nitric acid and precipitation of strontium as the carbonate. 3. Concentration of the strontium using the fuming nitric acid method. 4. Purification of the strontium on a resin by selective elution with ammonium citrate. 5. The strontium-90 is measured by separation at the {sup 90}Y equilibrium in the form of the oxalate which is then counted. II. Advantages of the method The concentration of the radio-activity starting from large volumes (100 l) is generally tedious but this method which makes use of a fixation on a cationic resin makes it very simple. The rest of the method consists of a series of simple chemical operations using ion-exchange on resins and coprecipitation. Finally, it is possible to dose stable strontium. (authors) [French] I. Principe du dosage 1. Fixation sur resine de tous les cations presents dans l'eau, 2. Elution par l'acide nitrique 5 N et precipitation du strontium sous forme de carbonate. 3. Concentration du strontium par la methode a l'acide nitrique fumant. 4. Purification du strontium sur resine par elution selective au citrate d'ammonium. 5. Le strontium-90 est dose par separation a l'equilibre du {sup 90}Y sous forme d'oxalate qui est compte. II. Interet de la methode La concentration de la radioactivite a partir de volumes importants (100 l) est generalement fastidieuse, la technique proposee rend cette phase tres simple en utilisant une fixation sur resine cationique. Le reste de la technique est une suite d'operations chimiques simples a realiser, faisant appel a l'echange d'ions sur resine et a la coprecipitation. Enfin, il est possible de realiser le dosage du strontium stable. (auteurs)

  14. Spectrographic measurement of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Artaud, J; Cittanova, J [Commissariat a l' Energie Atomique, Service d' Analyses et Recherches Chimiques Appliquees, Saclay (France). Centre d' Etudes Nucleaires; Crehange, G; Frequelin, S [Commissariat a l' Energie Atomique, Dir. des Applications Militaires, Service Chimie, Saclay (France). Centre d' Etudes Nucleaires; Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1961-07-01

    We describe here a method for the spectrographic determination of beryllium on filters which is valid for amounts varying between 0,01 and 30 {mu}g of beryllium and which is independent of the nature of the beryllium compound involved. This is a flux method (graphite-lithium carbonate mixture), the excitation being by a direct current arc. (author) [French] Nous decrivons ici, une methode de dosage spectrographique de beryllium sur filtre, valable pour des teneurs comprises entre 0,01 et 30 {mu}g de beryllium et independante de la nature du compose de beryllium a doser. C'est une methode de 'flux' (melange graphite-carbonate de lithium) l'excitation etant un arc a courant continu. (auteur)

  15. Determination of natural thorium in urines; Dosage du thorium dans les urines

    Energy Technology Data Exchange (ETDEWEB)

    Jeanmaire, L; Jammet, H [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    A procedure for the quantitative analysis of thorium in urine is described. After precipitation with ammonium hydroxide, dissolution of the precipitate, extraction at pH 4-4.2 with cupferron in chloroformic solution and mineralization, a colorimetric determination of thorium with thorin is performed. It is thus possible to detect about 2 {gamma} of thorium in the sample. (author) [French] Cet article decrit une technique de dosage du thorium dans l'urine. Apres precipitation par l'ammoniaque, remise en solution, extraction a pH 4-4,2 par le cupferron en solution chloroformique et mineralisation, le thorium est dose par colorimetrie avec le thorin. Cette methode permet de deceler environ 2 {gamma} de thorium dans l'echantillon. (auteur)

  16. Spectrographic determination of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Soudain, G; Morawek, T [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    Since the apparatus for continuous determination of beryllium is not yet perfect, a discontinuous method has been developed. The air to be analysed is filtered, and the dust laden filter is dissolved in a mixture of sulphuric and nitric acid. The pH and the conductivity of the solution obtained were adjusted to standard values, and it was then analysed spectro-graphically by the rotating sector method. Up to 0.01 x 10{sup 6} of Be per cm{sup 3} of solution can be detected. The precision is of the order of 10 per cent. (author) [French] Les appareils de dosage du beryllium en continu n'etant pas encore suffisamment au point, on a elabore une methode discontinue. L'air a analyser est filtre et le filtre charge de poussieres est mis en solution par une attaque sulfo-nitrique. La solution obtenue est normalisee par ajustage de son PH et de sa conductivite puis analysee spectrographiquement par la methode du disque tournant. On peut detecter jusqu'a 0,01.10{sup 6} de Be par cm{sup 3} de solution. La precision est de l'ordre de 10 pour cent. (auteur)

  17. Spectrographic determination of chlorine and fluorine; Dosage du chlore et du fluor par spectrographie d'emission en atmosphere inerte

    Energy Technology Data Exchange (ETDEWEB)

    Contamin, G. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1965-04-01

    Experimental conditions have been investigated in order to obtain the highest sensitivity in spectrographic determination of chlorine and fluorine using the Fassel method of excitation in an inert atmosphere. The influence of the nature of the atmosphere, of the discharge conditions and of the matrix material has been investigated. The following results have been established: 1. chlorine determination is definitely possible: a working curve has been drawn between 10 {mu}g and 100 {mu}g, the detection limit being around 5 {mu}g; 2. fluorine determination is not satisfactory: the detection limit is still of the order of 80 {mu}g. The best operating conditions have been defined for both elements. (author) [French] Nous avons recherche quelles etaient les conditions permettant d'obtenir la meilleure sensibilite dans le dosage spectrographique du chlore et du fluor par la methode d'excitation en atmosphere inerte (methode de Fassel). Nous avons etudie l'influence de l'atmosphere gazeuse, des conditions de la decharge et du materiau de pastillage. Les points suivants ont ete etablis: 1. le dosage du chlore est possible: une courbe de dosage a ete tracee entre 10 {mu}g et 100 {mu}g et la limite de detection est de l'ordre de 5 {mu}g; 2. le dosage du fluor n'est pas satisfaisant: la limite de detection obtenue etant encore de l'ordre de 80 {mu}g. Les conditions operatoires ont ete precisees pour ces deux elements. (auteur)

  18. Dosage of fission products in irradiated fuel treatment effluents (radio-chemical method); Dosage des produits de fission dans les effluents du traitement des combustibles irradies (methode radiochimique)

    Energy Technology Data Exchange (ETDEWEB)

    Auchapt, J [Commissariat a l' Energie Atomique, Marcoule (France). Centre d' Etudes Nucleaires

    1966-07-01

    The dosage methods presented here are applicable to relatively long-lived fission products present in the effluents resulting from irradiated fuel treatment processes (Sr - Cs - Ce - Zr - Nb - Ru - I). The methods are based on the same principle: - addition of a carrying-over agent - chemical separation over several purification stages, - determination of the chemical yield by calorimetry - counting of an aliquot liquid portion. (author) [French] Les methodes de dosage presentees concernent les produits de fission a vie relativement longue presents dans les effluents de traitement des combustibles irradies (Sr - Cs - Ce - Zr - Nb - Ru - I). Elles sont toutes basees sur le meme principe: - addition d'entraineur, - separation chimique en plusieurs stades de purification, - determination du rendement chimique par calorimetrie, - comptage d'une aliquote liquide. (auteur)

  19. Dosage of plutonium by isotopic dilution in irradiated fuels; Dosage du plutonium par dilution isotopique dans les combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Lucas, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Plutonium determination in irradiated fuels has been carried out for several years by isotopic dilution by Sebaci and SSM in collaboration. SECACI has made available to the SSM the necessary space and equipment in its Fontenay laboratories. This work has shown the importance of the valency cycle which should make it possible to obtain a uniform isotopic distribution in sample tracer mixtures, and also a satisfactory U/Pu separation. Now it has been noticed that the presence of an excess of uranium considerably modifies the oxidation-reduction reaction kinetics of the plutonium. We have therefore been led to change certain parts of the operational technique so as to have an efficient cycle and to thereby improve the U/Pu separation; the stability of the thermionic emission of the plutonium, connected to the quantity of residual uranium, has at the same time been improved and we can now carry out more precise isotopic analyses. We have also tried to eliminate as far as possible the isotopic contaminations by:using a more rational operational method; the equipment used has been the object of a special study. The evaporations are carried out so as to prevent the formation of saturated vapours inside the glove box. The material which cannot be changed after each operation is carefully cleaned every time a new sample is treated. With this technique, a second calibration of the tracer T{sub 2} has been undertaken using a new standard solution. This solution has been prepared very carefully by weighing uranium and plutonium of known chemical purity, and we believe that it can be guaranteed to be a good reference solution. The value of the {sup 233}U/{sup 242}Pu ratio of the tracer has been obtained with a relative accuracy of 0,5 per cent. This modified method is now being applied to the analysis of rods irradiated in G-3. (author) [French] La determination du plutonium par dilution isotopique dans les combustibles irradies est pratiquee depuis plusieurs annees en

  20. Effets de la radiolyse de l'air humide et de l'eau sur la corrosion de la couche d'oxyde du Zircaloy-4 oxydé

    OpenAIRE

    Guipponi , Claire

    2009-01-01

    Pas de résumé donné.; Les Colis Standards de Déchets Compactés (CSD-C) sont des déchets issus du retraitement des assemblages de combustibles nucléaires. Ils sont en partie constitués des gaines oxydées de Zircaloy-4. Ces pièces métalliques sont cisaillées avant d'être placées dans un étui en acier et compactées sous forme de galettes. Ces galettes contiennent des traces de produits d'activation, de produits de fission et d'actinides présents à la surface du Zircaloy-4 oxydé. Dans l'hypothèse...

  1. Contribution to the study of corrosion of zirconium and zircaloy-2 in superheated steam at 400 deg C (105 kg /cm{sup 2}); Contribution a l'etude de la corrosion du zirconium et du zircaloy-2 dans la vapeur d'eau surchauffee a 400 deg C (105 kg /cm{sup 2})

    Energy Technology Data Exchange (ETDEWEB)

    Coriou, H; Gauduchau, J; Grall, L; Hure, J; Pelras, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The corrosion kinetics of zircaloy-2 in water and steam at temperatures between 300 deg. C and 400 deg. C are represented by a curve sharply divided into two stages separated by a so-called transition point. After a first period of decreasing corrosion rate there follows a second period with much faster kinetics in which the speed is constant. After carrying out a methodical study of the corrosion of 'zircaloy-2 in the form of sheets and tubes. We have demonstrated, at 400 deg. C in steam, a systematic anomaly which appears at the transition point. The curve presents three quite distinct points; after the first period a fast corrosion is observed, followed by a third period at a slower speed. This leads us to believe that there may be not a single point but a transition zone, separating two types of kinetic behaviour and corresponding to modifications in the properties of the oxide layer. After this readjustment period a new corrosion law is established, lasting a considerable time, the corrosion speed being slower than that indicated so far. A study of the morphology of the oxide films which develop under these conditions has demonstrated the special part played by mechanical, physical and metallurgical factors in the case of zirconium. Deep penetration of oxide can thus show up on the inner wall of hammer-hardened tubes. Simultaneously a very considerable hydride formation occurs in the metal. (author) [French] La cinetique de corrosion du zircaloy-2 dans l'eau et la vapeur a des temperatures comprises entre 300 et 400 deg. C est representee par une courbe a deux periodes separees par un point singulier appele point de transition. A une premiere periode a vitesse de corrosion decroissante, succede une deuxieme periode a cinetique beaucoup plus rapide dont la vitesse est constante. Apres une etude systematique de la corrosion du zircaloy-2 sous forme de toles et de tubes, nous avons mis en evidence a 400 deg. C, dans la vapeur, une anomalie systematique qui se

  2. A rapid method of dosing plutonium in radioactive effluents; Methode de dosage rapide du plutonium dans les effluents radioactifs (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Scheidhauer, J; Messainguiral, L [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

    1961-07-01

    The plutonium is first separated by a lanthanum fluoride precipitation. The precipitated fluorides are dissolved in normal nitric acid solution in the presence of aluminium nitrate. The plutonium transformed to the tetravalent state is then extracted with thenoyltrifluoroacetone and returned to the aqueous phase with 10 N nitric acid. After evaporation on a watch glass the residue is calcined on a Meker burner and counted using a counting system fitted with a zinc sulphide scintillator. When necessary, the calcium is eliminated at the beginning of the dosage by a fluoride precipitation, the plutonium being oxidised to the valency IV. (authors) [French] Le plutonium est d'abord separe par entrainement au fluorure de lanthane. Le precipite des fluorures est remis en solution en milieu acide nitrique normal, en presence de nitrate d'aluminium. Le plutonium amene a la valence IV est alors extrait par la thenoyltrifluoroacetone et remis en phase aqueuse dans l'acide nitrique 10 N. Apres evaporation sur verre de montre, le residu est calcine sur bec Meker et compte sur un ensemble de comptage equipe d'un scintillateur au sulfure de zinc. Lorsque cela est necessaire, le calcium est elimine, au debut du dosage, par precipitation du fluorure, le plutonium etant oxyde a la valence VI. (auteurs)

  3. Effect of hydrogen and hydrides on the viscoplastic behaviour of the recrystallized zircaloy-4; Effet de l'hydrogene et des hydrures sur le comportement viscoplastique du zircaloy-4 recristallise

    Energy Technology Data Exchange (ETDEWEB)

    Rupa, N

    2000-04-15

    Zircaloy-4 is the main material of PWR fuel assemblies. In service as during the storage, the integrity of these compounds has to be guaranteed in spite of the presence of hydrogen (in solution in the zirconium matrix) and of hydrides (which precipitate when the amount of hydrogen is higher than the solubility limit). The aim of this work is to characterize the hydrogen and hydrides effect on the viscoplastic behaviour of the non irradiated recrystallized zircaloy-4. The presence of hydrogen in solid solution induces a decrease of the mechanical properties: the creep kinetics are then increased and the tensile stresses decreased. This decrease is particularly visible in conditions of oxygen/dislocations dynamic interactions (revealed on the material without hydrogen). The advanced hypothesis, strengthened by the atomic simulation results, is that the hydrogen facilitates the dislocations movement, in diminishing the effects of anchoring by the interstitials, and/or in increasing the intrinsic mobility of dislocations. The hydrides effect induces a hardening of the material (decrease of the creep kinetics, increase of the tensile stresses and of the relaxed stresses) compensating the decrease by hydrogen. The hardening mechanism is due to an increase of the internal constraints, determined by load-unload tests. For the very weak plastic deformations, the hydrides are an obstacle to the dislocations gliding. They are then passed (that corresponds to a saturation of the internal constraint). The TEM observations as well as the results obtained on the titanium indicate that the precipitates are then submitted to a deformation mechanism. (O.M.)

  4. Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4; Influence de l'orientation des hydrures sur les modes de deformation, d'endommagement et de rupture du zircaloy-4 hydrure

    Energy Technology Data Exchange (ETDEWEB)

    Racine, A

    2005-09-15

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

  5. Influence de l'orientation des hydrures sur les modes de déformation, d'endommagement et de rupture du Zircaloy-4 hydruré.

    OpenAIRE

    Racine , Aude

    2005-01-01

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300°C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embitterment depends on many parameters, among which hydrogen content and orientation of hydrid...

  6. Quantitative analysis of strontium 90 in the radioactive wastes by means of thenoyltrifluoroacetone; Dosage du strontium 90 dans les effluents radioactifs par le thenoyltrifluoroacetone

    Energy Technology Data Exchange (ETDEWEB)

    Testemale, G; Leredde, J L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1970-07-01

    A simple method of analysing the quantity of {sup 90}Sr has been perfected. It consists in a double extraction of it by means of thenoyltrifluoroacetone and tributylphosphate in tetrachloride of carbon followed by eliminating yttrium 90 by means of thenoyltrifluoroacetone in the benzene. Numberings on aliquot parts of wastes make the determination of that element possible. The yield is about 97 per cent. (authors) [French] Une technique simple du dosage du {sup 90}Sr a ete mise au point. Elle consiste en une double extraction par le thenoyltrifluoroacetone et le tributylphosphate dans le tetrachlorure de carbone, suivie de l'elimination de l'yttrium 90 par le thenoyltrifluoroacetone dans le benzene, Des comptages sur des parties aliquotes d'effluents permettent la dermination de cet element. Rendement environ 97 pour cent. (auteurs)

  7. Review of zircaloy oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, F.C. [Royal Military College of Canada, Kingston, Ontario (Canada); Lewis, B.J. [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2013-07-01

    This paper provides an overview of the kinetics for Zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. The effect of internal clad oxidation due to Zircaloy/UO{sub 2} interaction is also discussed. Low-temperature oxidation of Zircaloy due to water-side corrosion is further described. (author)

  8. Dosage du glyphosate par HPLC après extraction et dérivation à l'O ...

    African Journals Online (AJOL)

    Le glyphosate, premier herbicide utilisé au monde est une molécule difficile à quantifier par la chromatographie en phase liquide à haute performance (HPLC), eu égard à l'absence de chromophore dans sa structure. La chimie analytique est donc à la recherche perpétuelle de méthodes de détermination du glyphosate ...

  9. Precipitates in irradiated Zircaloy

    International Nuclear Information System (INIS)

    Chung, H.M.

    1985-10-01

    Precipitates in high-burnup (>20 MWd/kg U) Zircaloy spent-fuel cladding discharged from commercial boiling- and pressurized-water reactors have been characterized by TEM-HVEM. Three classes of primary precipitates were observed in the irradiated Zircaloys: Zr 3 O (2 to 6 nm), cubic-ZrO 2 (greater than or equal to 10 nm), and delta-hydride (35 to 100 nm). The former two precipitations appears to be irradiation induced in nature. Zr(Fe/sub x/Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/Ni/sub 1-x/) intermetallics, which are the primary precipitates in unirradiated Zircaloys, were largely dissolved after the high burnup. It seems, therefore, that the influence of the size and distribution of the intermetallics on the corrosion behavior may be quite different for the irradiated Zircaloys

  10. Zircaloy-sheathed element rods fitted with thermo-couples; Barre combustible a thermocouple gainee de zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Bernardy de Sigoyer, B; Jacques, F; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    In order to carry out thermal conductivity measurements on UO{sub 2} in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [French] Pour des mesures de conductivite thermique de l'UO{sub 2} dans des conditions voisines du fonctionnement des barres combustibles, il s'agissait de mesurer la temperature a l'interieur d'un element combustible gaine de zircaloy. Les prises de temperature sont faites par thermocouples du type Thermocoax, c'est-a-dire pourvu d'une gaine tres mince en inox ou inconel. Par ailleurs on sait que le soudage par fusion du zircaloy sur l'inox est impossible et que les brasures a haute temperature sont difficiles car tres agressives. La technique utilisee consiste a braser les thermocouples sur des pieces en inox relativement massives et de rapporter par soudage au bombardement electronique ces bouchons sur des raccords diffuses zircaloy-inox. Les proprietes de ces raccords diffuses et celles de joints brases ont ete etudiees; on expose egalement les diverses etapes de fabrication des containers. (auteurs)

  11. Plating on Zircaloy-2

    International Nuclear Information System (INIS)

    Dini, J.W.; Johnson, H.R.; Jones, A.

    1979-03-01

    Zircaloy-2 is a difficult alloy to coat with an adherent electroplate because it easily forms a tenacious oxide film in air and aqueous solutions. Procedures reported in the literature and those developed at SLL for surmounting this problem were investigated. The best results were obtained when specimens were first etched in either an ammonium bifluoride/sulfuric acid or an ammonium bifluoride solution, plated, and then heated at 700 0 C for 1 hour in a constrained condition. Machining threads in the Zircaloy-2 for the purpose of providing sites for mechanical interlocking of the plating also proved satisfactory

  12. Study of hydrogen migration produced during the corrosion of PWR reactors fuel cans in zircaloy 4 and zirconia; Etude du transport de l`hydrogene produit lors de la corrosion des gaines d`elements combustibles des reacteurs a eau sous pression dans la zircone et le zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Aufore, L

    1997-12-12

    The corrosion of Zircaloy-4-claddings by water from the primary circuit of nuclear power plant goes hand in hand with the release of hydrogen which penetrates the oxide and then the metal. This work focuses on the mechanisms of hydrogen transport in oxide and in metal. Hydrogen transport in oxide is studied on the basis of corrosion tests performed in the autoclave at 360 deg C. These tests are performed on Zircaloy-4 claddings under different chemical conditions (pure water, and pure water with lithium hydroxide). The distribution of hydrogen in oxide film is measured by SIMS. Hydrogen profiles in the oxide are dependent on the oxide microstructure and vary with oxidation time. These observations are confirmed by experiments in which some samples, previously oxidized in the autoclave, are immersed in heavy water. In the oxide layer, two zones are observed: one external porous zone and one internal dense zone. Deuterium diffusion coefficients in dense oxide are determined using SIMS profiles and Fischer diffusion model. Hydrogen transport in metal is also studied by means of gas-phase permeation experiments. These are set up at different temperature (400-500 deg. C) and under different hydrogen pressures and make it possible to determine the hydrogen diffusion coefficients in a Zircaloy-4 cladding under experimental conditions. All these results lead us to discuss of hydrogen transport evolution in cladding during oxidation. A model taking into account hydrogen transport in oxide and in metal, and the hydrides precipitations is proposed. (author) 110 refs.

  13. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  14. Analyses and quantitative determination of the strontium radioisotopes 89 and 90 in milk powder; Recherche et dosage des isotopes radioactifs 89 et 90 du strontium dans le lait en poudre

    Energy Technology Data Exchange (ETDEWEB)

    Jeanmaire, L; Michon, G [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The authors describe a procedure for the determination of the strontium radioisotopes 89 and 90. The concentration of strontium is made possible by the insolubility of its nitrate salt in strong nitric acid which allows the removal of greatest part of calcium. The purification is performed on a cation exchange column. The amount of radioisotope 90 is determined by means of its daughter product yttrium 90 necessary calibrations and computations are treated in special paragraphs. With regard to the reproducibility of the measurements, the fluctuations are less than 20 per cent. This seems satisfaction for such a technique which have great sensibility while being long and necessitative great carefulness. (author) [French] Les auteurs decrivent une technique de dosage des isotopes 89 et 90 du strontium. La concentration du strontium est assuree grace a l'insolubilite de son nitrate dans l'acide nitrique concentre, ce qui permet d'eliminer la plus grande partie du Ca. La purification se fait sur une colonne echangeuse de cations. L'isotope 90 est dose grace a son descendant l'yttrium 90. Les etalonnages et calculs necessaires font l'objet de paragraphes detailles. En ce qui concerne la reproductibilite des mesures, les fluctuations sont inferieures a 20 pour cent, ce qui semble satisfaisant devant la grande sensibilite de la methode qui reste cependant longue et delicate. (auteur)

  15. Simplified method for the determination of strontium-90 in large amounts of bone-ash; Methode simplifiee de dosage du strontium 90 sur des quantites importantes de cendres d'os

    Energy Technology Data Exchange (ETDEWEB)

    Patti, F; Jeanmaire, L [Commissariat a l' Energie Atomique, Fontenay-aux-roses (France). Centre d' Etudes Nucleaires

    1966-06-01

    The principle of the determination is based on a 3-step process: 1) concentrating the strontium by attacking the ash with nitric acid; 2) elimination of residual phosphoric ions by a double precipitation of strontium oxalate; and 3) extraction of yttrium 90, counted in the oxalate form. The advantages of the method: -) using simple techniques it makes it possible to process 50 g of ash; -) the initial concentration of strontium considerably reduces the volume of the solutions as well as the size of precipitates handled. Fuming nitric acid is used in a specially designed burette. (authors) [French] Le principe du dosage repos sur un procede en 3 etapes: 1) Concentration du strontium par une attaque nitrique des cendres; 2) Elimination des ions phosphoriques restants par une double precipitation de l'oxalate de strontium. 3) Extraction de l'yttrium 90, compte sous forme d'oxalate. Interet de la methode: dans des conditions techniques simples, elle permet le traitement de 50 g de cendres d'os; la concentration initiale du strontium reduit notablement le volume des solutions ainsi que l'importance des precipites manipules. L'acide nitrique fumant est utilise par l'intermediaire d'une burette specialement concue a cet effet.

  16. Technical note concerning the use of cellulose ester filtering membranes in the determination of plutonium in urine; Note technique sur l'utilisation des membranes filtrantes d'esters de cellulose dans le dosage du plutonium dans les urines

    Energy Technology Data Exchange (ETDEWEB)

    Harduin, J C; Montels, P [Commissariat a l' Energie Atomique, la Hague (France)

    1968-07-01

    During the last stage of the determination of plutonium in biological media, cellulose ester filtering membranes are used for collecting, with the help of a special device, the very fine precipitate resulting from the co-precipitation of plutonium and lanthanum fluorides. The membranes are then dried, and stuck on to flat watch-glasses for a {alpha} counting. A method is then given for purifying the lanthanum so as to keep the background noise during counting as low as possible. (author) [French] Dans la phase terminale du dosage du plutonium dans les milieux biologiques, on utilise les membranes filtrantes d'esters de cellulose pour recueillir, a l'aide d'un dispositif particulier, le precipite tres tenu resultant de la co-precipitation plutonium-lanthane sous forme de fluorure - Les membranes sont ensuite sechees puis collees sur verre de montre plat avant d'etre passees au compteur alpha. Un mode de purification du lanthane est ensuite donne afin de ne pas augmenter le bruit de fond des appareils de comptage. (auteur)

  17. Zircaloy oxidation studies

    International Nuclear Information System (INIS)

    Prater, J.T.; Beauchamp, R.H.; Saenz, N.T.

    1985-06-01

    The oxidation kinetics of Zircaloy-4 in steam have been determined at 1300-2400 0 C. Growth of the ZrO 2 and α-Zr layers display parabolic behavior over the entire temperature range studied. A discontinuity in the oxidation kinetics at 1510 0 C causes rates to increase above those previously established by the Baker-Just relationship. This increase coincides with the tetragonal-to-cubic phase transformation in ZrO/sub 2-x/. No discontinuity in the oxide growth rate is observed upon melting of Zr(0). The effects of temperature gradients have been taken into account and corrected values representative of near-isothermal conditions have been computed

  18. Dosage du mercure dans le gaz naturel par absorption atomique sans flammes Mercury Titration in Natural Gas by Flameless Atomic Absorption

    Directory of Open Access Journals (Sweden)

    La Villa F.

    2006-11-01

    Full Text Available Cet article présente la méthode mise au point par l'Institut Français du Pétrole pour déterminer par absorption atomique sans flamme, les traces de mercure métallique contenu dans un gaz naturel. La méthode d'analyse nécessite une extraction du mercure soit sous forme d'ion mercurique en faisant passer le gaz dans une solution oxydante, soit sous forme d'amalgame avec de l'or ou de l'argent. Le premier mode opératoire s'applique aux échantillons dont la concentration en mercure est supérieure à I ttg/Nm3, le second pour des concentrations inférieures à 5 pg/Nm3. Les seuils de détection sont respectivement 10 ng (en solution et 0,3 ng (en amalgame. La répétabilité pour 100 ng de mercure (en amalgame est de ± 7% pour une probabilité de.95 %. En conclusion, dans un échantillon de gaz naturel, compte tenu du volume des prélèvements effectués, il est possible de détecter des concentrations de l'ordre du nanogramme de mercure par mètre cube de gaz. This article describes the method developed by IFP using flameless atomic absorption to determine metallic mercury traces in a natural gas. The analyst method requires a mercury extraction either in the form of mercuric ions by making the gas pass through an oxidizing solution or in the form of an amalgam with gold or silver. The former operating method applies ta samples having a mercury concentration greater than I !ag/Nm3, and the latter for concentrations smaller than 5 (-Lg/Nm3. Detection thresholds are respectively 10 ng (in solution and 0.3 ng (in amalgam. The repeatability for 100 ng of mercury (in amalgam is ± 7 % with a probability of 95%. To conclude, in a sample of natural gas, considering the volume of the samples taken, it is possible ta detect concentrations in the vicinity of one nanogrom of mercury per cubic meter of gas.

  19. Stress corrosion crack growth in unirradiated zircaloy

    International Nuclear Information System (INIS)

    Pettersson, K.

    1978-10-01

    Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

  20. Radiation dosage

    Energy Technology Data Exchange (ETDEWEB)

    Finston, Roland [Health Physics, Stanford University, Stanford, CA (United States)

    1986-07-01

    Radiation dosage at Bikini Atoll is the result of current soil contamination, a relic of the nuclear weapons testing program of some 30 years ago. The principal contaminants today and some of their physical properties are listed: cesium-137, strontium-90, plutonium -239, 240 and americium-241. Cobalt-60 contributes less than 1 to the dose and is not considered significant. A resident of the atoll would accumulate radiation dose (rem) in two ways -- by exposure to radiation emanating from the ground and vegetation, and by exposure to radiation released in the spontaneous decay of radionuclides that have entered his body during the ingestion of locally grown foods. The latter process would account for some 90% of the dose; cesium-137 would be responsible for 0 90% of it. Since BARC's method of estimating dosage differs in some respects from that employed by the Lawrence Livermore National Laboratory (LLNL), (Ref.1, LLNL 1982) we are presenting our method in detail. The differences have two sources. First, the numbers used by BARC for the daily ingestion of radionuclides via the diet are higher than LLNL's. Second, BARC's calculation of dose from radionuclide intake utilizes the ICRP system. The net result is that BARC doses are consistently higher than LLNL doses, and in this respect are more conservative.

  1. Radiation dosage

    International Nuclear Information System (INIS)

    Finston, Roland

    1986-01-01

    Radiation dosage at Bikini Atoll is the result of current soil contamination, a relic of the nuclear weapons testing program of some 30 years ago. The principal contaminants today and some of their physical properties are listed: cesium-137, strontium-90, plutonium -239, 240 and americium-241. Cobalt-60 contributes less than 1 to the dose and is not considered significant. A resident of the atoll would accumulate radiation dose (rem) in two ways -- by exposure to radiation emanating from the ground and vegetation, and by exposure to radiation released in the spontaneous decay of radionuclides that have entered his body during the ingestion of locally grown foods. The latter process would account for some 90% of the dose; cesium-137 would be responsible for 0 90% of it. Since BARC's method of estimating dosage differs in some respects from that employed by the Lawrence Livermore National Laboratory (LLNL), (Ref.1, LLNL 1982) we are presenting our method in detail. The differences have two sources. First, the numbers used by BARC for the daily ingestion of radionuclides via the diet are higher than LLNL's. Second, BARC's calculation of dose from radionuclide intake utilizes the ICRP system. The net result is that BARC doses are consistently higher than LLNL doses, and in this respect are more conservative

  2. Effet du Pediococcus acidilactici sur le bilan lipidique sanguin du ...

    African Journals Online (AJOL)

    Les résultats relatifs aux performances zootechniques ont montré que l'addition du probiotique a amélioré significativement le gain de poids pendant la phase de croissance se traduisant par un indice de consommation meilleur. Les dosages du cholestérol total, des triglycérides, du HDL et du LDL ont été déterminés à la ...

  3. Obtaining zircaloy powder through hydriding

    International Nuclear Information System (INIS)

    Dupim, Ivaldete da Silva; Moreira, Joao M.L.

    2009-01-01

    Zirconium alloys are good options for the metal matrix in dispersion fuels for power reactors due to their low thermal neutron absorption cross-section, good corrosion resistance, good mechanical strength and high thermal conductivity. A necessary step for obtaining such fuels is producing Zr alloy powder for the metal matrix composite material. This article presents results from the Zircaloy-4 hydrogenation tests with the purpose to embrittle the alloy as a first step for comminuting. Several hydrogenation tests were performed and studied through thermogravimetric analysis. They included H 2 pressures of 25 and 50 kPa and temperatures ranging between from 20 to 670 deg C. X-ray diffraction analysis showed in the hydrogenated samples the predominant presence of ZrH 2 and some ZrO 2 . Some kinetics parameters for the Zircaloy-4 hydrogenation reaction were obtained: the time required to reach the equilibrium state at the dwell temperature was about 100 minutes; the hydrogenation rate during the heating process from 20 to 670 deg C was about 21 mg/h, and at constant temperature of 670 deg C, the hydride rate was about 1.15 mg/h. The hydrogenation rate is largest during the heating process and most of it occurs during this period. After hydrogenated, the samples could easily be comminuted indicating that this is a possible technology to obtain Zircaloy powder. The results show that only few minutes of hydrogenation are necessary to reach the hydride levels required for comminuting the Zircaloy. The final hydride stoichiometry was between 2.7 and 2.8 H for each Zr atom in the sample (author)

  4. Zircaloy 4 ingots' industrial fabrication

    International Nuclear Information System (INIS)

    Leyt, A.

    1987-01-01

    The technology developed for the industrial fabrication of Zircaloy-4 ingots is presented. According to the results obtained: a) the homogeneity of the ingots is analyzed, regarding the distribution of components (tin, iron, chromium, oxygen) and Brinell hardness as a function of different types of charge: zirconium sponge-recycling alloy material, sponge of zirconium-alloy; b) the distribution of the same parameters as a function of production is also analyzed. (Author)

  5. Study of the use of an electric discharge for hollow cathodes used as optical excitation sources in the spectrographic measurement of fluorine in thorium, uranium and plutonium; Etude de l'utilisation de la decharge electrique en cathode creuse comme source d'excitation optique pour le dosage spectrographique du fluor dans le thorium, l'uranium et le plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Bufpereau, M; Crehange, G; Poublan, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Previous works and phenomena concerned with a hollow cathode excitation are reviewed. Experiments aimed specially on the determination of the best conditions for an analysis of fluorine in oxides-metals and solutions. In that purpose, several factors have been pointed out. One started some researches about others elements that fluorine. Carrying fluorine into discharge and excitation have been more specially studied. A quantitative analysis method is given. The analysis limit is 45 ppm about but the detection limit is 5 ppm about. As a conclusion, various ways for optical excitation of fluorine are reviewed as other analytical possibilities a hollow cathode discharge offers. (authors) [French] On rappelle les travaux effectues jusqu'alors ainsi que les phenomenes mis en jeu dans l'excitation cathode creuse. Les experiences effectuees ont eu pour but essentiel la determination des conditions optima du dosage du fluor dans les oxydes, metaux et solutions. Pour cela de nombreux facteurs ont ete mis en evidence. Certaines etudes concernant d'autres elements que le fluor ont ete amorcees. Le passage du fluor dans la decharge et son excitation ont ete plus particulierement etudies. Une methode d'analyse quantitative est degagee, la limite de dosage est de l'ordre de 45 ppm, la limite de detection de 5 ppm. En conclusion, on passe en revue les differentes methodes d'excitation optique du fluor ainsi que les autres possibilites analytiques que peut offrir la cathode creuse. (auteurs)

  6. Ductility loss of ion-irradiated zircaloy-2 in iodine

    International Nuclear Information System (INIS)

    Shimada, M.; Terasawa, M.; Yamamoto, S.; Kamei, H.; Koizumi, K.

    1981-01-01

    An ion bombardment simulation technique for neutron irradiation was applied to 'thick' materials to study the effect of radiation damage on the ductility change in Zircaloy-2 in an iodine environment. Specimens were prepared from actual cladding tubes and, prior to the irradiation, they were heat-treated in vacuo at 450, 580, and 700/degree/C for 2 h. Irradiation was performed by 52-MeV alpha particles up to the 0.32 displacements per atom (dpa) at 340/degree/C. Ductility loss begins to appear after 0.03 dpa irradiation, both in iodine and argon gas environments. The iodine presence resulted in ductility reduction, compared with the argon result in all irradiation dose ranges examined. The stress applied during irradiation caused ductility loss to commence at lower dosage than in the case of stress-free irradiation. These results are discussed in relation to the existing stress corrosion cracking models

  7. In Vivo Measurements of Caesium-137 with a Human Body Counter; Dosage du Cesium 137 In Vivo au Moyen d'un Anthropogammametre; 0418 0417 041c 0414 ; Medidas de Cesio-137 In Vivo con un Antropogammametro

    Energy Technology Data Exchange (ETDEWEB)

    Melandri, C. [C.N.E.N., Divisione di Biologia e Protezione Sanitaria, Bologna (Italy); Rimondi, O. [Istituto di Fisica Dell' Universita di Bologna (Italy)

    1964-11-15

    Data are given of the first measurements of human body radioactivity made in Italy with the whole-body counter built by the authors in 1962. The counter shielding is made of iron bricks 16 cm thick. Measurements inside are 270 cm x 200 cm x 200 cm. The detectors are a 9 in x 4 in Nal (Tl) crystal and three 30 cm x 20 cm x 17 cm plastic scintillators which can be used separately, in parallel, or in anticoincidence and in coincidence with the crystal. The counter has been used for determining the behaviour of the caesium-137 body content from fall-out in the whole population and for estimating the caesium-137 retention in a subject accidentally contaminated. The first measurements of caesium-137 body content from fall-out were taken on 13 subjects in December 1962 and continued regularly at three-month intervals. A gradual increase was noted in the average caesium-137 content of the population throughout the whole of 1963 up to the value of 124 pc/g of potassium in September. In a case of accidental contamination from caesium-137 (about two years before the counter was built) measurements were made in order to determine both the effective half-life of the long-term component and the initial intake of the radioelement. (author) [French] Le memoire donne des renseignements sur les premiers dosages de l'activite du corps humain qui ont ete faits en Italie au moyen de l'anthropogammametre construit par les auteurs en 1962. La protection est en briques de fer de 16cmd'eDaisseur: ses dimensions interieures sont de 270 x 200 x 200 cm. L'appareil comporte comme detecteurs un cristal de Nal (Tl) de 22, 5 x 10 cm et trois scintillateurs en matiere plastique de 30 x 20 x 17 cm qui peuvent etre utilises separement, en parallele, ou en anticoincidence et en coincidence avec le cristal. Les auteurs ont employe l'appareil pour determiner comment varie la charge corporelle de cesium 137 due aux retombees dans l'ensemble de la population, et pour estimer la retention de cesium 137

  8. Recovery and recrystallisation of zircaloy-4

    International Nuclear Information System (INIS)

    Derep, J.L.; Rouby, D.; Fantozzi, G.

    1981-01-01

    Examination of the three mechanisms that control the recovery of zircaloy-4 workhardened by rolling: polygonisation leading to a cellular structure, annihilation of dislocations of opposite sign producing thinning of the cell walls, and growth of subgrains by coalescence [fr

  9. Chemical and microstructural characterization of recycled zircaloy

    International Nuclear Information System (INIS)

    Martinez, Luis G.; Pereira, Luiz A.T.; Rossi, Jesualdo L.; Takiishi, Hidetoshi; Sato, Ivone M.; Scapin, Marcos A.; Orlando, Marcos T.D.

    2011-01-01

    PWR reactors employ as nuclear fuel UO 2 pellets with Zircaloy clad. Brazil is autonomous in the nuclear fuel cycle, from uranium mining to enrichment and nuclear fuel manufacture. However, the industrial production of nuclear zirconium alloys does not meet the demand, leading to importation of Zircaloy for fuel manufacturing. In the fabrication of fuel elements parts, machining chips of alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is strategic in economical and environmental aspects. In this work are described two methods that are being developed to recycle Zircaloy chips. The first method the Zircaloy machining chips are melted using an electric arc furnace to obtain small laboratory ingots. The second method uses powder metallurgy technique. By this later method, the Zircaloy chips are submitted to a hydriding process and the resulting material is milled in a high-energy ball mill. The powder is cold isostatically pressed and vacuum sintered. The elemental composition of the materials obtained using both methods is being determined using X-ray fluorescence techniques and compared to the specifications of nuclear grade Zircaloy and to the composition of the starting chips. The phase composition of the laboratory ingots was determined using X-ray diffraction. The ingots were vacuum annealed and the microstructures resulting from both processing methods before and after heat treatments were characterized using optical and scanning electron microscopy. The hardness of the materials was evaluated. A methodology of chemical analysis using X-ray fluorescence spectrometry, for composition certification, was established and tested. The results showed that recycled Zircaloy presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding cap-ends, using near net shape sintering. (author)

  10. Stress corrosion cracking of Zircaloys. Final report

    International Nuclear Information System (INIS)

    Cubicciotti, D.; Jones, R.L.; Syrett, B.C.

    1980-03-01

    The overall aim has been to develop an improved understanding of the stress corrosion cracking (SCC) mechanism considered to be responsible for pellet-cladding interaction (PCI) failures of nuclear fuel rods. The objective of the present phase of the project was to investigate the potential for improving the resistance of Zircaloy to iodine-induced SCC by modifying the manufacturing techniques used in the commercial production of fuel cladding. Several aspects of iodine SCC behavior of potential relevance to cladding performance were experimentally investigated. It was found that the SCC susceptibility of Zircaloy tubing is sensitive to crystallographic texture, surface condition, and residual stress distribution and that current specifications for Zircaloy tubing provide no assurance of an optimum resistance to SCC. Additional evidence was found that iodine-induced cracks initiate at local chemical inhomogeneities in the Zircaloy surface, but laser melting to produce a homogenized surface layer did not improve the SCC resistance. Several results were obtained that should be considered in models of PCI failure. The ratio of axial to hoop stress and the temperature were both shown to affect the SCC resistance whereas the difference in composition between Zircaloy-2 and Zircaloy-4 had no detectable effect. Damage accumulation during iodine SCC was found to be nonlinear: generally, a given life fraction at low stress was more damaging than the same life fraction at higher stress. Studies of the thermochemistry of the zirconium-iodine system (performed under US Department of Energy sponsorship) revealed many errors in the literature and provided important new insights into the mechanism of iodine SCC of Zircaloys

  11. Zircaloy behaviour in high temperature irradiated water

    International Nuclear Information System (INIS)

    Urbanic, V.F.

    1982-04-01

    The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

  12. Characterization of electron beam welded Zircaloy-4

    International Nuclear Information System (INIS)

    Anishetty, Sharath; Manna, I.; Majumdar, J. Dutta

    2015-01-01

    Zirconium (Zr) alloys are the backbone materials for thermal reactors because of their low neutron absorption cross section and in addition have suitable properties like high temperature mechanical and corrosion properties. For various structural applications, different Zirconium based alloys are used. Zircaloy-4 (Zr-4) is most commonly used as channel boxes in boiling water reactors (BWRs), intermediate grid applications in pressurized water reactors (PWRs) and in fuel cladding. Zircaloy cladding acts as a barrier between the radioactive fuel and exterior coolants. Therefore, the structural integrity of the cladding tube is extremely important in the safe operation of reactors. Efforts are being made to produce Zircaloy-4 products with better mechanical properties. Different routes of processing are involved like forging, pilgering and extrusion are developed over years in fabricating components to improve in-reactor performance. In this study, microstructure and hardness properties of electron beam welded Zr-4 was evaluated

  13. Reaction of tellurium with Zircaloy-4

    International Nuclear Information System (INIS)

    Boer, R. de; Cordfunke, E.H.P.

    1994-09-01

    Interaction of tellurium vapour with Zircaloy during the initial stage of an accident will lead to retention of tellurium in the core. For reliable estimation of the release behaviour of tellurium, it is necessary to know which zirconium tellurides are formed during this interaction. In this work the reaction of tellurium with Zircaloy-4 has been studied, using various reaction temperatures and tellurium vapour pressures. The compound ZrTe 2-x is formed on the surface of the Zircaloy in a broad range of reaction temperatures and vapour pressures. It is found that the formation of the more zirconium-rich compound Zr 5 Te 4 is favoured at high reaction temperatures is combination with low tellurium vapour pressures. (orig.)

  14. Hydrogen terminal solubility in Zircaloy-4

    International Nuclear Information System (INIS)

    Vizcaino, Pablo; Banchik, Abrahan D.

    1999-01-01

    Terminal solubility temperature of hydrogen in zirconium and its alloys is an important parameter because hydrides precipitation embrittled these materials making them susceptible to the phenomenon known as retarded hydrogen cracking. This work continues the study presented in the 25 AATN Meeting. Within this framework, a study focused on determining these curves in recrystallized Zircaloy-4, using scanning differential calorimetric technique. Terminal solubility curves for Zircaloy-4 were constructed within a concentration range from 40 to 640 ppm in hydrogen weight and comparisons with results obtained by other authors were made. (author)

  15. Residual stresses in zircaloy welds

    International Nuclear Information System (INIS)

    Santisteban, J. R.; Fernandez, L; Vizcaino, P.; Banchik, A.D.; Samper, R; Martinez, R. L; Almer, J; Motta, A.T.; Colas, K.B; Kerr, M.; Daymond, M.R

    2009-01-01

    Welds in Zirconium-based alloys are susceptible to hydrogen embrittlement, as H enters the material due to dissociation of water. The yield strain for hydride cracking has a complex dependence on H concentration, stress state and texture. The large thermal gradients produced by the applied heat; drastically changes the texture of the material in the heat affected zone, enhancing the susceptibility to delayed hydride cracking. Normally hydrides tend to form as platelets that are parallel to the normal direction, but when welding plates, hydride platelets may form on cooling with their planes parallel to the weld and through the thickness of the plates. If, in addition to this there are significant tensile stresses, the susceptibility of the heat affected zone to delayed hydride cracking will be increased. Here we have measured the macroscopic and microscopic residual stressed that appear after PLASMA welding of two 6mm thick Zircaloy-4 plates. The measurements were based on neutron and synchrotron diffraction experiments performed at the Isis Facility, UK, and at Advanced Photon Source, USA, respectively. The experiments allowed assessing the effect of a post-weld heat treatment consisting of a steady increase in temperature from room temperature to 450oC over a period of 4.5 hours; followed by cooling with an equivalent cooling rate. Peak tensile stresses of (175± 10) MPa along the longitudinal direction were found in the as-welded specimen, which were moderately reduced to (150±10) MPa after the heat-treatment. The parent material showed intergranular stresses of (56±4) MPa, which disappeared on entering the heat-affected zone. In-situ experiments during themal cyclong of the material showed that these intergranular stresses result from the anisotropy of the thermal expansion coefficient of the hexagonal crystal lattice. [es

  16. Iodine stress corrosion cracking in Zircaloy

    International Nuclear Information System (INIS)

    Andrade, A.H.P. de; Pelloux, R.M.N.

    1983-01-01

    The subcritical growth of iodine-induced cracks in unirradiated Zircaloy plates is investigated as a function of the stress intensity factor K. The testing variables are: crystallographic texture (f-Number), microstructure (grain directionaly), heat treatment (stress relieved vs recrystallized plate), and temperature. The iodine partial pressure is 40Pa. (author) [pt

  17. Measurement of Caesium-137 in the Normal Person; Dosage du Cesium 137 chez un Sujet Normal; 0418 0417 041c 0414 ; Determinacion Cuantitativa del Cesio-137 en el Individuo Normal

    Energy Technology Data Exchange (ETDEWEB)

    Huycke, E. J.; Oberhausen, E. [United States Army Medical Research Unit, Europe, Landstuhl/Pfalz, Federal Republic of Germany (Germany)

    1964-11-15

    depuis juin 1962 met en evidence pour le cesium 137 une periode biologique moyenne de 140 j chez les sujets ages de plus de 22 a. En comparant la charge corpo- rellemoyenne de cesium 137 chez ces sujets a celle qui est presente chez de sujets ages de 8 a 17 a, on constate que la periode biologique du cesium 137 est plus courte chez les sujets plus jeunes. Cette observation concorde avec les donnees obtenues precedemment au moyen du meme appareil entre juin 1959 et decembre 1960; ces donnees indiquaient aussi que la charge corporelle de cesium 137 etait plus faible chez l'enfant que chez l'adulte. Une autre analyse des donnees obtenues recemment porte sur l'ecart type (sigma) de la charge corporelle en cesium 137 pour une population nombreuse. Dans cette population, l'ecart type des valeurs mesurees de la quantite de cesium par gramme de potassium de l'organisme est de 33%, soit plus de deux fois l'ecart type de la charge corporelle de potassium pour la meme population. En comparant les resultats obtenus pour un groupe de sujets qui ont ete soumis a plusieurs dosages au cours d'une periode prolongee aux-moyennes mensuelles pour l'ensemble de la population sur laquelle les mesures ont ete effectuees, les auteurs ont constate que l'importance de l'ecart type de la charge corporelle de cesium 137 ne peut pas etre uniquement due a la diversite des regimes alimentaires de la population sur laquelle les mesures ont ete effectuees. (author) [Spanish] Durante el perfodo de jumo de 1959 a octubre de 1963 el antropogammametro de Landstuhl permitio medir el contenido de cesio-137 de mas de 15 000 personas normales. La media mensual de estas medidas indica un aumento constante de contenido en cesio-137 desde junio de I962, aumento que se ha acelerado durante el periodo de junio a octubre de 1963. Una evaluacion subsiguiente de las medidas mensuales desde junio de 1962 indica para el cesio-137 un perfodo biologico de 140 d en el caso jde personas mayores de 22 anos. De la comparacion

  18. Dosage of boron traces in graphite, uranium and beryllium oxide; Dosage de traces de bore dans le graphite, l'uranium et l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Coursier, J [Ecole Nationale Superieure de Physique et Chimie Industrielles, 75 - Paris (France); Hure, J; Platzer, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [French] Le probleme du dosage du bore dans les materiaux servant a la construction de reacteurs nucleaires se pose de la facon suivante: determiner a environ 0,1 ppm pres des quantites de bore de l'ordre de quelques dixiemes de ppm. Nous avons choisit la colorimetrie a la curcumine comme methode de dosage. Pour atteindre les teneurs indiquees, il est necessaire d'effectuer une separation prealable du bore et des materiaux de base, soit par extraction du fluoborate de tetraphenylarsonium dans le cas du dosage de bore dans l'uranium et l'oxyde de beryllium, soit par l'utilisation d'une resine echangeuse de cations dans le cas du graphite. (M.B.)

  19. Characterization of Zircaloy-2 and Zircaloy-4 by X-Ray fluorescence

    International Nuclear Information System (INIS)

    Sato, I.M.; Imakuma, K.; Salvador, V.L.R.

    1981-03-01

    The analytical characterization of zircaloy-2 and zircaloy-4 is intimataly connected with the determination of Sn, Fe, Cr, Ni, O, N, H, and Hf. An analytical method developed in this laboratory is discribed for the determination of metallic elements like Sn, Fe, Cr and Ni using the technique of X-ray fluorescence. The samples are prepared in the form of double-layer pellets using boric acid as a binding agent. The zircaloy-4 is dissolved in hydrofluoric acid and the metallic elements are converted to fluorides. The standard samples used for calibration are prepared from synthetic materials. The elements are determined by measuring the characteristic first order K α lines. A Zircaloy-4 sample analysed yielded the following values: Sn=1.30+-0.03%, Fe=0.18+-0.01%, Cr=0.088+-0.004% and Ni=14+-3 ppm. The reproducibility, precision, as well as the theoretical limit of detection of the method are discussed. The determination of the elements O, N and H present as occluded gas in the zircaloy is nearing completion. These analyses are being carried out by a Mass Spectrometric technique where an aliquot of the released gas is analysed. (Author) [pt

  20. Device for sampling sodium from a circuit with a view to dosing its impurities (1963); Dispositif de prelevement de sodium sur un cuicuit en vue du dosage des impuretes (1063)

    Energy Technology Data Exchange (ETDEWEB)

    Sannier, J; Vingot, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    The device described was developed with the two following essential conditions in mind: - absence of any pollution during the sampling operation and the transfer to the analysis apparatus; - simultaneous extraction of several samples, each one remaining representative of the sodium in the circuit. The sampling is therefore carried out completely in a vacuum in a Pyrex-glass apparatus, this limiting the sodium temperature to a maximum of 240 deg. C. The samples are in the form of bulbs sealed in vacuo containing 2 to 3 grams of sodium each. The analysis of the oxygen by the amalgamation technique, carried out on samples obtained by this method show a very satisfactory reproducibility for concentrations of under 20 ppm. (authors) [French] Le dispositif decrit a ete mis au point compte tenu des deux imperatifs essentiels suivants: - absence de toute pollution pendant les operations de prelevement et de transport vers l'appareil d'analyse; - obtention simultanee de plusieurs echantillons, chacun restant representatif du sodium du circuit. A cet effet, le prelevement est effectue entierement sous vide dans un appareillage en verre Pyrex, ce qui limite la temperature du sodium a 240 deg. C. Les echantillons se presentent sous la forme d'ampoules scellees sous vide et contenant chacune 2 a 3 grammes de sodium. L'analyse de l'oxygene par la technique d'amalgamation, effectuee sur des echantillons preleves par cette methode, revele une reproductibilite tres satisfaisante, dans le domaine des teneurs inferieures a 20 ppm. (auteurs)

  1. Tensile creep of beta phase zircaloy-2

    International Nuclear Information System (INIS)

    Burton, B.; Reynolds, G.L.; Barnes, J.P.

    1977-08-01

    The tensile creep and creep rupture properties of beta-phase zircaloy-2 are studied under vacuum in the temperature and stress range 1300-1550 K and 0.5-2 MN/m 2 . The new results are compared with previously reported uniaxial and biaxial data. A small but systematic difference is noted between the uniaxial and biaxial creep data and reasons for this discrepancy are discussed. (author)

  2. Chemical analysis of radioactive liquid wastes. Volumetric dosage of sulfates in the presence of phosphates, 'nitrochromazo' being used as an indicator; Analyse chimique des effluents radioactifs - dosage volumetrique des sulfates en presence de phosphates a l'aide du 'nitrochromazo' comme indicateur

    Energy Technology Data Exchange (ETDEWEB)

    Testemale, G; Girault, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    A simple titration technique of SO{sub 4} ions in the presence of PO{sub 4} ions has been perfected. The pH of the medium is stated to 1,7-2,0 the colour of the indicator changes from violet to blue. The method is quick, accurate and can be fitted to biological studies or in the industry of fertilizes. The synthesis method of nitrochromazo (acid 2, 7 bis (4 nitro 2 sulfobenzene 1 azo) 1-8 hydroxynaphthalene 3-6 disulfonic) is described. (authors) [French] Une technique simple de titrage des ions SO{sub 4} en presence d'ions PO{sub 4} a ete mise au point. Le pH du milieu est fixe a 1,7-2,0, le virage de l'indicateur s'effectue du violet au bleu. La methode est rapide, precise et peut etre adaptee a des travaux de biologie ou dans l'industrie des engrais. La methode de synthese du 'Nitrochromazo' (acide 2, 7 bis (4 nitro 2 sulfobenzene 1 azo) 1-8 dihydroxynapthtalene 3-6 disulfonique) est decrite. (auteurs)

  3. Hydrogen isotope storage in zircaloy scrap

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. S.; Kuk, I. H.; Chung, H.; Paek, S. W.; Kang, H. S

    1999-08-01

    8 MCi of tritium a year will be produced after wolsong TRF is in operation. The metal hydride form is one of useful tritium storage. The metals in use for metal hydride are uranium, titanium, etc., however uranium is limited to use by regulation, and titanium is relatively costly. Both metals are not produced in country but whole amount is imported. On the other hand 2,000kg of zircaloy scrap is produced by CANDU nuclear fuel fabrication process, which is also useful for hydrogen storage. The purpose of this study is to evaluation of hydrogen absorption capacity for zircaloy scrap that is produced as waste by CANDU nuclear fuel fabrication process. The sample evacuated for an hour at 1000 deg C. The strip showed higher capacity : 0.7 at 25 deg C, 2.0 at 200 deg C, 2.0 at 200 deg C, 2.0 at 400 deg C, respectively. The H/M values for commercial zircaloy sponge were 2.0 at 25 deg C and 2.0 at 400 deg C.

  4. Hydrogen isotope storage in zircaloy scrap

    International Nuclear Information System (INIS)

    Lee, H. S.; Kuk, I. H.; Chung, H.; Paek, S. W.; Kang, H. S.

    1999-08-01

    8 MCi of tritium a year will be produced after wolsong TRF is in operation. The metal hydride form is one of useful tritium storage. The metals in use for metal hydride are uranium, titanium, etc., however uranium is limited to use by regulation, and titanium is relatively costly. Both metals are not produced in country but whole amount is imported. On the other hand 2,000kg of zircaloy scrap is produced by CANDU nuclear fuel fabrication process, which is also useful for hydrogen storage. The purpose of this study is to evaluation of hydrogen absorption capacity for zircaloy scrap that is produced as waste by CANDU nuclear fuel fabrication process. The sample evacuated for an hour at 1000 deg C. The strip showed higher capacity : 0.7 at 25 deg C, 2.0 at 200 deg C, 2.0 at 200 deg C, 2.0 at 400 deg C, respectively. The H/M values for commercial zircaloy sponge were 2.0 at 25 deg C and 2.0 at 400 deg C

  5. Hydriding and neutron irradiation in zircaloy-4

    International Nuclear Information System (INIS)

    Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

    2003-01-01

    The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

  6. Thermal creep of Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Murty, K.L.; Clevinger, G.S.; Papazoglou, T.P.

    1977-01-01

    Data on the hoop creep characteristics of Zircaloy tubing were collected at temperatures between 600 F and 800 F, and at stress levels ranging from 10 ksi to 25 ksi using internal pressurization tests. At low driving forces, exposures as long as 2000 hours were found insufficient to establish steady state creep. The experimental data at temperatures of 650 F to 800 F correlate well with an exponential stress dependence, and the activation energy for creep was found to be in excellent agreement with that for self-diffusion. The range of stresses and temperatures is too small to study the overall effect of these variables on the activation energy for creep. The experimental steady state creep-rates and those predicted from the creep equation used agree within a factor of 1.3. These correlations imply that the mechanism for hoop creep of Zircaloy-4 cladding is characterized by an activation energy of approximately 60 kcal/mole and an activation area of about 20b 3 . In addition, the exponential stress dependence implies that the activation area for creep is stress-independent. These results suggest that the climb of edge dislocations is the rate controlling mechanism for creep of Zircaloy-4. The transient creep regime was also analysed on the premise that primary creep is directly related to the rate of dispersal of dislocation entanglements by climb. (Auth.)

  7. Process for reliewing stresses in a zircaloy 2 or zircaloy 4 strip

    International Nuclear Information System (INIS)

    Charquet, D.; Dombre, M.

    1986-01-01

    Fabrication process of a zircaloy 2 or zircaloy 4 strip with an oxygen content between 900 and 1600 ppm with the following mechanical properties: E0.2≥250MPa at 315 deg C, parallel and perpendicular A% ≥4 at 20 deg C. The strip is rolled and stabilized by heat treatment between 490 and 580 deg C for 1 to 10 minutes and partially recrystallized for 0.5 to 5 vol.%. It is used for spacers of nuclear fuels [fr

  8. Intercavitary implants dosage calculation

    International Nuclear Information System (INIS)

    Rehder, B.P.

    The use of spacial geometry peculiar to each treatment for the attainment of intercavitary and intersticial implants dosage calculation is presented. The study is made in patients with intercavitary implants by applying a modified Manchester technique [pt

  9. Creep damage in zircaloy-4 at LWR temperatures

    International Nuclear Information System (INIS)

    Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

    1978-08-01

    The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

  10. Study on kinetic of strain-aging in zircaloy-4

    International Nuclear Information System (INIS)

    Gomes, P.A.

    1977-01-01

    The strain-aging in zircaloy-4 has been investigated in this work and a study of the general problems involving this phenomenon has been realized in Zirconium and its alloys. It has been verified that a yield point appears in the Zircaloy-4, when it is submitted to strain-aging treatment between the temperatures 200 0 C and 400 0 C. (author)

  11. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  12. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    International Nuclear Information System (INIS)

    Rudisill, T; John Mickalonis, J

    2006-01-01

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

  13. Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Santanam, L.; Raghavan, S.; Chin, B.A.

    1990-12-01

    Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340 degree C (613 K) for typically stressed rods (70--100 MPa) and 300 degree C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs

  14. Electromigration of hydrogen in zircaloy-2

    International Nuclear Information System (INIS)

    Parmeswaran, P.; Kamachi Mudali, U.; Raghunathan, V.S.; Govinda Rajan, K.

    1989-01-01

    Electromigration is a purification technique for removing interstitial impurities from metals like Zr, Ti and Nb. It uses an electric field to induce migration of atoms from one end to other. This paper describes an attempt to purify zircaloy-2 of its hydrogen content by this technique. Resistivity measurement has been used to evaluate the change in impurity concentration that occurs during the process. Results indicate the movement of hydrogen atoms towards the cathode end. The value of the effective charge number, Z * , calculated from the results confirms hydrogen migration to the cathode aided by a positive wind force. (author). 6 refs., 5 figs

  15. Study of the Zircaloy-2 welding

    International Nuclear Information System (INIS)

    Rodriguez-Solano, R.; Jimenez Moreno, J. M.

    1968-01-01

    After a bibliographical analysis of the Zircaloy-2 properties its welding was approached. The selected procedure is the TIG (Tungsten Inert Gas) d.c. arc-welding under an inert atmosphere vessel. A detailed description of the equipment and characteristics is given. During the tests two types of argon were used: one with 96 ppm. Impurities, the other with 7 ppm- impurities. It is al so mentioned the welding in helium atmosphere. The contamination of the welding was evaluated through hardness testing. (Author) 3 refs

  16. The anisotropic mechanical behaviour of zircaloy-2

    International Nuclear Information System (INIS)

    Ballinger, R.; Pelloux, R.M.

    1980-01-01

    Zirconium alloys used in the LWR industry crystallize in the hexagonal crystal structure below approximately 1136 K and many of the fabrication steps are performed below this temperature. The hexagonal structure possesses a limited number of slip systems and normal deformation processes result in extensive twinning. The twinning process results in the development of a fabrication texture, the type and extent of which is a function of the strain path used in the fabrication process. The texture which develops is important for two reasons. First, the texture at a given point in the fabrication process will determine the ease with which the next strain increment may be taken. Second, the texture of the completed part will have a significant effect on its in service performance because properties such as yield strength, creep strength, and fatigue and stress corrosion cracking resistance are a strong function of texture. Currently there is little data available concerning the evolution of textures as a function of strain path during the fabrication process of Zircaloy. Consequently this experimental investigation was conducted to determine the effect of textures on the mechanical behaviour of Zircaloy-2 with a primary emphasis on the evolution of texture during plastic deformation. (author)

  17. Embrittlement by hydrogen in zircaloy-4

    International Nuclear Information System (INIS)

    Almendariz M, M.C.

    1981-01-01

    The brittleness study of zircaloy-4 (nuclear quality) by hydrogen in the lattice was carried out with the purpose to watch the alterations at mechanic properties and fracture appearance for different thermal treatments. We used a statistical experimental method to watch both alterations. Fracture toughness property was evaluated in a semiquantitative way, and this property was calculated by integral J method but at a modified version, this modification lies in the area calculation under the curve of load versus head displacement plot; we used Instron machine to evaluate it. Three points bending proof was carried out in accordance with the device that specify A.S.T.M. standards. The samples were treated with hydrogen by means of catodic charged method and subsequently mechanic proof was realized. We used statistical analysis to get information of experimental results, and the watched general behaviour was a great disminution of the fracture toughness (in relation to not treated hydrogen sample), always that the hydrogen is present in the lattice, likewise we did watch that hydrogen does not influence at fracture appearance change, further there is a threshold hydrogen concentration at wich it starts to brittle and prior not influence it. We did conclude of results analysis that the fracture toughness is reduced by hydrogen and threshold concentration is subject to thermal treatment. Experimental results can be considered as semiquantitatives, but they gave us an explicit idea of hydrogen effect in zircaloy-4. (author)

  18. Biaxial mechanical tests in zircaloy-4

    International Nuclear Information System (INIS)

    Mintzer, S.R.; Bordoni, R.A.A.; Falcone, J.M.

    1980-01-01

    The texture of the zircaloy-4 tubes used as cladding in nuclear fuel elements determines anisotropy of the mechanical properties. As a consequence, the uniaxial tests to determine the mechanical behaviour of the tubes are incomplete. Furthermore, the cladding in use is subject to creep with a state of biaxial tensions. For this reason it is also important to determine the biaxial mechanical properties. The creep tests were performed by internal pressure for a state of axial to circumferential tensions of 0.5. Among the experimental procedures are described: preparation of the test specimens, pressurizing equipment, and the implementation of a device that permits a permanent register of the deformation. For the non-irradiated Atucha type zircaloy-4 sheaths, experimental curves of circumferential deformation versus time were obtained, in tests at constant pressure and for different values of temperature and pressure. An empirical function was determined to adjust the experimental values for the speed of the circumferential deformation in terms of the initial tension applied, temperature and deformation, and the change of the corresponding parameters in accordance to the range of the tensions. Also the activation energy for creep was determined. (M.E.L.) [es

  19. A dosing method in the same time of the radiocesium and the radiostrontium in natural waters; Dosage simultane du cesium 137 et des strontium 89 et 90 dans les eaux naturelles

    Energy Technology Data Exchange (ETDEWEB)

    Scheidhauer, J; Messainguiral, L [Commissariat a l' Energie Atomique, Marcoule (France).Centre d' Etudes Nucleaires

    1960-07-01

    Prior a concentration and an unselective elution are effected by the way of an exchange resin. The cesium is absorbed on ammonium molybdophosphate precipitated with hydrofluoric acid after radium elimination. The strontium is stripped in concentrated nitric acid, precipitated in the form of strontium carbonate and then counted near the radioactive balance. At last the yttrium is separated by the thenoyltrifluoroacetone and counted on a counting device. From the outset of this paper, an analysis of the method is explained. (author) [French] On opere d'abord une concentration sur resine et une elution non selective. Le cesium est absorbe sur un precipite de phosphomolybdate d'ammonium en presence d'acide fluorhydrique, apres separation du radium. Le strontium est traite a l'acide nitrique concentre, precipite sous forme de carbonate et compte au voisinage de l'equilibre radioactif ({sup 90}Sr {sup 9O}y ). L'yttrium 90 est ensuite separe a la thenoyltrifluoracetone et compte. Une etude de la methode est presentee au debut de ce memoire. (auteur)

  20. Development of zircaloy deformation model to describe the zircaloy-4 cladding tube during accidents

    International Nuclear Information System (INIS)

    Raff, S.

    1978-01-01

    The development of a high-temperature deformation model for Zircaloy-4 cans is primarily based on numerous well-parametrized tensile tests to get the material behaviour including statistical variance. It is shown that plastic deformation may be described by a power creep law, the coefficients of which show strong dependence on temperature in the relevant temperature region. These coefficients have been determined. A model based on these coefficients has been established which, apart from best estimate deformation, gives upper and lower bounds of possible deformation. The model derived from isothermal uniaxial tests is being verified against isothermal and transient tube burst tests. The influence of preoxidation and increased oxygen concentration during deformation is modeled on the basis of the pseudobinary Zircaloy-oxygen phase diagram. (author)

  1. The testing of a method for dosing plutonium by {alpha}-counting in the presence of strong concentrations of salts or of uranium; Essai d'une methode de dosage du plutonium par comptage {alpha} en presence de fortes concentrations en sels ou en uranium

    Energy Technology Data Exchange (ETDEWEB)

    Fontaine, A M; Baude-Malafosse, L M; Cunq, M J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    This report describes a method for dosing small quantities of plutonium in a solution having a high concentration of salts. It shows the possibility of dosing up to 5.10{sup -3} {mu}g of Pu in the presence of 10 mg of NaNO{sub 3} with out decreasing the counting-rate. The only error possible is that in the counting. It is also possible to dose 10{sup -3} {mu}g of Pu in the presence of 1,7 mg of uranyl nitrate. (author) [French] Ce rapport decrit une methode de dosage de faibles quantites de plutonium dans une solution de forte concentration en sels. Il montre la possibilite de doser jusqu'a 5.10{sup -3} {mu}g de Pu en presence de 10 mg de NO{sub 3}Na sans diminution du taux de comptage. La seule erreur que l'on puisse faire est l'erreur de comptage. On peut aussi doser 10{sup -3} {mu}g de Pu en presence de 1,7 mg de nitrate d'uranyle. (auteur)

  2. The effect of texture, heat treatment and elongation rate on stress corrosion cracking in irradiated zircaloy

    International Nuclear Information System (INIS)

    Pettersson, K.; Stany, W.; Hellstrand, E.

    1979-03-01

    Irradiated zircaloy samples with different textures and heat treatments have been tested concerning stress corrosion. Irradiated samples of Zr-1Nb, pure Zr and beta quenched zircaloy have also been investigated. Stress-relieve annealled zircaloy is even after irradiation more sensitive to stress corrosion than recrystallized zircaloy. Zr-1Nb and beta quenched zircaloy are much more sinsitive to stress corrosion than the samples with different textures. As a rule irradiated zircaloy is sensitive to stress corrosion at stresses far below the yield point. The breaking stress decreases with the elongation rate. The extension of cracks is much faster in irradiated zircaloy than in unirradiated zircaloy. There is no simple failure criterium for irradiated zircaloy. However for a certain stress and a certain elongation rate the probability for a failure before this stress is reached with a constant elongation rate can be given. (E.R.)

  3. Instrumented impact properties of zircaloy-oxygen and zircaloy-hydrogen alloys

    Energy Technology Data Exchange (ETDEWEB)

    Garde, A.M.; Kassner, T.F.

    1980-04-01

    Instrumented-impact tests were performed on subsize Charpy speciments of Zircaloy-2 and -4 with up to approx. 1.3 wt % oxygen and approx. 2500 wt ppM hydrogen at temperatures between 373 and 823/sup 0/K. Self-consistent criteria for the ductile-to-brittle transition, based upon a total absorbed energy of approx. 1.3 x 10/sup 4/ J/m/sup 2/, a dynamic fracture toughness of approx. 10 MPa.m/sup 1/2/, and a ductility index of approx. 0, were established relative to the temperature and oxygen concentration of the transformed BETA-phase material. The effect of hydrogen concentration and hydride morphology, produced by cooling Zircaloy-2 specimens through the temperature range of the BETA ..-->.. ..cap alpha..' = hydride phase transformation at approx. 0.3 and 3 K/s, on the impact properties was determined at temperatures between 373 and 673 K. On an atom fraction basis, oxygen has a greater effect than hydrogen on the impact properties of Zircaloy at temperatures between approx. 400 and 600 K. 34 figures.

  4. Hydrogen pickup and redistribution in alpha-annealed Zircaloy-4

    International Nuclear Information System (INIS)

    Kammenzind, B.F.; Franklin, D.G.; Duffin, W.J.; Peters, H.R.

    1996-01-01

    Zircaloy-4, which is widely used as a core structural material in Pressurized-Water Reactors (PWR), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and hydrides precipitate after the Zircaloy-4 matrix becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4. To study hydrogen pickup and concentration, a postirradiation nondestructive radiographic technique for measuring hydrogen concentration was developed and qualified. Experiments on hydrogen pickup were conducted in the Advanced Test Reactor (ATR). Ex-reactor tests were conducted to determine the conditions for which hydrogen would dissolve, migrate, and precipitate. Finally, a phenomenological model for hydrogen diffusion was indexed to the data. This presentation describes the equipment and the model, presents the results of experiments, and compares the model predictions to experimental results

  5. Fluorimetric determination of uranium in zirconium and zircaloy alloys

    International Nuclear Information System (INIS)

    Acosta L, E.

    1991-05-01

    The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

  6. Quenching behaviour of hot zircaloy tube

    International Nuclear Information System (INIS)

    Chinchole, A.S.; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.

    2015-01-01

    The quenching process plays a very important role in case of safety of nuclear reactors. During large break Loss of Coolant Accident in a nuclear reactor, the cooling water from the system is lost. Under this condition, cold water is injected from emergency core cooling system. Quenching behaviour of such heated rod bundle is really complex. It is well known that nanofluids have better heat removal capability and high heat transfer coefficient owing to enhanced thermal properties. Alumina nano-particles result in better cooling abilities compared with the traditionally used quenching media. In this paper, the authors have carried out experiments on quenching behaviour of hot zircaloy tube with demineralized water and nanofluids. It was observed that, the tube got quenched within few seconds even with the presence of decay heat and shows slightly reduced quenching time compared with DM water. (author)

  7. Diffusionless bonding of aluminum to Zircaloy-2

    International Nuclear Information System (INIS)

    Watson, R.D.

    1965-04-01

    Aluminum can be bonded to zirconium without difficulty even when a thin layer of oxide is present on the surface of the zirconium . No detectable diffusion takes place during the bonding process. The bond layer can be stretched as much. as 8% without affecting the bond. The bond can be heated for 1000 hours at 260 o C (500 o F), and can be water quenched from 260 o C (500 o F) without any noticeable change in the bond strength. An extrusion technique has been devised for making transition sections of aluminum bonded to zirconium which can then be used to join these metals by conventional welding. Welding can be done close to the bond zone without seriously affecting the integrity of the bond. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 26, 1965. (author)

  8. High-pressure hydriding of Zircaloy

    International Nuclear Information System (INIS)

    Kim, Y.S.

    1996-01-01

    The hydriding characteristics of Zircaloy-2(Zry), sponge zirconium (as a liner on Zry plate), and crystal-bar zirconium exposed to pure H 2 at 0.1 MPa or 7 MPa and 400 C were determined in a thermogravimetric apparatus. The morphology of the hydrided specimens was also examined by optical microscopy. For all specimen types, the rate of hydriding in 7 MPa H 2 was two orders of magnitude greater than in 0.1 MPa H 2 . For Zry, uniform bulk hydriding was revealed by hydride precipitates at room temperature and on one occasion, a sunburst hydride. In addition, all specimen types exhibited a hydride surface layer. In a duplex Zry/sponge-Zr specimen, Zry is more heavily hydrided than the sponge Zr layer. (orig.)

  9. Comparison between zircaloy oxidation in steam and air surroundings

    International Nuclear Information System (INIS)

    Shawkat, M.E.; Hasaneln, H.; Ali, M.; Parlatan, Y.; Albasha, H.

    2013-01-01

    The available experimental data for Zircaloy oxidation in air were reviewed. The behavior of the oxidation kinetics at different temperature ranges was described. It was shown that maintaining the oxidation kinetics within the oxide pre-breakaway region can prevent elevated sheath temperatures due to the oxidation process during postulated accidents. The available correlations to model the oxidation kinetics for pre-breakaway region were reviewed and assessed. Zircaloy-air oxidation correlation based on Leistikow-Berg data was determined to be the most suitable correlation to model pre-breakaway kinetics and it was compared to Urbanic-Heidrick correlation which is widely used for Zircaloy oxidation in steam environment. The results showed that the energy release due to the Zircaloy-steam oxidation bounds the energy released due to Zircaloy-air oxidation up to a sheath temperature referred as the “crossover temperature”. Below this temperature, the impact of Zircaloy-air oxidation on fuel sheath temperature transient can be predicted conservatively using the Urbanic-Heidrick steam correlation. The crossover temperature was calculated for isothermal sheath heating as well as transient sheath heat-up assuming three linear heating rates of 0.6, 1.0, and 1.3 K/s. (author)

  10. Dosage of trace carbon in sodium (1963); Dosage de traces de carbone dans le sodium (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Sannier, J; Vasseur, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    A wet method for dosing carbon in sodium has been developed. The carbon is oxidised in a vacuum using Van SLYKE'S solution. The carbonic acid formed is measured volumetrically; its purity can be controlled by chromatographic analysis. The results obtained show that this method makes it possible to measure carbon in concentrations of about 10 ppm. (authors) [French] Une methode de dosage par voie humide du carbone dans le sodium a ete mise au point. L'oxydation du carbone par la solution de Van SLYKE est realisee sous vide. Le gaz carbonique forme est dose volumetriquement; sa purete peut etre controlee par analyse chromatographique. Les resultats obtenus montrent que cette methode permet de doser des teneurs en carbone de l'ordre de 10 ppm. (auteurs)

  11. Thermal expansion studies on zircaloy-2

    International Nuclear Information System (INIS)

    Sivabharathy, M.; Senthilkumar, A.; Palanichamy, P.; Ramachandran, K.

    2016-01-01

    Zircaloy-2 and Zr-2.5% Nb alloys are widely used in the pressurized heavy water reactors (PHWR) as the material for the pressure tubes. The pressure tube operates at 573 K, 11 MPa internal pressures and is subjected to neutron flux of the order of 1013 n/cm 2 /s. These conditions lead to degradations in the pressure tube with respect to dimensional changes, deterioration in mechanical properties due to irradiation embrittlement, thereby reducing its flaw tolerance, the growth of existing flaws, which were too small or 'insignificant' at the time of installation. Physical and chemical properties of materials are also very essential in nuclear industry and the relations among them is of interest in the selection of materials when they are used in the design and manufacturing of devices particularly for atomic reactors.Studies on the relations between mechanical and thermal properties are of interest to the steel and metal industries as these would give useful information on the relation between hardness and thermal diffusivity (α) of steel. Jayakumar et al have already carried out the ultrasonic and metallographic investigations to see that all the heat-treated specimens retained essentially the martensite structure. In this present work, thermal expansion measurements on useful reactor material, Zircaloy-2 with different sample. Given a β-quenching treatment by heating to 1223 K and holding for 2 h, followed by water quenching. These specimens were then thermally aged for 1 h in the temperature range 473 to 973 K and air-cooled. For all samples, the thermal expansion was carried out and the results are correlated with ultrasonic measurements, metallographic and photoacoustic studies. (author)

  12. Fluorimetric determination of uranium in zirconium and zircaloy alloys; Determinacion fluorimetrica de uranio en aleaciones de zirconio y zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Acosta L, E [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-05-15

    The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

  13. Study of the Zircaloy-2 welding; Estudio de la soldadura de Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez-Solano, R; Jimenez Moreno, J M

    1968-07-01

    After a bibliographical analysis of the Zircaloy-2 properties its welding was approached. The selected procedure is the TIG (Tungsten Inert Gas) d.c. arc-welding under an inert atmosphere vessel. A detailed description of the equipment and characteristics is given. During the tests two types of argon were used: one with 96 ppm. Impurities, the other with 7 ppm- impurities. It is al so mentioned the welding in helium atmosphere. The contamination of the welding was evaluated through hardness testing. (Author) 3 refs.

  14. A study of stress reorientation of hydrides in zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Yourong, Jiang; Bangxin, Zhou [Nuclear Power Inst. of China, Chengdu, SC (China)

    1994-10-01

    Under the conditions of circumferential tensile stress from 70 to 180 MPa for Zircaloy tubes or the tensile stress from 55 to 180 MPa for Zircaloy-4 plates and temperature cycling between 150 and 400 degree C, the effects of stress and the number of temperature cycling on hydride reorientation in Zircaloy-4 tubes and plates and Zircaloy-2 tubes containing about 220 {mu}g/g hydrogen have been investigated. With the increase of stress and/or the number of temperature cycling, the level of hydride reorientation increases. When hydride reorientation takes place, there is a threshold stress concerned with the number of temperature cycling. Below the threshold stress, hydride reorientation is not obvious. When applied stress is higher than the threshold stress, the level of hydride reorientation increases with the increase of stress and the number of temperature cycling. Hydride reorientation in Zircaloy-4 tubes develops gradually from the outer surface to inner surface. It might be related to the difference of texture between outer surface and inner surface. The threshold stress is affected by both the texture and the value of B. So controlling texture could still restrict hydride reorientation under tensile stress.

  15. Phase transformations in neutron-irradiated Zircaloys

    International Nuclear Information System (INIS)

    Chung, H.M.

    1986-04-01

    Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after ∼3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr 3 O and cubic-ZrO 2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ∼4 x 10 21 ncm -2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs

  16. Irradiation growth of Zircaloy (LWBR) development program

    International Nuclear Information System (INIS)

    Williard, H.J.

    1984-01-01

    Irradiation growth of recrystallized annealed (RXA) Zircaloy is divided into four stages and a model is presented to account for each stage. Stage I is a short time, low-strain transient caused by the accumulation of point defects, small interstitial loops, and vacancy clusters. Stage II is a quasi-steady-state region of relatively low strain rate during which the loops grow and intrinsic dislocations climb. Stage III is a transient during which the strain rate increases due to the production and motion of irradiation-induced dislocation lines. Stage IV is a high-strain-rate, steady-state region during which nonrecoverable strain is caused predominantly by glide of the irradiationinduced dislocations. The proposed model is based on two new mechanisms: (1) direct production of an interstitial dislocation loop accompanied by a vacancy cluster in the primary damage event, and (2) production of dislocations due to the activation of Frank-Read sources by internal stresses caused by interaction of the loops with themselves and with intrinsic (cold work) dislocations. Nonconservative, recoverable strain is due to climb of all dislocations, whereas conservative, nonrecoverable strain is caused by glide of irradiation-induced and intrinsic dislocations under the action of the internal stress. The conservative strain follows a (1-3f) texture dependence

  17. High-temperature oxidation of Zircaloy-2 and Zircaloy-4 in steam

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Heidrick, T.R.

    1978-01-01

    At temperatures above the (α + β)/β transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO 2 ) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO 2 + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050-1850 o C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO 2 + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram. (author)

  18. Endotoxin dosage in sepsis

    Directory of Open Access Journals (Sweden)

    Vincenzo Rondinelli

    2012-03-01

    Full Text Available Introduction. Endotoxin, a component of the cell wall of Gram-negative bacteria is a major contributor to the pathogenesis of septic shock and multiple organ failure (MOF. Its entry into the bloodstream stimulates monocytes/macrophages which once activated produce and release cytokines, nitric oxide and other mediators that induce systemic inflammation, endothelial damage, organ dysfunction, hypotension (shock and MOF.The aim of this study is to evaluate the usefulness of a quantitative test for the dosage of endotoxin to determine the risk of severe Gram-negative sepsis. Materials and methods. In the period January 2009 - June 2011 we performed 897 tests for 765 patients, mostly coming from the emergency room and intensive care, of which 328 (43% women (mean age 53 and 437 (57% male (mean age 49. Fifty-nine patients, no statistically significant difference in sex, were monitored by an average of two determinations of EA.All patients had procalcitonin values significantly altered.The kit used was EAA (Endotoxin Activity Assay Estor Company, Milan, which has three ranges of endotoxin activity (EA: low risk of sepsis if <0.40 units, medium if between 0.40 and 0.59; high if 0.60. Results. 78 out of 765 patients (10% had a low risk, 447 (58% a medium risk and 240 (32% a high risk.The dosage of EA, combined with that of procalcitonin, has allowed a more targeted antibiotic therapy. Six patients in serious clinical conditions were treated by direct hemoperfusion with Toraymyxin, a device comprising a housing containing a fiber polypropylene and polystyrene with surface-bound polymyxin B, an antibiotic that removes bacterial endotoxins from the blood. Conclusions.The test is useful in risk stratification as well as Gram negative sepsis, to set and monitor targeted therapies, also based on the neutralization of endotoxin.

  19. Zircaloy-sheathed element rods fitted with thermo-couples

    International Nuclear Information System (INIS)

    Bernardy de Sigoyer, B.; Jacques, F.; Thome, P.

    1963-01-01

    In order to carry out thermal conductivity measurements on UO 2 in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [fr

  20. Mechanical analysis of surface-coated zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Lee, You Ho; Lee, Jeong Ik; No, Hee Cheon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2017-08-15

    A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

  1. Zircaloy nodular corrosion analysis by an image processing technique

    International Nuclear Information System (INIS)

    Kawashima, Junko; Sato, Kanemitsu; Kuwae, Ryosho; Higashinakagawa, Emiko

    1987-01-01

    An image processor has been fabricated to examine out-of-pile nodular corrosion for Zircaloy-2 tubings. The covering fraction, which is the percentage of the nodule occupying area on the Zircaloy surface, was measured with the processor. The covering fraction showed a strong correlation with the weight gain at any corrosion time of this experiment. The correlation observed can be explained by a model for the lenticular shape of the nodules. The image processor also gives unfolded pictures of the whole Zircaloy surface. By analyzing the picture, the location of the nodules generated was found to be determined in an early stage of corrosion. New nodules were not produced later, and the nodules only grew larger with time. (orig.)

  2. A tem investigation on intermetallic particles in zircaloy-2

    International Nuclear Information System (INIS)

    Sudarminto, Harini Sosiati; Kuwano, Noriyuki; Oki, Kensuke

    1996-01-01

    Tem investigation were conducted on the heat treated zircaloy-2 having the composition of Zr containing 1.6% Sn, 0.2% Fe, 0.1% Cr and 0.05% Ni (%wt) in order tostudy the characteristics of intermetallic particles related to the microstructural basis on the corrosion effect. Forged zircaloy-2 was annealed in the β-phase at 1050 C degrees for various isothermally in the α-phase region at 650 and 750 C degrees, followed by water quenching. The size precipates, the lower became their number. By increasing the annealing temperature, the growth of precipitates formed in this zircaloy-2 were of the Zr(Cr,Fe) 2 and Zr 2 (Fe,Cr,Ni) types. These kinds of precipitates and the ratios of Fe/Cr were independent of size and shape of precipitates and annealing time and temperature. (author), 16 refs, 2 tabs, 5 figs

  3. Fatigue properties of Zircaloy-2 in a PWR water environment

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The continuing trend of operation of light water reactors is towards power cycling as a means of operating the systems more efficiently. Depending upon the reactor design and mode of power cycling this could lead to significant fatigue usage in Zircaloy structural components. In order to design against the possibility of gross yielding or fast fracture of such components as a result of this it is obviously necessary to be able to predict conservatively the fatigue properties of Zircaloy under the reactor operating conditions

  4. The oxidation kinetics of zircaloy - 4 under isothermal conditions

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos; Cardoso, P.E.

    1982-01-01

    The oxidation kinetics of zircaloy-4 tubes was studied by means of isothermal tests in the temperature interval 500 0 C to 900 0 C. Dry oxygen and water steam, were used as oxidant agents. The results show that the oxidation kinetics law exhibits a behaviour from cubic to parabolic in the range of the time and temperatures of the experiment. Dry oxygen shows a stronger oxidation effect than water steam. A special mechanical test to study the embrittlement effect in the small samples of zircaloy tubes was used. (Author) [pt

  5. Influence of foreign matter on the flammability of Zircaloy

    International Nuclear Information System (INIS)

    Praetorius, R.; Muenzel, H.

    1990-01-01

    When cutting Zircaloy cladding in the head end of a reprocessing plant, fine particles with a high chemical reactivity are produced. Spontaneous ignition may cause fire or dust explosion. Therefore their ignition and fire behaviour was studied. As a result it can be stated that sugar or a concentrated sugar solution (syrup) poured over a Zircaloy fire is particularly suited as a fire-extinguishing agent. The developing caramel melt prevents air access and sparking. In addition, the sugar can be washed out easily before cementing, and so additional waste arising can be avoided. (DG) [de

  6. Corrosion performance of new Zircaloy-2-based alloys

    International Nuclear Information System (INIS)

    Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

    1994-01-01

    A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

  7. Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1979-02-01

    A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

  8. Influence of temperature on the Zircaloy-4 plastic anisotropy

    International Nuclear Information System (INIS)

    Limon, R.; Bechade, J.L.; Lehmann, S.; Maury, R.; Soniak, A.

    1995-01-01

    In order to improve the comportment modelling of PWR fuel pin, and more precisely their canning tubes, Framatome and the CEA have undertake an important study program of Zircaloy-4 mechanical properties. It includes in particular the study of the plasticity between 20 and 400 degree Celsius. This material being not isotropic because of the zirconium hexagonal crystal network and the texture presented by the canning tubes, its plastic anisotropy has been measured. The obtained results for the canning in *slack* and recrystallized before irradiation Zircaloy-4 are presented and the deformation systems able to explain the observed anisotropy is researched. (O.L.). 6 refs., 4 figs., 1 tab

  9. Glucinium dosimetry in beryl; Dosage du glucinium dans le beryl

    Energy Technology Data Exchange (ETDEWEB)

    Kremer, M

    1949-05-01

    The application of the method developed by Kolthoff and Sandell (1928) for the dosimetry of glucinium (beryllium) in beryl gives non-reproducible results with up to 20% discrepancies. This method recommends to separate beryllium and aluminium using 8 hydroxyquinoline and then to directly precipitate glucinium in the filtrate using ammonia. One possible reason of the problems generated by this method should be the formation of a volatile complex between beryllium and the oxine. This work shows that when the oxine is eliminated before the precipitation with ammonia the dosimetry of beryllium becomes accurate. The destruction of the oxine requires the dry evaporation of the filtrate, which is a long process. Thus the search for a reagent allowing the quantitative precipitation of beryllium in its solutions and in presence of oxine has been made. It has been verified also that the quantitative precipitation of the double beryllium and ammonium phosphate is not disturbed by the oxine in acetic buffer. This method, which gives good results, has also the advantage to separate beryllium from the alkaline-earth compounds still present in the filtrate. The report details the operation mode of the method: beryllium dosimetry using ammonium phosphate, aluminium-beryllium separation, application to beryl dosimetry (ore processing, insolubilization of silica, precipitation with ammonia, precipitation with oxine, precipitation of PO{sub 4}NH{sub 4}Gl, preciseness). (J.S.)

  10. Refusion of zircaloy scraps by VAR (vacuum arc remelting): preliminary results; Fusao de cavacos de zircaloy por VAR: resultados preliminares

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, L.A.T.; Mucsi, C.S.; Sato, I.M.; Rossi, J.L.; Martinez, L.G., E-mail: lgallego@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Correa, H.P.S. [Universidade Federal do Mato Grosso do Sul (UFMS), Campo Grande, MS (Brazil); Orlando, M.T.D. [Universidade Federal do Espirito Santo (UFES), Vitoria, ES (Brazil)

    2010-07-01

    Fuel elements and structural components of the core of PWR nuclear reactors are made in zirconium alloys known as Zircaloy. Machining chips and shavings resulting from the manufacturing of these components can not be discarded as scrap, once these alloys are strategic materials for the nuclear area, have high costs and are not produced in Brazil on an industrial bases and, consequently, are imported for the manufacture of nuclear fuel. The reuse of Zircaloy chips has economic, strategic and environmental aspects. In this work is proposed a process for recycling Zircaloy scraps using a VAR (vacuum arc remelting) furnace in order to obtain ingots suitable for the manufacture of components of the reactors. The ingots obtained are being studied in order to verify the influence of processing on composition and microstructure of the remelted material. In this work are presented preliminary results of the composition of obtained ingots compared to start material and the resulting microstructure. (author)

  11. An Appraisal of Analytical Methods for Plutonium and their Applications to the Analysis of Nuclear Materials; Evaluation des Methodes Analytiques de Dosage du Plutonium et de Leur Application a l'Analyse des Matieres Nucleaires; Otsenka analiticheskikh metodov opredeleniya plutoniya i ikh primenenie dlya analiza yadernykh materialov; Metodos Analiticos de Determinacion del Plutonio y su Empleo en el Analisis de Materiales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Milner, G. W.C.; Phillips, G. [Atomic Energy Research Establishment, Harwell, Berks. (United Kingdom)

    1966-02-15

    programmes for new nuclear fuels. (author) [French] Il existe plusieurs methodes de dosage de la teneur en plutonium des matieres nucleaires. Pour les quantites de l'ordre du milligramme, les methodes utilisables sont la spectro- photometiie differentielle fondee sur la couleur de Pu (III), la gravimetrie fondee sur PuO{sub 2}, le comptage gamma et les methodes de reduction/oxydation comprenant les titrages poientiometriques et amperemetriques et la coulombmetrie a potentiel constant. Pour les quantites de Tordre du microgramme, le comptage alpha, la dilution isotopique et les methodes polarigraphiques sont a utiliser. Certaines methodes conviennent mieux que d'autres a des types determines d'echantillons et l'analyste soucieux d'obtenir les meilleurs resultats se heurte a un choix difficile. Les auteurs exposent les avantages et les inconvenients des methodes citees tels qu'ils se sont degages de l'experience acquise au cours des annees a l'Atomic Energy Research Establishment, et ils discutent l'exactitude, la precision, la sensibilite de ces methodes, et d'autres caracteristiques presentant un interet particulier. Certaines methodes ne peuvent etre utilisees si l'on n'a, dans une certaine mesure, separe le plutonium des autres constituants de l'echantillon et le memoire commente l'experience acquise avec l'echange d'anions et les procedes de chromatographie a phase inversee utilises a cette fin, en insistant surtout sur la mesure dans laquelle cette methode convient aux echantillons radioactifs. Les auteurs etudient en outre les nombreux problemes qui se sont poses lors de l'application (d'ailleurs couronnee de succes) de ces methodes a l'analyse des alliages de plutonium, des ceramiques et des cermets dans differentes combinaisons contenant de l'uranium, du thorium, du fer, du chrome, du molybdene, du cerium et du cobalt. Us exposent les difficultes de la dissolution des echantillons et de la reduction du plutonium a l'etat de valence voulu, ainsi que les avantages

  12. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  13. Quantitative analysis of nickel in zirconium and zircaloy

    International Nuclear Information System (INIS)

    Rastoix, M.

    1957-01-01

    A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [fr

  14. Evolution of deformation velocity in narrowing for Zircaloy 2

    Energy Technology Data Exchange (ETDEWEB)

    Cetlin, P R [Minas Gerais Univ., Belo Horizonte (Brazil). Dept. de Engenharia Metalurgica; Okuda, M Y [Goias Univ., Goiania (Brazil). Inst. de Matematica e Fisica

    1980-09-01

    Some studies on the deformation instability in strain shows that the differences in this instability may lead to localized narrowing or elongated narrowing, for Zircaloy-2. The variation of velocity deformation with the narrowing evolution is expected to be different for these two cases. The mentioned variation is discussed, a great difference in behavior having been observed for the case of localized narrowing.

  15. Simulation of the interaction between uranium dioxide and zircaloy

    International Nuclear Information System (INIS)

    Denis, A.; Garcia, E.A.

    1984-01-01

    The code solves the oxygen diffusion equations of the five phases formed during the UO 2 /Zircaloy interaction, using an implicit finite difference method with parabolic interpolation at the interfaces. Uranium and Zirconium mass conservation are considered. The code gives a good simulation of the experimental results for isothermal conditions. (orig.)

  16. Influence of hydrogen on the oxygen solubility in Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Guilbert-Banti, Séverine, E-mail: severine.guilbert@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Lacote, Pauline; Taraud, Gaëlle [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Berger, Pascal [NIMBE, CEA, CNRS, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Desquines, Jean; Duriez, Christian [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France)

    2016-02-15

    Despite the influence of hydrogen on the behavior of zirconium fuel cladding in many nuclear safety issues, the pseudo-binary Zircaloy-4 – oxygen phase diagram still lacks of data, especially above 1000 °C. The aim of this study was to provide experimental data to better assess the influence of hydrogen on the oxygen solubility in Zircaloy-4. Homogenized two-phase Zircaloy-4 samples were elaborated from 300 to 1000 wppm pre-hydrided samples. Local distributions were characterized thoroughly using Electron Probe Micro-Analysis (EPMA) for oxygen and Elastic Recoil Detection Analysis (ERDA) for hydrogen. The data obtained in this work were included in the pseudo-binary Zircaloy-4 – oxygen phase diagram and have shown that hydrogen has limited influence on the α + β → β transus. Regarding the α → α + β transus, no influence of hydrogen concentration in the α phase below 400 wppm was evidenced.

  17. Heat transfer coefficient between UO2 and Zircaloy-2

    International Nuclear Information System (INIS)

    Ross, A.M.; Stoute, R.L.

    1962-06-01

    This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

  18. Microstructural characterization of second phase irradiated Zircaloy-4 particles

    International Nuclear Information System (INIS)

    Flores, Alejandra V.; Vizcaino, Pablo; Banchik, Abraham D.; Bozzano, Patricia B.; Versaci, Raul A.

    2007-01-01

    X-ray diffraction diagrams of neutron irradiated Zircaloy-4 were obtained at the LNLS with the aim to obtain bulk information about the amorphization process in which the Zircaloy-4 second phase particles (SPPs) undergoes due to neutron irradiation. Owing to the low concentration of the SPPs in the alloy (∼0.4 V %), no data regarding to the bulk were obtained until now. The synchrotron experiences allowed to detect five of the more intense lines of the phase C 14 (SPPs structure) in unirradiated Zircaloy-4: (110) θ, (103) θ, (112) θ, (201) θ and (004) θ in the 34 degrees ≤ θ2≤45 degrees Bragg angle range and others of minor intensity. The diagrams of the samples irradiated at moderate doses (1020n/cm 2 ) show these lines even in the as received samples. In contrast, none of these lines are observed for high fluence samples (∼1022neutrons/cm 2 ). In addition, in similar high fluence samples annealed 24 h or 72 h at 600 C degrees the intensity rises just at the 2q range where the C 14 lines were observed, showing a wide peak. That peak is interpreted as a result of the superposition of unresolved diffraction lines corresponding to the Zircaloy SPPs which are in a reconstitution process of crystallization. Analytical Electron Microscopy techniques were used, in order to study the effects on the Zircaloy-4 SPPs and compared with samples of the same material without irradiation. Spots in SAD patterns of non irradiated SPPS, evidences the presence of a C 14 structure, but in irradiated SSP SAD patterns evidences the beginning of an amorphization process. Another important feature to point out is the different Fe / Cr ratio presented in both irradiated and non irradiated SSPs. In non irradiated precipitates the Fe / Cr ratio is approximately 1.5, while in irradiated precipitates the Fe / Cr ratio becomes near 1.0. (author) [es

  19. Microstructural aspects of zircaloy nodular corrosion in steam

    International Nuclear Information System (INIS)

    Taylor, D.F.

    1999-01-01

    Zircaloy-2 becomes susceptible to nodular corrosion in high-temperature, high-pressure steam when the total solute concentration of the β-stabilizing alloying elements Fe, Ni and Cr in the α-zirconium matrix falls below a critical value C c that is characteristic of the test conditions. C c for typical commercial Zircaloy-2 in a 24hr/510 C/10.4MPa steam-test is the precipitate-free a-matrix concentration in equilibrium with solute-saturated β phase at about 840 C, the corresponding critical temperature T c .Thus, immunity to nodular corrosion is a metastable condition for α-Zircaloy that requires fast cooling from above T c to achieve adequate solute concentration throughout the matrix. Annealing Zircaloy at any temperature below T c for a sufficiently long time makes it susceptible to nodular corrosion. In the (α+χ) phase field, where χ collectively designates the Fe-, Cr-, and Ni-containing precipitate phases, lowering the solute concentration to less than C c by Ostwald ripening can require many hundreds of hours. Above about 825 C, the temperature of the (α+χ)/(α+β+χ) transus, solute-saturated β phase surrounds each precipitate and a strong inverse activity gradient promotes equilibration with the much lower solute concentration in the α matrix. Sensitization to nodular corrosion occurs most rapidly at about 835 C between the (α+χ)/(α+β+χ) transus and T c . Annealing Zircaloy at temperatures above T c for a sufficiently long time will raise the solute concentration above C c and, with rapid cooling, heal any degree of susceptibility. Annealing within the protective coarsening window between T c and about 850 C, the temperature of the (α+β+χ)/(α+β) transus, achieves rapid precipitate growth in a matrix immune to nodular corrosion

  20. Corrosion Characteristics and Kinetics of Zircaloys and Aluminium Alloys

    International Nuclear Information System (INIS)

    Sugondo; Chaidir, A

    1998-01-01

    Corrosion rate characterization of cladding materials has been done by dynamic method. The materials are zircaloy-2,zircaloy-4,AIMg2,and AIMgSi.The zircaloy alloys are characterized in the electrolytes of boric ion,iodide ion,lithium ion and cesium ion with a pH variation.The aluminum alloys are characterized in the cooling water of RSG-GAS reactor in different temperatures and Ph values .The results, show that corrosion product of iodine on zircaloy is not passivated, meanwhile the corrosion product of cesium undergoes passivation. However, the deposited substance in the surface of the specimens as indicated using WDX-SEM shows the same deposition rate.it is concluded therefore that iodine is diffused into the materials without getting resistance from the deposited substances on the surface. The effect of pH to corrosion rate of iodine on the zircaloy fluctuates meanwhile the cesium has the minimum corrosion rate at pH 7.5 At the concentration of 0.1 gram/1,cesium ion is more reactive than iodine but at higher concentration the reactivity becomes competitive . Furthermore , the interaction between zircaloy and boric ion at concentration of 300 ppm and lithium ion at 10 ppm shows an outstanding corrosion rate, i.e. 0.1 mpy. if both substances are mixed then the corrosion rate decreases drastically in the order of 10 -2 mpy.The reason of such a decrease may be due to the formation of complexes of boron lithium on the electrode surface. The arrhenius activation energies for such reaction have been found to be 37629.322 joule/mole 0 K for Al Mg 2 and 41609.822 joule /mole 0 K for AIMgSi ,respectively. This underlies the argument that AI Mg 2 is more reactive than AI Mg Si besides , AI Mg 2 is more reactive under acid condition meanwhile AI Mg Si more reactive under basic condition. Both alloys over come the minimum corrosion rate at the pH in between 4.7 to 7.5 and the level of the corrosion rate in the pH interval was outstanding

  1. Determination of Boron in Zircaloy by using ICP-AES and Colorimetry

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong-Goo; Pyo, Hyung-Ryul; Choi, Kwang-Soon; Han, Sun-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    Zircaloy has been being widely used in the nuclear industry because of the low cross section of Zirconium against a thermal neutron. Accurate composition data of Zircaloy for Hf, B, and so on having a high cross section against thermal neutron is important to use it as a nuclear material. Accordingly proper determination methods of these elements in Zircaloy are needed. In this study, the application of two methods, ICP-AES and a colorimetry using methylene blue were investigated in order to establish a proper analysis method of Boron in the range from tens to hundreds ug B/g sample of Zircaloy.

  2. Determination of Boron in Zircaloy by using ICP-AES and Colorimetry

    International Nuclear Information System (INIS)

    Kim, Jong-Goo; Pyo, Hyung-Ryul; Choi, Kwang-Soon; Han, Sun-Ho

    2007-01-01

    Zircaloy has been being widely used in the nuclear industry because of the low cross section of Zirconium against a thermal neutron. Accurate composition data of Zircaloy for Hf, B, and so on having a high cross section against thermal neutron is important to use it as a nuclear material. Accordingly proper determination methods of these elements in Zircaloy are needed. In this study, the application of two methods, ICP-AES and a colorimetry using methylene blue were investigated in order to establish a proper analysis method of Boron in the range from tens to hundreds ug B/g sample of Zircaloy

  3. Nondestructive characterization of hydrogen concentration in zircaloy cladding tubes with laser ultrasound technique

    International Nuclear Information System (INIS)

    Yang, Che Hua; Lai, Yu An

    2006-01-01

    This paper describes a laser ultrasound technique (LUT) for nondestructive characterization of hydrogen concentration (HC) in Zircaloy cladding tubes. With the LUT, guided ultrasonic waves are generated remotely and then propagate in the axial direction of Zircaloy tubes, and finally detected remotely by an optical probe. By measuring the dispersion spectra with the LUT, relations between the dispersion spectra and the HC of the Zircaloy tubes can be established. The LUT is non-contact, capable of remote inspection, and therefore suitable for nondestructive inspection of HC in Zircaloy cladding tubes used in nuclear power plant.

  4. Apparatus for study of transient oxidation of Zircaloy-4 tubing

    International Nuclear Information System (INIS)

    Sagat, S.; Iglesias, F.C.; Newell, G.W.

    1985-11-01

    Complex transient oxidation tests on Zircaloy-4 tubing were performed to provide data for validation of the computer code FROM2. This code was developed to calculate oxygen distribution through oxidized Zircaloy tubing. The test temperature histories consisted of ramp, hold and cool cycles. The heating and cooling rates were in the range of 1 to 100 K/s and the maximum temperature was 1875 K. The apparatus developed to perform these experiments is described. In principle, Joule heating is used to heat the specimen and the temperature is controlled by a computer in conjunction with temperature and SCR power controllers. Using this combination, fast heating and cooling rates were achieved without sacrificing the accuracy of temperature control

  5. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

    2005-09-01

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  6. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    Grigoriev, V.; Andersson, Stefan

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

  7. Modeling of Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Santanam, L.; Raghavan, S.; Chin, B.A.

    1989-07-01

    Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

  8. Spectrochemical determination of impurities in zircaloy 2 and 4

    International Nuclear Information System (INIS)

    Paula Reino, L.C. de; Lordello, A.R.

    1987-06-01

    A method has been developed for the determination of Hf,Co,Mo,Pb,Ti,V,Al,Si,W,Cu,Mg,Mn,B and Cd in zircaloy 2 and 4. For hafnium determination 10% CuF 2 is added as spectrographic buffer on a previously oxidized zircaloy; the samples are loaded in a shallow cup electrode of Scribner Mullins type and excited in a direct current arc. The carrier distillation technique has been used for the other elements. Better results were obtained with 25% AgCl as carrier. The precision of the method varies from 4% for copper to 29% for boron but it does not exceed 17% for most elements. (Author) [pt

  9. Electrolytic hydriding and hydride distribution in zircaloy-4

    International Nuclear Information System (INIS)

    Gomes, M.H.L.

    1974-01-01

    A study has been made of the electrolytic hydriding of zircaloy-4 in the range 20-80 0 C, for reaction times from 5 to 30 hours, and the effect of potential, pH and dissolved oxygen has been investigated. The hydriding reaction was more sensitive to time and temperature conditions than to the electrochemical variables. It has been shown that a controlled introduction of hydrides in zircaloy is feasible. Hydrides were found to be plate like shaped and distributed mainly along grain-boundaries. It has been shown that hydriding kinetics do not follow a simple law but may be described by a Johnson-Mehl empirical equation. On the basis of this equation an activation energy of 9.400 cal/mol has been determined, which is close to the activation energy for diffusion of hydrogen in the hydride. (author)

  10. Zircaloy cladding ID/OD oxidation studies. Final report

    International Nuclear Information System (INIS)

    Westerman, R.E.; Hesson, G.M.

    1977-11-01

    The ID/OD oxide ratio that forms on Zircaloy tubing at temperatures relevant to postulated LOCA conditions was measured as a function of time, temperature, and distance from the rupture. The average ratio at the rupture position was less than unity, and decreased with decreasing test time and increasing distance from the point of rupture. The maximum observed ID/OD oxide ratio was 1.4. Ratios in excess of unity were typically found to be a consequence of the OD oxide being thinner than would have been anticipated from the nominal test conditions. Confirmatory data were also obtained on the isothermal oxidation kinetics of Zircaloy. These data are in good agreement with those obtained by other investigators and confirm the conservative nature of the Baker-Just equation that is required for use in licensing calculations

  11. Iodine stress-corrosion cracking in irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Mattas, R.F.; Yaggee, F.L.; Neimark, L.A.

    1979-01-01

    Irradiated Zircaloy cladding specimens, which had experienced fluences from 0.1 to 6 x 10 21 n/cm 2 (E>0.1 MeV), were gas-pressure tested in an iodine environment to investigate their stress-corrosion cracking (SCC) susceptibility. The test temperatures and hoop stresses ranged from 320 to 360 0 C and 150 to 500 MPa, respectively. The results indicate that irradiation, in general, increases the susceptibility of Zircaloy to iodine SCC. For specimens that experienced fluences >2 x 10 21 n/cm 2 (E>0.1 MeV), the 24-h failure stress was 177+-18 MPa, regardless of the preirradiation metallurgical condition. An analytical model for iodine SCC has been developed which agrees reasonably well with the test results

  12. Interactions of Zircaloy cladding with gallium: 1998 midyear status

    International Nuclear Information System (INIS)

    Wilson, D.F.; DiStefano, J.R.; Strizak, J.P.; King, J.F.; Manneschmidt, E.T.

    1998-06-01

    A program has been implemented to evaluate the effect of gallium in mixed-oxide (MOX) fuel derived from weapons-grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in a light-water reactor. The graded, four-phase experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of a series of tests for Phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests will determine corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Although continued migration of gallium into the initially formed intermetallic compound can result in large stresses that may lead to distortion, this was shown to be extremely unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

  13. Oxidation of zircaloy-2 in high temperature steam

    International Nuclear Information System (INIS)

    Ikeda, Seiichi; Ito, Goro; Ohashi, Shigeo

    1975-01-01

    Oxidation tests were conducted for zircaloy-2 in steam at temperature ranging from 900 to 1300 0 C to clarify its oxidation kinetics as a nuclear fuel cladding materials in case of a loss-of-coolant accident. The influence of maximum temperature and heating rate of the specimen on its oxidation rate in steam was investigated. The changes in mechanical properties of the specimens after oxidation tests are also studied. The results obtained were summarized as follows: (1) The weight of the specimen after oxidation in steam increased two times as the time required to reach the maximum temperature increased from 1 to 10 mins. (2) The kinetics of oxidation of zircaloy-2 in steam were not affected by the difference in the surface condition before test such as chemical polishing or pre-oxidation in steam. (3) The dominant growth of oxide film on the surface of zircaloy-2 was observed at the initial stage of oxidation in steam. However, the thickness of oxygen-rich solid solution layer under the film increased gradually with the progress of oxidation and the ratio of oxygen in oxide to that in solid solution has a constant value of 8:2. (4) The breakaway took place only in the specimen subjected to 900 0 C repeated heating. This penomenon was caused by the local growth of the oxide below a crack of the oxide film resulting from the reheating of the specimen. (5) The results of bending tests showed that the deflection until fracture of the specimen was smaller for the one heated at a higher temperature even if the weight increase was of the same order of magnitude for both specimens. (6) It was concluded that the ductility of zircaloy-2 decreased remarkably at a heating temperature in excess of 1100 0 C for more than 5 min. (auth.)

  14. Interim report on the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Hobson, D.O.; Dodd, C.V.

    1977-01-01

    This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

  15. Interactions of zircaloy cladding with gallium -- 1997 status

    International Nuclear Information System (INIS)

    Wilson, D.F.; DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P.

    1997-11-01

    A four phase program has been implemented to evaluate the effect of gallium in mixed oxide (MOX) fuel derived from weapons grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in LWR. This graded, four phase experimental program will evaluate the performance of prototypic Zircaloy cladding materials against: (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of an initial series of tests for phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement (LME), and (3) corrosion mechanical. These tests are designed to determine the corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥ 300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (in parts per million) of gallium in the MOX fuel. While continued migration of gallium into the initially formed intermetallic compound results in large stresses that can lead to distortion, this is also highly unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

  16. Prediction of water droplet evaporation on zircaloy surface

    International Nuclear Information System (INIS)

    Lee, Chi Young; In, Wang Kee

    2014-01-01

    In the present experimental study, the prediction of water droplet evaporation on a zircaloy surface was investigated using various initial droplet sizes. To the best of our knowledge, this may be the first valuable effort for understanding the details of water droplet evaporation on a zircaloy surface. The initial contact diameters of the water droplets tested ranged from 1.76 to 3.41 mm. The behavior (i.e., time-dependent droplet volume, contact angle, droplet height, and contact diameter) and mode-transition time of the water droplet evaporation were strongly influenced by the initial droplet size. Using the normalized contact angle (θ*) and contact diameter (d*), the transitions between evaporation modes were successfully expressed by a single curve, and their criteria were proposed. To predict the temporal droplet volume change and evaporation rate, the range of θ* > 0.25 and d* > 0.9, which mostly covered the whole evaporation period and the initial contact diameter remained almost constant during evaporation, was targeted. In this range, the previous contact angle functions for the evaporation model underpredicted the experimental data. A new contact angle function of a zircaloy surface was empirically proposed, which represented the present experimental data within a reasonable degree of accuracy. (author)

  17. Treatment of zircaloy cladding hulls by isostatic pressing

    International Nuclear Information System (INIS)

    Tegman, R.; Burstroem, M.

    1984-12-01

    A method for the treatment of Zircaloy fuel hulls is proposed. It involves hot isostatic pressing (HIP) for making large, completely densified metallic bodies of the waste. The hulls are packed into a bellows-shaped container of steel. On packing the fuel hulls give a filling factor of only 14%, which is too low for non-deformable compaction in a normal container, but by using a belloped container, a non-deformable compaction can be obtained without any pretreatment of the hulls. Fully dense and mechanically strong blocks of Zircaloy can be fabricated by holding them at temperatures of around 1000 degrees C for three hours. It is also feasible to incorporate the other metallic parts of the fuel bundle, such as top and bottom tie plates and spacers, in the pressing. The HIP-densified hulls provide an effective means of self-containment of radioactive waste due to the excellent corrosion resistance of Zircaloy. A waste loading factor of close to 100% can be realized. Futher, a volume reduction factor of 7 and a surface reduction factor of aout 250 for a 1-ton canister can be achieved. Equilibrium calculations have shown that tritium present in the hulls can quantitatively be contained in the HIPed block. A study has been made of a possible process for industrilscale use. (Author)

  18. Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

    Science.gov (United States)

    Huang, F. H.

    1993-12-01

    Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

  19. Semi-empirical corrosion model for Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Nadeem Elahi, Waseem; Atif Rana, Muhammad

    2015-01-01

    The Zircaloy-4 cladding tube in Pressurize Water Reactors (PWRs) bears corrosion due to fast neutron flux, coolant temperature, and water chemistry. The thickness of Zircaloy-4 cladding tube may be decreased due to the increase in corrosion penetration which may affect the integrity of the fuel rod. The tin content and inter-metallic particles sizes has been found significantly in the magnitude of oxide thickness. In present study we have developed a Semiempirical corrosion model by modifying the Arrhenius equation for corrosion as a function of acceleration factor for tin content and accumulative annealing. This developed model has been incorporated into fuel performance computer code. The cladding oxide thickness data obtained from the Semi-empirical corrosion model has been compared with the experimental results i.e., numerous cases of measured cladding oxide thickness from UO 2 fuel rods, irradiated in various PWRs. The results of the both studies lie within the error band of 20μm, which confirms the validity of the developed Semi-empirical corrosion model. Key words: Corrosion, Zircaloy-4, tin content, accumulative annealing factor, Semi-empirical, PWR. (author)

  20. Cyclic deformation of zircaloy-4 at room temperature

    International Nuclear Information System (INIS)

    Armas, A. F; Herenu, S; Bolmaro, R; Alvarez-Armas, I

    2003-01-01

    Annealed materials hardens under low cyclic fatigue tests.However, FCC metals tested with medium strain amplitudes show an initial cyclic softening.That behaviour is related with the strong interstitial atom-dislocation interactions.For HCP materials the information is scarce.Commercial purity Zirconium and Zircaloy-4 alloys show also a pronounced cyclic softening, similar to Titanium alloys.Recently the rotation texture induced softening model has been proposed according to which the crystals are placed in a more favourable deformation orientation by prismatic slip due to the cyclic strain.The purpose of the current paper is the presentation of decisive results to discuss the causes for cyclic softening of Zircaloy-4. Low cycle fatigue tests were performed on recrystallized Zircaloy-4 samples.The cyclic behaviour shows an exponential softening at room temperature independently of the deformation range.Only at high temperature a cyclic hardening is shown at low number of cycles.Friction stresses, related with dislocation movement itself, and back stresses, related with dislocation pile-ups can be calculated from the stress-strain loops.The cyclic softening is due to diminishing friction stress while the starting hardening behaviour is due to increasing back stresses.The rotation texture induced softening model is ruled out assuming instead a model based on dislocation unlocking from interstitial oxygen solute atoms

  1. High temperature properties of Zircaloy--oxygen alloys

    International Nuclear Information System (INIS)

    Mellinger, G.B.; Bates, J.L.

    1977-03-01

    The effect of oxygen on three properties of Zircaloy-4 cladding relevant to LOCA evaluation codes was determined. Thermal expansion, elastic moduli, and thermal diffusivity were measured over the range room temperature--1200 0 C (2192 0 F) and 0.7 to 28 at.% oxygen. Thermal expansion and elastic moduli showed increases with oxygen concentration, while thermal diffusivity tended to decrease. Zircaloy-2 was examined over the same temperature range, but only to 5 at.% oxygen, differences in the properties between the two alloys were minor. The thermal emittance of Zircaloy-4 was measured in argon over the wavelength range 1.5 to 2.5 μm on previously oxidized tubing and on surfaces in the process of oxidizing in unlimited steam. For the latter, a high emittance (approximately 0.9) was reached at an oxide thickness of about 100 mg/dm 2 , and the tubing surface remained black and substoichiometric as oxidation continued at temperatures to 1200 0 C

  2. Diffused zircaloy 2/stainless steel junctions; Jonctions diffusees zircaloy 2 - acier inoxydable

    Energy Technology Data Exchange (ETDEWEB)

    Jacques, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The diffusion permits to realize joints between two different materials, in fact of the formation of a liquid phase at the contact face. The study of the tensile properties allowed the determination of the ideal conditions for the diffusion treatment which are, within 2 and 3 minutes for a temperature within 1020 C and 1030 C. The characteristics of the so obtained joints were, studied: mechanical properties, tightness, resistance to thermal cycling. Analysis of the thermal stress, owing to the differential dilatation of the two materials mode the object of a particular study. The investigation on the diffusion zone, includes specially, an analysis of the constituents distribution formed during the diffusion treatment. (author) [French] La diffusion permet de realiser des joints entre deux materiaux differents, du fait de la formation d'une phase liquide a l'interface de contact. L'etude de la resistance a la traction a permis de determiner les conditions optimum du traitement de diffusion: une duree de 2 a 3 minutes pour une temperature comprise entre 1020 C et 1030 C. Les caracteristiques des jonctions ainsi obtenues ont ete etudiees: proprietes mecaniques, etancheite, resistance au cyclage thermique. L'analyse des contraintes thermiques dues a la difference de dilatation des deux materiaux, a fait l'objet d'une etude particuliere. L'etude metallurgique de la zone diffusee comporte en particulier une analyse de la repartition des constituants formes lors du traitement de diffusion. (auteur)

  3. The long range migration of hydrogen through Zircaloy in response to tensile and compressive stress gradients

    International Nuclear Information System (INIS)

    Kammenzind, B.F.; Berquist, B.M.; Bajaj, R.; Kreyns, P.H.; Franklin, D.G.

    1998-01-01

    Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed

  4. Zircaloy-steam reaction under a simulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kawasaki, Satoru; Furuta, Teruo; Hashimoto, Masao

    1975-07-01

    Under a simulated loss-of-coolant condition, the reaction between zircaloy and steam and the embrittlement of the zircaloy oxidized by this reaction have been studied. The parabolic rate constant, ksub(p), in the zircaloy-steam reaction is represented as ksub(p)=3.24x10 6 exp(-40500/RT) (mg 2 /cm 4 . sec) Ring compression test was made on the steam-reacted zircaloy tubes, and following results were obtained: Embrittlement of the steam-reacted zircaloy tube increases with oxidation at each oxidation temperature. For a given quantity of the oxidation, the incursion of α-phase into β-phase is more remarkable in the specimens reacted at low temperatures than those at high temperatures. The embrittlement, however, is larger in the specimens oxidized at high temperatures than those at low temperatures. (auth.)

  5. Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy; Caracterizacion superficial por XPS de nanoparticulas de plata y su deposito hidrotermal sobre zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L., E-mail: aida.contreras@inin.gob.mx [ININ, Departamento de Tecnologia de Materiales, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

  6. Preparation, extraction and dosage of labelled cholesterol (D and C{sup 14}); Preparation, extraction et dosage de cholesterol marque (D et C{sup 14})

    Energy Technology Data Exchange (ETDEWEB)

    Bugnard, L; Chevallier, F; Coursaget, J [Commissariat a l' Energie Atomique, Saclay(France). Centre d' Etudes Nucleaires

    1953-07-01

    We returned in this note the techniques that we used for the preparation of labelled cholesterol. The chemical exchange of hydrogen enabling to contain deutero-cholesterol until 4 percent deuterium. The biologic synthesis, done on living rats or on their liver maintained in survival, permits, on the other hand, to get active cholesterol from acetate of containing sodium of the carbon 14. We indicated the techniques of extraction and dosage of the marked cholesterol. The radioactivity is measured with a Geiger-Muller counter. (M.B.) [French] Nous avons rapporte dans cette note les techniques que nous avons utilisees pour la preparation de cholesterol marque. L'echange chimique d'hydrogene conduit a du deuterio-cholesterol pouvant contenir jusqu'a 4 pour cent de deuterium. La synthese biologique, effectuee sur des rats vivants ou sur leur foie maintenu en survie, permet, d'autre part, d'obtenir du cholesterol radio-actif a partir d'acetate de sodium contenant du carbone 14. Nous avons indique les techniques d'extraction et de dosage du cholesterol marque. Sa radioactivite est mesuree au compteur de Geiger-Muller. (M.B.)

  7. Temperature effect on Zircaloy-4 stress corrosion cracking

    International Nuclear Information System (INIS)

    Farina, Silvia B.; Duffo, Gustavo S.; Galvele, Jose R.

    1999-01-01

    Stress corrosion cracking (SCC) susceptibility of Zircaloy-4 alloy in chloride, bromide and iodide solutions with variables as applied electrode potential, deformation rate and temperature have been studied. In those three halide solutions the susceptibility to SCC is only observed at potentials close to pitting potential, the crack propagation rate increases with the increase of deformation rate, and that the temperature has a notable effect only for iodide solutions. For chloride and bromide solutions and temperatures ranging between 20 to 90 C degrees it was not found measurable changes in crack propagation rates. (author)

  8. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  9. Threshold values characterizing iodine-induced SCC of zircaloys

    International Nuclear Information System (INIS)

    Une, K.

    1984-01-01

    Threshold values of stress, stress intensity factor, strain, strain rate and iodine concentration for SCC of unirradiated and irradiated Zircaloys are reviewed. The ratio of σsub(th)/σsub(y) adequately represents the effects of cold work and irradiation damage on the SCC susceptibility, where threshold stress σsub(th) is defined as the minimum stress to cause SCC to failure after -6 and 10 -3 min -1 . A comparison of SCC data between constant strain rate and constant stress tests is presented in order to examine the validity of a cumulative-damage concept under SCC conditions. (author)

  10. Identification of the zirconium hydrides metallography in zircaloy-2

    International Nuclear Information System (INIS)

    Garcia Gonzalez, F.

    1968-01-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs

  11. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

  12. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    International Nuclear Information System (INIS)

    Clayton, J.C.

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

  13. Determination of I-SCC crack propagation rate of zircaloy-4

    International Nuclear Information System (INIS)

    Woo-Seog, Ryu

    2002-01-01

    Threshold stress intensity (K ISCC ) and propagation rate of iodine-induced SCC in recrystallized and stress-relieved Zircaloy-4 were determined using a DCPD method. Dynamic system flowing Ar gas through iodine chamber at 60 deg C provided a constant iodine pressure of 1000 Pa during test. The SCC curves of crack velocity vs. stress intensity showed the typical SCC curves that are composed of stages I, II and III. The threshold K ISCC at 350 deg C was about 9 and 9.5 MPa √m for the stress- relieved Zircaloy-4 and the recrystallized Zircaloy-4, respectively. The plateau velocity in the stage II at 350 deg C was 4-8x 10 -4 mm/sec in the range of 20-40 MPa√m. In comparison with recrystallized Zircaloy-4, stress-relieved Zircaloy-4 had a lower threshold stress intensity factor and a little higher SCC velocity, indicating that SRA Zircaloy-4 was more sensitive to SCC in respect of velocity. The fracture mode in recrystallized Zircaloy was mostly a transgranular fracture with river pattern. An intergranular mode and the flutting were scarcely observed. (author)

  14. Studies of irradiated zircaloy fuel sheathing using XPS

    International Nuclear Information System (INIS)

    Chan, P.K.; Irving, K.G.; Hocking, W.H.; Duclos, A.M.; Gerwing, A.F.

    1995-01-01

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO 2 ) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs

  15. a Study on the Fretting Fatigue Life of Zircaloy Alloys

    Science.gov (United States)

    Kwon, Jae-Do; Park, Dae-Kyu; Woo, Seung-Wan; Chai, Young-Suck

    Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr, and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.

  16. Irradiation creep and growth of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Lansiart, S.; Darchis, L.; Pelchat, J.

    1990-01-01

    The influence of temperature and fast neutron flux on irradiation creep and growth of stress relieved zircaloy-4 pressurized tubes has been derived from experimental irradiations in NaK, performed up to 2.5 10 25 n.m -2 in the temperature range [280, 350] 0 C. A significant influence of temperature on axial growth has been observed: at 280 0 C the elongation can no longer be expressed as a linear function of fluence as for the 350 0 C irradiation temperature; diametral growth, on the other hand, always appears negligible. Irradiation creep obviously depends on temperature too; the diametral strain (including thermal part) has been modelled as a sum of primary and secondary terms, the former being independent of fluence. For the tubing considered it is observed that the ranking of the different batches, with respect to diametral creep resistance, is the same before and under irradiation. Concerning axial creep strain the stress relieved material behaves as does an isotropic tube. This is not the case of recrystallized zircaloy-4 F, which shows a non negligible axial deformation, related to the diametral creep one, even though this diametral irradiation creep strain is strongly reduced comparatively to that of the stress relieved material. The comparison of the two materials growth rates is more complex since their dependence on temperature and flux differs

  17. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Grigoriev, V. [Studsvik Material AB, Nykoeping (Sweden); Andersson, Stefan [Royal Inst. of Tech., Stockholm (Sweden)

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill`s theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture. With a 2 page summary in Swedish. 32 refs, 18 figs.

  18. A phenomenological model for iodine stress corrosion cracking of zircaloy

    International Nuclear Information System (INIS)

    Miller, A.K.; Tasooji, A.

    1981-01-01

    To predict the response of Zircaloy tubing in iodine environments under conditions where either crack initiation or crack propagation predominates, a unified model of the SCC process has been developed based on the local conditions (the local stress, local strain, and local iodine concentration) within a small volume of material at the cladding inner surface or the crack tip. The methodology used permits computation of these values from simple equations. A nonuniform distribution of local stress and strain results once a crack has initiated. The local stress can be increased due to plastic constraint and triaxiality at the crack tip. Iodine penetration is assumed to be a surface diffusion-controlled process. Experimental data are used to derive criteria for intergranular failure, transgranular failure, and ductile rupture in terms of the local conditions. The same failure criteria are used for both crack initiation and crack propagation. Irradiation effects are included in the model by changing the value of constants in the equation governing iodine penetration and by changing the values used to represent the mechanical properties of the Zircaloy. (orig./HP)

  19. Conversion of zircaloy to a massive chemically inert form

    International Nuclear Information System (INIS)

    Atkinson, A.; Kearsey, H.A.; Knibbs, R.H.; Mercer, A.C.; Nickerson, A.K.; Pearson, D.; Sambell, R.A.J.; Taylor, R.I.

    1985-01-01

    The report covers work carried out in the period July 1980 - December 1982 on the development and assessment of an aqueous route for the conversion of Zircaloy fuel element cladding to a stable oxide form and on alternative methods for incorporating the oxide into monolithic waste forms suitable for long-term storage and disposal. The work included two aspects, preliminary process development studies aimed at demonstrating the key steps in the process, and studies on the alternative immobilization techniques and the properties of the resulting waste forms. Experimental studies have shown that the ''hydrous zirconium oxide'' (with a residual fluoride content), following calcination at about 500 0 C, can be hot-pressed at 800-1000 0 C and 22.5 MPa to a high density ceramic waste form with good capacity for the incorporation of active species, such as U 4+ and Sr 2+ , and high leach resistance. Parallel studies have been carried out on the incorporation of the washed ''hydrous zirconium oxide'' in a range of cement matrices. A preliminary chemical engineering assessment of the overall process has been made and flowsheets for a plant to convert 250 kg Zircaloy/day have been prepared

  20. Investigation of Zircaloy-2 oxidation model for SFP accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

    2017-05-15

    The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

  1. Studies of irradiated zircaloy fuel sheathing using XPS

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P K; Irving, K G [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Hocking, W H; Duclos, A M; Gerwing, A F [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO{sub 2}) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs.

  2. Adsorption and diffusion of hydrogen in Zircaloy-4

    International Nuclear Information System (INIS)

    Torres, E.; Desquines, J.; Baietto, M.C.; Coret, M.; Wehling, F.; Blat-Yrieix, M.; Ambard, A.

    2015-01-01

    Hydrogen in zirconium alloys is considered in many nuclear safety issues. Below 500 Celsius degrees, rather limited knowledge is available on the combined hydrogen adsorption at the sample surface and diffusion in the metal. A modeling of hydrogen gaseous charging has been established starting with a set of relevant laws and parameters derived from open literature. Simulating the hydrogen charging process requires simultaneous analysis of gaseous surface adsorption, hydrogen solid-solution diffusion and precipitation, when exceeding the material solubility limit. The modeling has been extended to reproduce the solid-gas exchange. Gaseous charging experiments have been performed at 420 C. degrees on Stress Relieved Annealed (SRA) Zircaloy-4 cladding samples to validate the model. The sample hydrogen content has been systematically measured after charging and compared to the calculated value thus providing a validation of the adsorption modeling. Complementary tests have been carried out on Recrystallized Annealed (RXA) Zircaloy-4 rods to characterize the combined diffusion and adsorption process. The hydrogen concentration distribution has been characterized using an inverse technique based on destructive analyses of the samples. This additional set of data was relevant for the validation of the hydrogen combined adsorption/diffusion modeling up to 420 C. degrees. (authors)

  3. Quantification and characterization of zirconium hydrides in Zircaloy-4 by the image analysis method

    International Nuclear Information System (INIS)

    Zhang, J.H.; Groos, M.; Bredel, T.; Trotabas, M.; Combette, P.

    1992-01-01

    The image analysis method is used to determine the hydrogen content in specimens of Zircaloy-4. Two parameters, surface density of hydride, S v , and degree of orientation, Ω, are defined to represent separately the hydrogen content and the orientation of hydrides. By analysing the stress-relieved Zircaloy-4 specimens with known hydrogen content from 100 to 1000 ppm, a relationship is established between the parameter S v and the hydrogen content when the magnifications of the optical microscope are 1000 and 250. The degree of orientation for the hydride in the stress-relieved Zircaloy-4 cladding is about 0.3. (orig.)

  4. Effect of the aluminum flow pattern on the bonding of aluminum to oxidized Zircaloy-2

    International Nuclear Information System (INIS)

    Watson, R.D.; Lambert, J.P.

    1965-04-01

    The bonds produced when hot aluminum is allowed to flow smoothly from an extrusion die to the oxidized surface of a heated tube of Zircaloy-2 are consistently inferior to those produced with back-extruded flow. The difference is believed to be due to the reduction in, or elimination of, the oxide layer on the aluminum that comes in contact with the surface of the Zircaloy-2. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 1965. (author)

  5. Modification of hydrogen determinator for total hydrogen analysis in irradiated zircaloy cladding tube

    International Nuclear Information System (INIS)

    Park, Soon Dal; Choi, Kwnag Soon; Kim, Jong Goo; Joe, Kih Soo; Kim, Won Ho

    1999-01-01

    A hydrogen determinator was modified and installed in the glove box to analyse total hydrogen content in irradiated zircaloy tube. The analysis method of hydrogen is Inert Gas Fusion(IGF)-Thermal Conductivity Detection(TCD). The hydrogen recoveries of no tin method using Ti and Zr matrix standards, respectively, were available within 3 μg of hydrogen. Also the smaller size of sample showed the better hydrogen recovery. It was found that the hydrogen standard of Ti matrix is available to hydrogen analysis in zircaloy sample. The mean radioactivity of irradiated zircaloy sample was 10 mR/hr and hydrogen concentration was 130 ppm

  6. Nucleation and growth of intermetallic precipitates in Zircaloy-2 and zircaloy-4 and correlation to nodular corrosion behavior

    International Nuclear Information System (INIS)

    Maussner, G.; Steinberg, E.; Tenckhoff, E.

    1987-01-01

    One of the fundamental aspects in the history of the development of zirconium alloys for nuclear applications is the corrosion behavior under in-pile conditions. In boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) the zirconium alloys Zircaloy-2 and Zircaloy-4 are the most commonly used materials, permitting attainment of a very high level of integrity and reliability. Nevertheless, efforts are required to optimize these well-established alloys with regard to their resistance to nodular corrosion, where improvements will give long-term advantages in fuel integrity and fuel economy. Phenomenological studies allow correlation of the nodular corrosion behavior with the morphological appearance of precipitated intermetallic phases in the microstructures of Zry-2 and Zry-4. To understand the fundamental processes of precipitation, particle nucleation-and-growth studies were made with Zry-2 and Zry-4 in different fabrication dimensions and with variations in β-quenching rates followed by isothermal and isochronical heat treatments. The microstructural characteristics of the precipitates were investigated by optical and transmission-electron microscopy. The macroscopic behavior was studied by electrical-resistivity measurements and hardness measurements. The nodular-corrosion susceptibility was determined by weight-gain and nodule distribution measurements after a 500 0 C laboratory-autoclave test

  7. Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    McClelland, R.G.; O'Leary, P.M.

    1992-01-01

    Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an ∼0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current 'lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4

  8. Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy

    International Nuclear Information System (INIS)

    Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L.

    2012-10-01

    The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

  9. Zirconium metal-water oxidation kinetics. III. Oxygen diffusion in oxide and alpha Zircaloy phases

    International Nuclear Information System (INIS)

    Pawel, R.E.

    1976-10-01

    The reaction of Zircaloy in steam at elevated temperature involves the growth of discrete layers of oxide and oxygen-rich alpha Zircaloy from the parent beta phase. The multiphase, moving boundary diffusion problem involved is encountered in a number of important reaction schemes in addition to that of Zircaloy-oxygen and can be completely (albeitly ideally) characterized through an appropriate model in terms of oxygen diffusion coefficients and equilibrium concentrations for the various phases. Conversely, kinetic data for phase growth and total oxygen consumption rates can be used to compute diffusion coefficients. Equations are developed that express the oxygen diffusion coefficients in the oxide and alpha phases in terms of the reaction rate constants and equilibrium solubility values. These equations were applied to recent experimental kinetic data on the steam oxidation of Zircaloy-4 to determine the effective oxygen diffusion coefficients in these phases over the temperature range 1000--1500 0 C

  10. Comparison study between GTWA and PAW welding techniques in zircaloy-4

    International Nuclear Information System (INIS)

    Martinez, R.L.; Boccanera, L.; Ortiz, L.; Fernandez, L.; Corso, H.

    2003-01-01

    The wide use of zirconium alloys in different structural parts of nuclear reactors mainly under severe environmental conditions has encouraged the study of Zircaloy-4 and specifically welded joints of this material.Many different factors affect mechanical properties, specifically hydrides, formed by absorbed hydrogen.Hydrogen solubility in Zircaloy-4 is low and because Zircaloy-4 picks up hydrogen during service the potential exist that zirconium hydrides phase precipitate causing loss of ductility, the most undesirable consequence. Therefore, the study and characterization of welded joint of nuclear materials assumes fundamental importance in the safety of nuclear reactors.This paper presents experimental results regarding of hardness and hydrogen concentration in Zircaloy-4 plates obtained by two different welding techniques GTWA (Gas Tungsten Arc Welding) and PAW (Plasma Arc Welding).In this work following these remarks the difference observed between these two techniques are presented and point out some aspects of PAW for further discussion

  11. Modelling of zircaloy-4 corrosion in nitrogen and oxygen mixtures at high temperature

    International Nuclear Information System (INIS)

    Lasserre, M.; Peres, V.; Pijolat, M.; Coindreau, O.; Duriez, C.; Mardon, J.P.

    2015-01-01

    Previous studies of zircaloy-4 corrosion in air have shown accelerated corrosion in the 600-1000 Celsius degrees temperature range with Zr nitrides precipitating near the metal/oxide surface. The aim of this series of slides is to assess the influence of N 2 and O 2 partial pressures on the kinetic rate of growth of a new phase and to propose a kinetic modelling of zircaloy-4 corrosion

  12. Characterisation of metallic glass incorporated Zircaloy-2 weldments

    International Nuclear Information System (INIS)

    Mishra, S.; Savalia, R.T.; Bhanumurthy, K.; Dey, G.K.; Banerjee, S.

    1995-01-01

    In this study the effect of incorporation of Zr based Fe and Ni bearing metallic glass in spot welds in Zircaloy components has been examined. A comparison of strength and microstructure of the welded joint with and without glass has been carried out. The welded joint with metallic glass has been found to be stronger than the one without metallic glass. The microstructure of the welded region with metallic glass has been found to comprise a large region having martensite. This large martensitic region has also been found to have considerable amount of excess solute (Fe, Ni). The higher strength of the weld with metallic glass seems to originate due to solid solution strengthening, small grain size and the presence of martensitic structure over a large region. (orig.)

  13. Effect of current density on the anodization of zircaloy-2

    International Nuclear Information System (INIS)

    Bhaskar Reddy, P.; Panasa Reddy, A.

    2005-01-01

    The effect of current density on the kinetics of anodization of Zircaloy-2 in 0.1 M potassium tartarate have been studied at various constant current densities ranging from 2 to 10 mA.cm -2 and at room temperature to investigate the exponential dependence of ionic current density on the field across the oxide. The rate of anodic film formation (dV/dt), the current efficiency the differential field of formation (F) and the ionic current density (i i ) were calculated. It was found that all these parameters were increased with increase of current density. The induction period was decreased with the increase of current density. It was also found that the plot of log (ionic current density) vs differential field gave fairly a linear relationship. The kinetic parameters, half jump distance (a) and height of the energy barrier (W) were calculated. (author)

  14. Creep modeling of textured zircaloy under biaxial stressing

    International Nuclear Information System (INIS)

    Adams, B.L.; Murty, K.L.

    1984-01-01

    Anisotropic biaxial creep behavior of textured Zircaloy tubing was modeled using a crystal-plastic uniform strain-rate upper-bound and a uniform stress lower-bound approach. Power-law steady-state creep is considered to occur on each crystallite glide system by fixing the slip rate to be proportional to the resolved shear stress raised to a power. Prismatic, basal, and pyramidal slip modes were considered. The crystallographic texture is characterized using the orientation distribution function determined from a set of three pole-figures. This method is contrasted with a Von-Mises-Hill phenomenological model in comparison with experimental data obtained at 673 deg K. The resulting creep-dissipative loci show the importance of the basal slip mode on creep in heavily cold-worked cladding, whereas prismatic slip is more important for the recrystallized materials. (author)

  15. Deformation texture and microtexture development in zircaloy-2

    International Nuclear Information System (INIS)

    Vanitha, C.; Kiran Kumar, M.; Samajdar, I.; Vishvanathan, N.N.; Dey, G.K.; Tewari, R.; Srivastava, D.; Banerjee, S.

    2002-01-01

    In the present study, two starting materials used were as-cast Zircaloy-2 with random texture and the finished tube with relatively stronger starting texture. Specimens of the alloys were hot rolled to various strains at different temperature. The texture measurement was carried out and was represented in the form of Orientation Distribution Function which showed a sluggish texture development on high temperature deformation. In the case of as cast alloy with increase in strain at a constant deformation temperature, development in the texture was significant. Upon increasing the working temperature, rate of the overall texture development has been found to reduce. This could be due to reduced slip-twin activities, recovery or due to recrystallization. Microstructural and relative hardening studies were carried out for understanding the mechanisms of deformation texture developments at warm and hot working stages. In the case of finished tube having initially strong texture exhibited slower development in texture on warm and hot rolling. (author)

  16. Creep behavior of Zircaloy cladding under variable conditions

    International Nuclear Information System (INIS)

    Matsuo, Y.

    1989-01-01

    Various creep tests of Zircaloy cladding tubes under variable conditions were conducted to investigate which hardening rule can be applicable for the creep behavior associated with condition changes. The results show that the strain-hardening rule is applicable in general when either the stress or temperature conditions change, provided that a certain amount of creep strain recovery is observed in case of stress drop. In stress reversal conditions, however, softening of the material was observed. Strain rate after stress reversal is much higher than that predicted by the strain-hardening rule. In this case, the modified strain-hardening model, considering a recoverable creep-hardening range together with the strain recovery, predicts the creep behavior well. The applicability of the model is ascertained through a verification test that includes stress reversal, strain recovery, stress changes, and temperature changes

  17. High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Suzuki, Motoye

    1982-03-01

    In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

  18. Out-of-pile fatigue tests on Zircaloy CANDU sheaths

    International Nuclear Information System (INIS)

    Roth, Maria; Ciocanescu, Marin; Gheorghiu, Constantin; Pitigoi, Vasile; Ducu, Catalin; Malinovschi, Viorel

    2005-01-01

    The paper outlines the achievements in the nuclear research field of cooperation on Nuclear Fuel performed as part of the collaboration under the Memorandum of Understanding, settled between Atomic Energy of Canada Limited (AECL) and Institute for Nuclear Research (ICN), The sheath behavior was simulated using out-of-pile fatigue tests, in conditions identical with those met during the operation in power cycling of CANDU reactor, except for irradiation. A special test rig, designed and carried-out at ICN ensured the experimental requirements according to the Canadian testing procedure. The description of the experimental setup and monitoring of testing parameters were also done. The fatigue life time, expressed as number of cycles to rupture (N), was measured as a function of the total strain amplitude (e) induced in the Zircaloy-4 sheath samples. Strain-Life time fatigue dependence (e-N) under low cycle fatigue conditions was also verified using the Coffin-Manson correlation. (authors)

  19. Pressure effects on high temperature steam oxidation of Zircaloy-4

    International Nuclear Information System (INIS)

    Park, Kwangheon; Kim, Kwangpyo; Ryu, Taegeun

    2000-01-01

    The pressure effects on Zircaloy-4 (Zry-4) cladding in high temperature steam have been analyzed. A double layer autoclave was made for the high pressure, high temperature oxidation tests. The experimental test temperature range was 700 - 900 deg C, and pressures were 0.1 - 15 MPa. Steam partial pressure turns out to be an important one rather than total pressure. Steam pressure enhances the oxidation rate of Zry-4 exponentially. The enhancement depends on the temperature, and the maximum exists between 750 - 800 deg C. Pre-existing oxide layer decreases the enhancement about 40 - 60%. The acceleration of oxidation rate by high pressure team seems to be originated from the formation of cracks by abrupt transformation of tetragonal phase in oxide, where the un-stability of tetragonal phase comes from the reduction of surface energy by steam. (author)

  20. Reaction diffusion in chromium-zircaloy-2 system

    International Nuclear Information System (INIS)

    Xiang Wenxin; Ying Shihao

    2001-01-01

    Reaction diffusion in the chromium-zircaloy-2 diffusion couples is investigated in the temperature range of 1023 - 1123 K. Scanning electron microscope (SEM) and energy dispersive spectrum (EDS) were used to measure the thickness of the reaction layer and to determine the Zr, Fe and Cr concentration penetrate profile in reaction layer, respectively. The growth kinetics of reaction layer has been studied and the results show that the growth of intermetallic compound is controlled by the process of volume diffusion as the layer growth approximately obeys the parabolic law. Interdiffusion coefficients were calculated using Boltzmann-Matano-Heumann model. Calculated interdiffusion coefficients were compared with those obtained on the condition that Cr dissolves in Zr and merely forms dilute solid solution. The comparison indicates that Cr diffuses in dilute solid solution is five orders of magnitude faster than in Zr(Fe, Cr) 2 intermetallic compound

  1. SIMS and TEM study on oxide characteristics of Zircaloy

    International Nuclear Information System (INIS)

    Jung, Y. H.; Baek, J. H.; Kim, S. J.; Kim, K. H.; Choi, B. K.; Jung, Y. H.

    1998-01-01

    Long-term corrosion test, SIMS analysis, and TEM study were carried out to investigate the corrosion characteristics and corrosion mechanism of Zircaloy-4 in LiOH solution. The corrosion tests were performed in alkali solutions at 350 deg C for 500days. SIMS analysis was performed for the specimens prepared to have an equal oxide thickness to measure the cation content. TEM studies on the samples formed in various alkali solutions were also conducted. Based on the corrosion test, SIMS analysis, and TEM study, the cation is considered to control the corrosion in LiOH solution and its effect is dependent on the concentration of alkali and the oxide thickness. The slight acceleration of corrosion rate at a low concentration is thought to be caused by the cation incorporation into oxide while the significant acceleration at a high concentration is due to the transformation of oxide microstructure that would be induced by the cation incorporation

  2. Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes

    Directory of Open Access Journals (Sweden)

    Yang-Il Jung

    2018-03-01

    Full Text Available An oxide-dispersion-strengthened (ODS layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

  3. Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy

    International Nuclear Information System (INIS)

    Pereira, Luiz Alberto Tavares

    2014-01-01

    PWR reactors employ, as nuclear fuel, UO 2 pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

  4. Interaction between zircaloy tube and inconel spacer grid at high temperature

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi; Furuta, Teruo

    1990-09-01

    In order to investigate the interaction between fuel cladding and spacer grid of the pressurized water reactor during a severe accident, isothermal reaction tests were performed at the temperature range from 1248 to 1673K. A specimen consisted of a short Zircaloy-4 cladding tube and a piece of spacer grid of Inconel-718. In the tests in an argon atmosphere, eutectic reaction between Zircaloy and Inconel was observed at the contact points at 1248K. Rapid reaction was observed at higher test temperatures. For example, in the test at 1373K for 300s, Zircaloy reacted with Inconel over the entire thickness (0.62mm) of the tube in the vicinity of the contact point. In the present tests, Zircaloy which has higher melting point than Inconel was dissolved preferentially due to eutectic formation. In the tests in an oxygen atmosphere, no eutectic reaction was observed at temperatures below 1437K. A trace of interaction was found at the contact point of specimen heated at 1573 and 1623K. However, decrease in Zircaloy thickness was not measured. The possibility of eutectic reaction between Zircaloy cladding and Inconel spacer grid seems to be quite limited when sufficient oxygen is supplied. (author)

  5. Influence of Zircaloy cladding composition on hydride formation during aqueous hydrogen charging

    Energy Technology Data Exchange (ETDEWEB)

    Rajasekhara, S. [Intel Corporation, 2501 NW 229th Av., Hillsboro, OR 97124 (United States); Kotula, P.G.; Enos, D.G.; Doyle, B.L. [Sandia National Laboratories, Albuquerque, NM, 87185 (United States); Clark, B.G., E-mail: blyclar@sandia.gov [Sandia National Laboratories, Albuquerque, NM, 87185 (United States)

    2017-06-15

    Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. - Highlights: •Aqueous charging was performed to produce a layer of zirconium hydride for three different Zr-alloy claddings. •Hydride thicknesses were analyzed by elastic recoil detection and transmission electron microscopy. •Zircaloy-2 and Zircaloy-4 formed thicker hydride layers than ZIRLO™ for the same charging durations.

  6. Thermal gradient effects on the oxidation of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

    1990-01-01

    A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

  7. Understanding thermally activated plastic deformation behavior of Zircaloy-4

    Science.gov (United States)

    Kumar, N.; Alomari, A.; Murty, K. L.

    2018-06-01

    Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973 K for 4 h) condition was subjected to uniaxial tensile testing at a constant cross-head velocity at temperatures in the range 293 K-1073 K and repeated stress relaxation tests at 293 K, 573 K, and 773 K. The minimum in the total elongation was indicative of dynamic strain aging phenomenon in this alloy in the intermediate temperature regime. The yield stress of the alloy was separated into effective and athermal components and the transition from thermally activated dislocation glide to athermal regime took place at around 673 K with the athermal stress estimated to be 115 MPa. The activation volume was found to be in the range of 40 b3 to 160 b3. The activation volume values and the data analyses using the solid-solution models in literature indicated dislocation-solute interaction to be a potential deformation mechanism in thermally activated regime. The activation energy calculated at 573 K was very close to that found for diffusivity of oxygen in α-Zr that was suggestive of dislocations-oxygen interaction during plastic deformation. This type of information may be helpful in alloy design in selecting different elements to control the deformation behavior of the material and impart desired mechanical properties in those materials for specific applications.

  8. Coating of Zircaloy sheaths with silica glass using the Sol-Gel technique for protection against oxidation

    International Nuclear Information System (INIS)

    De Sanctis, O.; Pellegri, N.; Gomez, L.

    1990-01-01

    With the aim of improving corrosion resistance of Zircaloy, a few Zircaloy sheaths were covered with vitreous silica. Deposition was made by dip coating in tetraetilortosilicate (TEOS) solutions and later densification treatment at 500 degrees C. Oxidation tests were performed and compared with sheaths not covered with silica. As a result, an effective increase in the resistance to dry oxidation was found in sheaths which had been protected. The coating-Zircaloy interface was studied using XPS (scanner). (Author). 6 refs., 3 figs

  9. Hydraulic Modular Dosaging Systems for Machine Drives

    Directory of Open Access Journals (Sweden)

    A. J. Kotlobai

    2005-01-01

    Full Text Available The justified principle of making modular dosaging systems for positive-displacement multimotor hydraulic drives used in running gear and technological equipment of mobile construction, road and agricultural machines makes it possible to synchronize motion of running parts. The examples of the realization of modular dosaging systems and an algorithm of their operation are given in the paper.

  10. [Pharmaceutical advice concerning different pharmaceutical dosage forms].

    Science.gov (United States)

    Szakonyi, Gergely; Zelkó, Romána

    2010-01-01

    The present paper summarizes the commonly applied types of drug uptake and the pharmacists' advice concerning a certain dosage form. The manuscript also deals with the modified release dosage forms and their abbreviations in the name of the marketing authorized products.

  11. Surface analytical investigations of the thermal behaviour of passivated Zircaloy-4 surfaces and of the reaction behaviour of iodine with Zircaloy-4 surfaces

    International Nuclear Information System (INIS)

    Kaufmann, R.

    1988-07-01

    In the first part of the present work the thermal behaviour of atmospherically oxidized Zircaloy-4 samples was investigated at various temperatures. In a next step the amount of iodine adsorbed at the metallic surface was determined as well at room temperature with varying iodine exposures as for constant exposure but varying temperatures. Furthermore, the zirconium iodide species resulting from the interaction of iodine with the Zircaloy-4 and desorbed at higher temperatures were identified by means of residual gas analysis. During these studies it was found that the oxidic overlayer of the passivated Zircaloy-4 samples is decomposed at temperatures above 200 0 C. The iodine uptake at metallic surfaces (cleaned by Ar-ion sputtering) at room temperature slows markedly down after formation of a closed zirconium-iodide overlayer and consequently the further reaction proceeds diffusion-controlled. At 200 0 C ZrI 4 is formed being the thermodynamically most stable Zr-iodide. During desorption experiments using iodine exposed Zircaloy-4 samples the release of ZrI 4 was proved. The results obtained from the various experiments are finally discussed with respect to the iodine-induced stress corrosion cracking process and the underlying basic mechanisms and a transport mechanism for the SCC in nuclear fuel rods is proposed. (orig./RB) [de

  12. Microstructural examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

    2005-01-01

    Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples from a Zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit 1, for a period for 6.77 effective full power years (EFPYs), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

  13. Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hofmann, P.

    1983-01-01

    Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

  14. High temperature interaction between UO2 and Zircaloy-4/silver mixture

    International Nuclear Information System (INIS)

    Uetsuka, Hiroshi; Nagase, Fumihisa; Otomo, Takashi

    1995-12-01

    The reaction between UO 2 and Zircaloy is a main material interaction in the reactor core during a severe accident of LWR. With a view of examining the influence of the core materials having low melting temperatures on the reaction, the effect of silver that is main component of PWR control rod alloy was investigated in the temperature range from 1373 to 1703K. Zircaloy was completely liquefied by the same weight of liquid silver at tested temperatures. The reaction between UO 2 and (Zircaloy+silver) mixture roughly obeyed a parabolic rate law. The determined reaction rate below about 1600K was much lower than that obtained by Hofmann et al. for the reaction between UO 2 and Zircaloy. However, it sharply increased with temperature and became comparable with the rate of UO 2 /Zircaloy reaction at about 1700K. Metallurgical examination including EPMA analysis revealed that Zr(O) layer formed at the reaction interface only for the tests below about 1600K correlated with the discontinuity of the temperature dependence of reaction rate. (author)

  15. Influence of sintering time on distribution of alloying elements composition in Zircaloy pellet

    International Nuclear Information System (INIS)

    Sigit; Muchlis B; Widjaksana; Eric, J.; Suryana, RA; Gunawan

    1996-01-01

    Influence of sintering time on distribution of alloying elements composition in zircaloy pellet has been studied. Zircaloy pellets were obtained by pressing of Zr, Fe, Cr and Sn powders mixture in adequate composition of zircaloy-4, than the green pellets were sintered at 1100 o C for 1 - 3 hours. The alloying elements (Fe, Cr and Sn) composition in zircaloy pellets as sintering product were determined by Scanning Electron Microscope - Energy Dispersive X-Ray Analyser (SEM-EDAX). The experiments showed that there was an accumulation of Sn in a site of the zircaloy green pellet of 17.46 %, but after sintering process, the Sn was distributed everywhere. The influence of sintering time up to 1 hour showed a decreasing Sn composition from 9 % to 2 % which then relatively constant, while for Fe and Cr its decreasing was relatively small, i.e. : 1.86 % to 0.6 % and 1.04 % to 0.17 % respectively. The sintering process revealed no clear grain boundaries and powder homogenization did not complete. Observation on metallographic photos showed that this condition was in initial stage of sintering process where there was a complex phenomenon i.e.: no powder homogenization in green pellet or initial heating rate was extremely quick

  16. Effect of the anodization variables in the corrosion resistence of the zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Figueiredo, M.E.

    1981-02-01

    The anodization effect in the oxidation of the zircaloy-4 in steam atmosphere at 10,06MPa was investigated. It was also studied how the voltage and the types of electrolytes at several values of pH affect the growing of the anodic oxide film and the performance of the zircaloy-4 in relation to corrosion. Anodizations of zircaloy-4 tubes have been made with voltages ranging from zero to 280V and using electrolytic solutions of Na 2 B 4 O 7 , CH 3 COOH and NaOH in the concentrations of 1,0N, 0,1N and 0,01N. After anodization, the tubes were oxidized in autoclave under steam at 400 0 C and 10,06 MPa during 3 and 14 days. The results show that the anodization inhibit the oxidation process of zircaloy-4, and that this protection increases with the voltage applied for film formation. The relationship between the weight gain after oxidation in autoclave and the anodization voltage is of the exponential type: (σM/A) sub(AC) = Ce sup(-DV). The observed relationship between the applied voltage and the weight gain due to anodization is of the linear type: (σM/A) sub(AN) = aV. Concerning the influence of different electrolytes, it was observed a similar behaviour between them with respect to the thickness of the anodic oxide and the weight gain of zircaloy-4 after the autoclave test. (Author) [pt

  17. Formation of Lamellar Structured Oxide Dispersion Strengthening Layers in Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang-Il; Park, Jung-Hwan; Park, Dong-Jun; Kim, Hyun-Gil; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lim, Yoon-Soo [Hanbat National University, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures. According to our previous investigations, the tensile strength of Zircaloy-4 was increased by up to 20% with the formation of a thin dispersed oxide layer with a thickness less than 10% of that of the Zircaloy-4 substrate. However, the tensile elongation of the samples decreased drastically. The brittle fracture was a major concern in development of the ODS Zircaloy-4. In this study, a lamellar structure of ODS layer was formed to increase ductility of the ODS Zircaloy-4. The mechanical properties were varied depending on the structure of ODS layer. For example, the partial formation of ODS layer with the thickness of 10% to the substrate thickness induced the increase in tensile strength up to about 20% than fresh Zircaloy-4.

  18. Implications of Y-fluting microstructures in zircaloy stress-corrosion fracture and analogous systems

    International Nuclear Information System (INIS)

    Banks, T.M.; Garlick, A.

    1982-01-01

    Transgranular cleavage is an important mode of crack propagation during stress-corrosion cracking (SCC) of Zircaloy in iodine vapour; and another characteristic feature is the presence of parallel closely spaced ridges. These are often referred to as Y-flutings because each ridge takes the form of an inverted Y when viewed along the direction of crack growth. The flutings are shown here to be formed by localised ductile parting of the Zircaloy near the tips of cleavage cracks; high mechanical constraints in those regions and the limited number of available slip systems result in the formation of a planar array of parallel tunnels. Upon final separation these appear as a pattern of parallel ridges on each fracture face. Striking similarities in morphology have been noted here between Y-flutings in Zircaloy and those produced during tests on unstable fluid interfaces: the direction of motion of the fluid interface can be determined from the Y-morphology and is in agreement with observations from Zircaloy SCC tests. It is further demonstrated that equations governing thermodynamic and kinetic instability of fluid interfaces can be adapted to relate the fluting spacing in Zircaloy to standard fracture mechanics parameters. (author)

  19. Uniaxial ratcheting behavior of Zircaloy-4 tubes at room temperature

    International Nuclear Information System (INIS)

    Wen, Mingjian; Li, Hua; Yu, Dunji; Chen, Gang; Chen, Xu

    2013-01-01

    In this study, a series of uniaxial tensile, strain cycling and uniaxial ratcheting tests were conducted at room temperature on Zircaloy-4 (Zr-4) tubes used as nuclear fuel cladding in Pressurized Water Reactors (PWRs) for the purpose to investigate the uniaxial ratcheting behavior of Zr-4 and the factors which may influence it. The experimental results show that at room temperature this material features cyclic softening remarkably within the strain range of 1.6%, and former cycling under larger strain amplitude cannot retard cyclic softening of later cycling under lower strain amplitude. Uniaxial ratcheting strain accumulates in the direction of mean stress, and the ratcheting stain level is larger under tensile mean stress than that under compressive mean stress. Uniaxial ratcheting strain level increases with the increase of mean stress and stress amplitude, and decreases with the increase of loading rate. The sequence of loading rate appears to have no effects on the final ratcheting strain accumulation. Loading history has great influence on the uniaxial ratcheting behavior. Lower stress level after loading history with higher stress level leads to the shakedown of ratcheting. Higher loading rate after loading history with lower loading rate brings down the ratcheting strain rate. Uniaxial ratcheting behavior is sensitive to compressive pre-strain, and the decay rate of the ratcheting strain rate is slowed down by pre-compression

  20. Irradiation effect on fatigue behaviour of zircaloy-4 cladding tubes

    International Nuclear Information System (INIS)

    Soniak, A.; Lansiart, S.; Royer, J.; Waeckel, N.

    1993-01-01

    Since nuclear electricity has a predominant share in French generating capacity, PWR's are required to fit grid load following and frequency control operating conditions. Consequently cyclic stresses appear in the fuel element cladding. In order to characterize the possible resulting clad damage, fatigue tests were performed at 350 deg C on unirradiated material or irradiated stress relieved Zircaloy-4 tube portions, using a special device for tube fatigue by repeated pressurization. It appears that, for high stress levels, the material fatigue life is not affected by irradiation. But the endurance fatigue limit undergoes a decrease from the 350 MPa value for unirradiated material to the 210 MPa value for the material irradiated for four cycles in a PWR. However, this effect seems to saturate with irradiation dose: no difference could be detected between the two cycles results and the corresponding four cycles results. The corrosion effect and the load following influence were also investigated: they do not appear to modify the fatigue behaviour in our experimental conditions

  1. Delayed hydride cracking behavior for zircaloy-2 plate

    International Nuclear Information System (INIS)

    Mills, J.W.; Huang, F.H.

    1991-01-01

    The delayed hydride cracking (DHC) behaviour for Zircaloy-2 plate was characterized at temperatures ranging from 300 to 550 o F. Specimens with a longitudinal (T-L) orientation exhibited a classic two-stage DHC response. At K values slightly above the threshold level (K th ), crack-growth rates increased dramatically with increasing K values (stage I). The K th value was found to be 11 and 14 ksi√ in at 400 and 500 o F. At high K values (stage II), cracking rates were relatively insensitive to applied K levels. Stage II crack growth was a thermally activated process described by an Arrhenius-type relationship with an activation energy of 65 kJ/mol. This energy level agreed with the theoretical activation energy for hydrogen diffusion into the triaxial stress field ahead of a crack. Above a critical temperature (300 o F), an overtemperature cycle was required to initiate DHC. The magnitude of the thermal excursion required to initiate cracking was found to increase at higher test temperatures. Specimens with a transverse(L-T) orientation showed a very low sensitivity to DHC because of an unfavorable crystallographic orientation for hydride reorientation. Metallographic and fractographic examinations were performed to understand the DHC mechanism. (author)

  2. Propagation of stress-corrosion cracks in unirradiated zircaloy

    International Nuclear Information System (INIS)

    Norring, K.; Haag, Y.; Wikstroem, C.

    1982-01-01

    Propagation of iodine-induced stress-corrosion cracks in Zircaloy was studied using pre-cracked and internally pressurized cladding tubes. These were recrystallized at different temperatures, to obtain grain sizes between 4 μm and 10 μm. No statistically significant difference in propagation rate due to the difference in grain size was observed. If the obtained data, with Ksub(I) values ranging from 4 to 11 MNmsup(-3/2), were log-log plotted (da/dt = CKsub(I)sup(N)), as usual, they fell within the scatter-band of data reported earlier. But from this plot it could also be seen that the Ksub(I) interval can be divided into two separate parts having different da/dt-Ksub(I) relations. The transition takes place at a Ksub(I) value of about 8 MNmsup(-3/2). The region with lower Ksub(I) values shows a substantially lower n value than the upper region (2.4 and 9.8 respectively), and earlier reported values (n = 7 to 10). This transition is in good agreement with a transition from an intergranular to a transgranular propagation mode of the stress-corrosion crack. (orig.)

  3. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10 21 n cm -2 to 5.9 x 10 21 n cm -2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

  4. Treatment of stainless steels and zircaloy cladding hulls

    International Nuclear Information System (INIS)

    Jenkins, I.L.; Taylor, R.F.

    1978-01-01

    Results are reported on the fissile material content and the distribution of alpha and beta-gamma emitters in both types of cladding. Apart from very small amounts of residual fuel, fissile material is present as a deposit formed during the dissolution of fuel and also as material driven into the cladding by fission recoil. Alpha-emitters penetrate to depths of 1-2 μm into both S.S. and Zircaloy claddings. The surface deposits on individual hulls can be effectively removed by refluxing with nitric acid or by cleaning with nitric acid in an ultrasonic bath. The physical structural and handling behavior of hull assemblies are examined as being of key importance to the establishment of an efficient cleaning process. The reference leaching target is to extract residual fuel fragments and to remove surface deposits. Preferred routes for compaction, drumming, and encapsulation are briefly reviewed with regard to achieving a final package volume half that of the original hulls with associated hardware

  5. Crack resistance curve determination of zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Bertsch, J.; Alam, A.; Zubler, R.

    2009-03-01

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 o C and 350 o C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could be

  6. Ratcheting fatigue behavior of Zircaloy-2 at room temperature

    Energy Technology Data Exchange (ETDEWEB)

    Rajpurohit, R.S., E-mail: rsrajpurohit.rs.met13@iitbhu.ac.in [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India); Sudhakar Rao, G. [Nuclear Energy and Safety Department, Paul Scherrer Institute, Villigen, CH-5232 (Switzerland); Chattopadhyay, K.; Santhi Srinivas, N.C.; Singh, Vakil [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India)

    2016-08-15

    Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain. - Highlights: • Ratcheting strain accumulation occurred due to asymmetric cyclic loading. • Accumulation of ratcheting strain increased with mean stress and stress amplitude. • Ratcheting strain accumulation decreased with increase in stress rate. • With increase in mean stress and stress amplitude there was reduction in fatigue life. • Fatigue life is improved with increase in stress rate.

  7. Crack resistance curve determination of zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J.; Alam, A.; Zubler, R

    2009-03-15

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 {sup o}C and 350 {sup o}C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could

  8. Threshold values characterizing iodine-induced SCC of zircaloys

    International Nuclear Information System (INIS)

    Une, K.

    1981-01-01

    In this paper, threshold values of stress, stress intensity factor, strain, strain rate and iodine concentration for SCC of unirradiated and irradiated Zircaloys are reviewed. The ratio of σ sub(th)/σ sub(y) adequately represents the effects of cold-work and irradiation on the SCC susceptibility, where threshold stress σ sub(th) is defined as the minimum stress to cause SCC to failure after 10-20 hours and σ sub(y), the yield stress obtained in an inert atmosphere. The ratio becomes gradually smaller with larger σ sub(y) and is less than 1 for materials with yield strengths above about 350MPa. Plastic strain appears to be necessary for SCC; plastic strains to failure range from 0.1 to 1% for high strength materials, even when data for irradiated materials are included. Strain rate significantly affects the susceptibility. A comparison of SCC data between constant strain rate and constant stress tests is presented. (author)

  9. Zircaloy spacer grid for boiling light water reactors

    International Nuclear Information System (INIS)

    Borgiani, F.; Cali', G.P.; Cerretti, P.; Pazzo, P.

    1975-01-01

    The need to increase the neutronic efficiency of the new cores of BWR's, lead to study types of spacer-grids made of low neutronic absorption materials as zircaloy-4. The particular mechanical behaviour of this material suggested to design a spacer-grids such as to utilize only blanking, slotting and bending operations as plastic forming and to avoid therefore drawing effects. The optimization of the bending procedures lead to a final spacer-grids configuration equally stiff in all directions and planes. Only for the ''elastic constraints'' nichel alloy sheets were used to made easy the whole spacer design. The ''rigid constraints'', supporting the rods, have been obtained directly from the spacer structure. Calculations were performed to verify the mechanical strength of the main grid components. In this framework a computer code was developed to find the best elastic characteristic of the ''elastic constraints'' taking into account the machining tolerances. Some original methods to test the integral behaviour of the grid assembled as well as the procedures to be adopted for its best maintenance, are described

  10. Characterization of Zircaloy-4 oxide layers by impedance spectroscopy

    International Nuclear Information System (INIS)

    Barberis, P.

    1999-01-01

    Two Zircaloy-4 type alloys with different tin contents (0.5 and 1.2 wt%) have been oxidized in autoclave (400 C in steam) for several durations (1-140 days). The film has been characterized by electrochemical impedance spectroscopy (EIS). Several soaking times have been investigated (up to 40 days). The Cole-Cole representation has been used to display and study the data. A simple electrical model has been derived from the observed spectra: the electrical circuit includes two RC loops in series, whose capacitances are frequency dispersed. It is thoroughly related to the layer structure. It has been shown that even before the kinetic transition, the film is constituted of three parts: an inner layer which is compact, an outer layer subdivided in an external region immediately soaked by the electrolyte, and an internal one in which electrolyte diffusion processes can take place. The kinetic transition is interpreted in terms of an abrupt 'compacity' change, both layers degrading at this point. The alloy with high tin content exhibits higher dispersive properties of the oxide layer formed on it, in correlation with its faster oxidation kinetics. (orig.)

  11. Stress corrosion cracking behavior of zircaloy-2 in iodine environment

    International Nuclear Information System (INIS)

    Ikeda, Seiichi

    1983-01-01

    The effects of strain rates, iodine partial pressure and testing temperature on SCC behavior of zircaloy-2 in iodine environment were studied by means of slow strain rate technique (SSRT). SCC behavior of recrystallized specimens in iodine environment was remarkably influenced by the testing temperatures, and the susceptibility to SCC of specimens tested at 623 K was higher than that at 573 K. The susceptibility to SCC of recrystallized specimens increased with increasing iodine partial pressure at the lower strain rates of 4.2 x 10 -6 s -1 and 8.3 x 10 -7 s -1 . Cold worked specimens indicate no SCC failure in iodine environment regardless of strain rates, although those were tested only at 573 K. Fractographic observation revealed that SCC features of recrystallized specimens can be classified into two groups. One group, mostly specimens tested at 573 K, are characterized by the fact that cracks are initiated from corrosion pits. The other group are characterized by transgranuler SCC in the absence of pitting. This type of crack is found on specimens tested in environments containing more than 570 Pa iodine and seems to be produced by iodine embrittlement. (author)

  12. Crystal plasticity modeling of irradiation growth in Zircaloy-2

    Science.gov (United States)

    Patra, Anirban; Tomé, Carlos N.; Golubov, Stanislav I.

    2017-08-01

    A physically based reaction-diffusion model is implemented in the visco-plastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. The reaction-diffusion model accounts for the defects produced by the cascade of displaced atoms, their diffusion to lattice sinks and the contribution to crystallographic strain at the level of single crystals. The VPSC framework accounts for intergranular interactions and irradiation creep, and calculates the strain in the polycrystalline ensemble. A novel scheme is proposed to model the simultaneous evolution of both, number density and radius, of irradiation-induced dislocation loops directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behaviour of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture and external stress on the coupled irradiation growth and creep behaviour are also studied and compared with available experimental data.

  13. Delayed hydride cracking of Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

    2007-01-01

    Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

  14. Analysis of zircaloy oxide thickness data from PWRs

    International Nuclear Information System (INIS)

    Sheppard, K.D.; Speyer, D.M.; Chan, Y.Y.; Frankl, I.; Strasser, A.A.

    1990-02-01

    Prior EPRI funded research (Project 1250-1) resulted in a set of Zircaloy waterside corrosion models. These models were based principally on KWU reactor data. The objective of this study was to evaluate the ability of the KWU corrosion models to predict available domestic USA data for all domestic PWR vendors in order to further validate the models and to provide a consistent basis to judge the corrosion data of the domestic plants. A methodology for analyzing the large amount of data was developed and implemented in a single channel model. This model includes the capability, by a method described herein, of accounting for open core related effects (crossflow) and the effect of the immediately adjacent fuel rods, guide tubes, etc., on the coolant conditions around the fuel rods that were measured for oxide thickness. Data from the Arkansas Unit number-sign 2 (ANO-2) Combustion Engineering (C-E), Oconee Units 1 and 2 built by Babcock ampersand Wilcox (B ampersand W), and the Trojan reactor built by Westinghouse (W) were used in this study. The corrosion models previously developed, and the present single channel model methodology, were able to predict the corrosion data quite well. The maximum corrosion thickness was on the order of 20 to 40 microns in all plants studied. 13 refs., 11 figs., 5 tabs

  15. Fiche technique du spermogramme et du spermocytogramme ...

    African Journals Online (AJOL)

    En Afrique la stérilité du couple constitue un drame social. Selon l'OMS, environ 8 à 12 % des couples africains sont touchés par une infertilité. La responsabilité masculine dans la stérilité est comprise entre 30 à 40%. Les causes de l'infertilité masculine peuvent être l'impuissance et/ ou l'altération du sperme. L'étude de ...

  16. The Use of Direct Tritium Assay Techniques in Studies with Tritiated Thymidine; Application des Methodes de Dosage Direct du Tritium aux Etudes avec la Thymidine Tritiee; 041f 0440 0438 043c 0435 043d 0435 0414 ; Aplicacion de Tecnicas de Medicion Directa del Tritio a Estudios con Timidina Tritiada

    Energy Technology Data Exchange (ETDEWEB)

    Gordon Steel, G. [Physics Department, Institute of Cancer Research, Royal Cancer Hospital, Clifton Avenue, Belmont, Surrey (United Kingdom)

    1962-02-15

    Results are described of investigations into the catabolism of tritiated thymidine in the rat, during the period of its initial localization in tissues, and into the retention of tritium-labelled cells in intestine and bone marrow up to 16 d after labelling. During the first hour after the injection of tritiated thymidine, whilst the concentration of nonvolatile tritium in most tissues rose to a saturation level, the concentration in liver fell rapidly from an initially high level. Using a device which enabled a sample of tissue water to be extracted from a cold tissue specimen, measurements were made of the specific activity of tritiated water from various organs. By one hour after injection, the tritium concentration was almost constant throughout the body water, but during the period of establishment of this equilibrium, considerable gradients of tissue water specific activity could be detected. In liver, the gradient indicated a flow of tritiated water from the liver into the blood; in spleen, testis and muscle the flow was from the blood into the tissue. Measurements of tritium retention in intestine and bone marrow after giving tritiated thymidine, indicated that in both tissues there was an initial plateau and that the subsequent decay had two prominent exponentials. In bone marrow the observation of a plateau conflicts with other published work; the plateau is not observed in animals which have received protracted irradiation. (author) [French] L'auteur decrit les resultats de recherches sur le catabolisme de la thymidine tritiee dans le rat, au cours de la periode de fixation initiale dans les tissus, et sur la retention de cellules marquees au tritium par l'intestin et la moelle osseuse dans un delai allant jusqu'a 16 jours apres le marquage. Pendant la premiere heure suivant l'injection de thymidine tritiee, alors que la concentration du tritium non volatil dans la plupart des tissus atteignait rapidement un degre de saturation, la concentration dans le

  17. Advances in solid dosage form manufacturing technology.

    Science.gov (United States)

    Andrews, Gavin P

    2007-12-15

    Currently, the pharmaceutical and healthcare industries are moving through a period of unparalleled change. Major multinational pharmaceutical companies are restructuring, consolidating, merging and more importantly critically assessing their competitiveness to ensure constant growth in an ever-more demanding market where the cost of developing novel products is continuously increasing. The pharmaceutical manufacturing processes currently in existence for the production of solid oral dosage forms are associated with significant disadvantages and in many instances provide many processing problems. Therefore, it is well accepted that there is an increasing need for alternative processes to dramatically improve powder processing, and more importantly to ensure that acceptable, reproducible solid dosage forms can be manufactured. Consequently, pharmaceutical companies are beginning to invest in innovative processes capable of producing solid dosage forms that better meet the needs of the patient while providing efficient manufacturing operations. This article discusses two emerging solid dosage form manufacturing technologies, namely hot-melt extrusion and fluidized hot-melt granulation.

  18. The hydrogen generated as a gas and storage in Zircaloy during water quenching

    International Nuclear Information System (INIS)

    Garcia, Eduardo A.

    1999-01-01

    A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during water quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 , 1400 and 1600 C degrees using as-received Zircaloy-4 (no pre oxidation) and with Zircaloy specimens pre oxidised to give oxide thicknesses of 100μm and 300μm. The results are relevant to accident management in light water reactors. (author)

  19. High temperature interaction between Zircaloy-4 and stainless steel type 304

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi

    2001-03-01

    The chemical interactions between Zircaloy-4 and stainless steel type 304 were investigated in the temperature range from 1273 to 1573 K to obtain the basic information on the melt progress in the fuel bundle during an LWR severe accident. Reaction layers were formed at the contact interface and grew as the temperature and the time increase. The Zircaloy was preferentially dissolved by the reaction. The SEM/EDX analyses showed that the main process of the reaction was diffusion of Fe, Cr and Ni into the Zircaloy which resulted in the formation of a Zr-rich eutectic through the tested temperature range. Reaction rates for decrease in the materials thickness were evaluated and the reaction generally obeyed a parabolic rate law. The reaction rate constant was determined at every examined temperature and Arrhenius type rate equations were estimated for the temperature range. (author)

  20. Investigation of microstructure and mechanical properties of proton irradiated Zircaloy 2

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Kumar, Ajay [Nuclear Physics Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Mukherjee, S.; Sharma, S.K.; Dutta, D.; Pujari, P.K. [Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Agarwal, A.; Gupta, S.K.; Singh, P. [Ion Accelerator Development Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India)

    2016-10-15

    Samples of Zircaloy 2 have been irradiated with 4 MeV protons to two different doses. Microstructures of the unirradiated and irradiated samples have been characterized by Electron Back Scatter Diffraction (EBSD), X-ray diffraction line profile analysis (XRDLPA), Positron Annihilation Lifetime Spectroscopy (PALS) and Coincident Doppler Broadening (CDB) Spectroscopy. Tensile tests and micro hardness measurements have been carried out at room temperature to assess the changes in mechanical properties of Zircaloy 2 due to proton irradiation. The correlation of dislocation density, grain size and yield stress of the irradiated samples indicated that an increase in dislocation density due to irradiation is responsible for the change in mechanical behavior of irradiated Zircaloy.

  1. Electrochemical Studies on Important Elements for Zirconium Recovery Form Irradiated Zircaloy-4 Cladding

    International Nuclear Information System (INIS)

    Park, J.; Sohn, S.; Hwang, I.S.

    2015-01-01

    Since Zircaloy cladding accounts for about 16 wt. % of used nuclear fuel assembly, decontamination process is required to reduce the final waste volume from spent nuclear fuel. To develop Zircaloy-4 electrorefining process as an irradiated Zircaloy cladding decontamination process, electrochemical studies on Sn, Cr, Fe and Co which are major or important elements in the irradiated cladding were conducted based on cyclic voltammetry in LiCl-KCl at 500 deg. C. Cyclic voltammetry for Sn, Fe, Cr and Co elements that should be eliminated was conducted and revealed that redox reactions of these ions are much simpler than Zr and more reductive than Zr. The reliability of cyclic voltammetry was verified by comparing diffusion coefficients and formal reduction potentials of these ions obtained in this study to previous studies. (authors)

  2. The hydrogen generated as a gas and storage in Zircaloy during steam quenching

    International Nuclear Information System (INIS)

    Garcia, Eduardo A.

    2000-01-01

    A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during steam quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 centigrade, 1400 centigrade and 1600 centigrade using as-received Zircaloy-4 (no pre-oxidation) and with Zircaloy specimens pre-oxidized to give oxide thickness of 100μm and 300μm. The results are relevant to accident management in nuclear power plants. (author)

  3. Les Cahiers du CREAD

    African Journals Online (AJOL)

    Admin

    politique de bas prix exercée par la Russie et le Qatar vient confirmer ce constat ; s'ajoute à cela l'entrée éventuelle du gaz non conven- tionnel, dont son prix actuel de 3/4 $US, offre aux USA l'opportunité d'être exportateur de ..... les compagnies à produire en matière du gaz naturel, tels le prix du gaz naturel, le prix des ...

  4. Bulletin du CRDI #124

    International Development Research Centre (IDRC) Digital Library (Canada)

    Les femmes jouent un rôle important dans les exploitations minières artisanales et à petite échelle en Afrique subsaharienne. De concert ... Couverture du livre: Une vie saine pour les femmes et les enfants vulnérables · Couverture du livre: Entre el activismo y la intervención · Couverture du livre: Revitalizing Health for All.

  5. Bulletin du CRDI #125

    International Development Research Centre (IDRC) Digital Library (Canada)

    L'IOSRS remporte le prix de la diplomatie scientifique · GrowInclusive : la plateforme tant attendue est en construction · Toutes les nouvelles. Activités à venir. Semaine du développement international 2018. Le CRDI célébrera la Semaine du développement international du 4 au 10 février 2018. Suivez-nous sur Twitter et ...

  6. Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

    International Nuclear Information System (INIS)

    Holub, F.

    1987-08-01

    The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

  7. Interaction between aluminium oxide pellets and Zircaloy tubes in steam atmospheres at temperatures above 12000C

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1988-09-01

    The burnable poison rods in light water reactors (LWR) consist of Al 2 O 3 /B 4 C pellets surrounded by Zircaloy-4 cladding tubes. In the Al 2 O 3 /B 4 C pellets of a LWR rod alumina is the main constituent (98.6 wt.-%) whereas boron carbide acts as neutron absorber. Failure of the Al 2 O 3 /Zircaloy test rods started at 1350 0 C when first droplets of molten material were observed running down the test bundle forming bundle blockages upon solidification. Post test examinations revealed that the process of liquefaction was initiated by a reduction of alumina by Zircaloy resulting in a (Zr, Al, O) melt which decomposed on cooldown into two metallic phases, a (Zr, Al) alloy and oxygen-stabilized a-Zr(O). The components of an extremely porous ceramic melt were also Zr, Al, and oxygen but with a higher oxygen content compared to the metallic melt. The ceramic melt decomposes on cooldown into an Al 2 O 3 /ZrO 2 eutectic with various amounts of primary constituents. Other types of relocated material were due to melting of essentially unreacted Zircaloy cladding and to debris formation by fracturing of oxidized cladding and Al 2 O 3 pellets stack residues. The interactions between Al 2 O 3 and Zircaloy occurring in a burnable poison rod are furthermore important for the behavior of the entire LWR core because the generated metals are able to attack the UO 2 chemically and dissolve or liquefy the fuel even below the melting point of Zircaloy (1760 0 C). As a result, fuel elements which contain burnable poison rods are expected to fail under severe accident conditions at about 1500 0 C. (orig./HP) [de

  8. Thermal diffusion of hydrogen in zircaloy-2 containing hydrogen beyond terminal solid solubility

    International Nuclear Information System (INIS)

    Maki, Hideo; Sato, Masao.

    1975-01-01

    The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy. In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20 0 --30 0 C was applied between the inside and outside surfaces of the specimen in a thermal simulator. To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300 0 C are Dsub(p)=2.3x10 5 exp(-32,000/RT), (where R:gas constant, T:temperature) and the apparent heat of transport Qsub(p) =-60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours. (auth.)

  9. Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

    International Nuclear Information System (INIS)

    Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

    1990-01-01

    Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

  10. Temperature estimates from the zircaloy oxidation kinetics in the α plus β phase region

    International Nuclear Information System (INIS)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near thermocouple locations

  11. Substructure evolution of Zircaloy-4 during creep and implications for the Modified Jogged-Screw model

    Energy Technology Data Exchange (ETDEWEB)

    Morrow, B.M., E-mail: morrow@lanl.gov [The Ohio State University, 2041 College Rd., 477 Watts Hall, Columbus, OH 43210 (United States); Los Alamos National Laboratory, P.O. Box 1663, MS G755, Los Alamos, NM 87545 (United States); Kozar, R.W.; Anderson, K.R. [Bettis Laboratory, Bechtel Marine Propulsion Corp., West Mifflin, PA 15122 (United States); Mills, M.J., E-mail: millsmj@mse.osu.edu [The Ohio State University, 2041 College Rd., 477 Watts Hall, Columbus, OH 43210 (United States)

    2016-05-17

    Several specimens of Zircaloy-4 were creep tested at a single stress-temperature condition, and interrupted at different accumulated strain levels. Substructural observations were performed using bright field scanning transmission electron microscopy (BF STEM). The dislocation substructure was characterized to ascertain how creep strain evolution impacts the Modified Jogged-Screw (MJS) model, which has previously been utilized to predict steady-state strain rates in Zircaloy-4. Special attention was paid to the evolution of individual model parameters with increasing strain. Results of model parameter measurements are reported and discussed, along with possible extensions to the MJS model.

  12. The anisotropic creep behaviour of zircaloy-4 fuel cladding at 1073 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Bowden, J.; Shewfelt, R.S.W.

    1982-04-01

    The anisotropy coefficients (F, G and H) of Hill's equation, suitably modified for creep deformation, have been determined for Zircaloy-4 fuel cladding from steady-state creep tests at an elevated temperature. Creep specimens were subjected to both uniaxial and biaxial loads (via internal pressure) at 1073 K and the strain measured concurrently in the axial and tangential directions. It has been found that Zircaloy-4 fuel cladding is almost, but not completely, isotropic at 1073 K; the values of F, G and H are 0.57, 0.48 and 0.45 respectively

  13. Influence of neutron irradiation on the stability of recipitates in zircaloy: a critical review

    International Nuclear Information System (INIS)

    Lobo, Raquel M.; Andrade, Arnaldo H. P.

    2013-01-01

    The realization of RMB enterprise (Brazilian Multipurpose Reactor) will give the country a powerful tool to investigate the behavior materials subjected to irradiation. Among them, zirconium alloys, used as cladding of nuclear fuel in reactors type LWR. It is know that neutron irradiation can affect the stability of precipitates in zircaloys, generating as a result changes in theirs mechanical properties, important application of this alloys. This paper present a critical review of neutron irradiation effects on microstructural stability of zircaloys (2 and 4). (author)

  14. In-situ neutron diffraction study of Zircaloy 4 subjected to biaxial tension

    Energy Technology Data Exchange (ETDEWEB)

    Gharghouri, M.A. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, ON (Canada); McDonald, D.; Xiao, L. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Zircaloy-4 is widely used as fuel element cladding in nuclear reactors. Pellet-clad interaction (PCI) failure is a concern for many water reactor fuel designs. Extensive work on the mechanism of PCI failure has led to the conclusion that stress corrosion cracking (SCC) induced by iodine vapour in the temperature range relevant to fuel operation is the most probable cause of PCI failure in zirconium alloy fuel element cladding. In-situ neutron diffraction measurements performed on tubular Zircaloy-4 specimens simultaneously pulled in tension and pressurized internally will provide information on the effects of stress biaxiality on the distribution of stresses at the crystal level during loading. (author)

  15. Tensile properties in zircaloy-II after 590 MeV proton irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Victoria, M. [Ecole Polytechnique Federale, Lausanne (Switzerland)

    1997-09-01

    In order to investigate radiation potential damage effects on the SINQ Zircaloy-rod target, four Zircaloy-II tensile specimens were irradiated at the PIREX facility in 1995 to a proton fluence about 3x10{sup 20} p/cm{sup 2}, which produced a radiation damage of about 1.35 displacements per atom (dpa). Tensile test results show that, although there is some reduction in tensile elongation, substantial ductility still exists after such irradiation dose which corresponds to the peak value obtained in the SINQ target for 23 days operation at 1 mA. (author) 1 fig., 2 refs.

  16. The effect of repeated melting of zircaloy-4 to the distribution of volatile constituents

    International Nuclear Information System (INIS)

    Johneri, E.; Wijaksana; Badruzzaman, M.

    1996-01-01

    The effect of repeated fusion on the composition and distribution of zircaloy volatile elemental constituents (especially Sn) has been investigated. The results showed that the higher the number of repeated fusion is, the more evenly distributed the constituents are, but the composition decreased until reached constant values. This phenomenon occurred due to the relatively faster diffusion movement of one element compared to the others. Further investigation needs to be done to find other proofs of the phenomenon. Moreover, continued research is in demand in order to answer technological problems regarding the zircaloy production and metal alloy production in general. (author)

  17. Plastic behaviour of Zircaloy-4 in the temperature range 77-1000 K

    International Nuclear Information System (INIS)

    Derep, J.L.; Ibrahim, S.; Rouby, D.; Fantozzi, G.; Gobin, P.

    1979-01-01

    Tensile tests were carried out on Zircaloy-4 over a temperature range 77-1000 K. So, we have determined the flow stress variations as a function of temperature and strain rate. Two thermally activated zones were observed between about 77 and 600 K, a plateau stress between 600 and 700 K and an other thermally activated zone above 700 K. The various mechanisms which can be responsible for the thermally activated and athermal zones are discussed in the light of experimental results. The mechanical behaviour of Zircaloy-4 appears similar to the zirconium-oxygen alloys one. (orig.) [de

  18. Observations on deformation systems in zircaloy-2 deformed at room temperature

    International Nuclear Information System (INIS)

    Pettersson, K.; Bergqvist, H.

    1975-08-01

    Different polycrystalline samples of Zircaloy-2 with textures such that the c-axis of most of the grains are oriented near the sheet normal were subjected to loading conditions such that sheet thinning was accomplished. Metallography showed that no twinning was involved. Electron microscopy showed the presence of dislocations which were usually confined to deformation bands. With the help of stereo micrographs the most likely plane of slip was determined to be (1011). The possibility of slip as a means of breaking the oxide film in iodine induced stress corrosion cracking of Zircaloy-2 is briefly discussed. (author)

  19. Identification and analcime quantification; Identification et dosage de l'analcime

    Energy Technology Data Exchange (ETDEWEB)

    Chantret, F; Guillemaut, A; Pouget, R

    1962-07-01

    The authors are comparing thermal analysis and X-ray diffraction methods for the estimation of analcime in rocks. From application to the analcimolithes of Agades - Republic of Niger - it appears that X-ray diffractometry is better convenient, both for identification and estimation; nevertheless, thermal analysis combined with chemical analysis allows to detect variations in the composition of analcime inside a given series. [French] Les auteurs comparent les techniques d'analyse thermique et de diffraction X pour le dosage de l'analcime dans les roches. L'application aux analcimolites d'Agades - Republique du Niger - montre que la diffractometrie X est mieux adaptee a la fois dans l'identification et le dosage; neanmoins, l'analyse thermique, associee a l'analyse chimique, permet de suivre les fluctuations de composition de l'analcime a l'interieur d'une serie determinee. (auteurs)

  20. Fluorometric determination of uranium in urine; Dosage fluorimetrique de l'uranium urinaire

    Energy Technology Data Exchange (ETDEWEB)

    Ronteix, C; Hugot, G [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    For the medical supervision of personnel the most sensitive analytical methods must be used. The fluorimetric method, enabling determinations to be made without previous concentration, is undoubtedly the quickest and most effective. This report is intended to help users of the method to avoid loss of time in the practical application. It therefore describes in great detail the material used, and also gives precise technical information acquired by experience. (author) [French] Pour la surveillance medicale du personnel, il est necessaire d'utiliser les methodes de dosages les plus sensibles. La methode fluorimetrique qui permet d'effectuer le dosage sans concentration prealable est, sans conteste, la plus rapide et la plus efficace. Le present rapport a pour but de permettre aux utilisateurs de cette methode, d'eviter des pertes de temps dans l'application pratique. Il contient donc, tres detaille, le materiel utilise ainsi que des precisions techniques acquises par l'experience. (auteur)

  1. The electroreduction of pentavalent protactinium; Reduction electrolytique du protactinium pentavalent

    Energy Technology Data Exchange (ETDEWEB)

    Musikas, C [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-05-01

    The reduction of pentavalent protactinium to tetravalent protactinium, in sulfuric and hydrochloric media, on a milligram scale was demonstrated by electrolysis in a separate-compartment cell. There was no indication that protactinium may exist at the trivalent state in these solutions. Polarograms in fluoride solutions showed only one reduction wave. The principle of a volumetric method for the titration of protactinium is given. (author) [French] La reduction du protactinium pentavalent en protactinium tetravalent, dans des solutions sulfuriques et chlorhydriques a ete realisee a l'echelle du milligramme, par electrolyse, dans une cellule a compartiments separes. Aucun indice ne permet de penser que le protactinium puisse exister dans ces solutions a l'etat trivalent. De plus les polarogrammes traces en milieu fluorhydrique ne font apparaitre qu'une seule vague de reduction. Le principe d'une methode volumetrique de dosage du protactinium est donne. (auteur)

  2. Microstructure evolution of recrystallized Zircaloy-4 under charged particles irradiation

    Science.gov (United States)

    Gaumé, M.; Onimus, F.; Dupuy, L.; Tissot, O.; Bachelet, C.; Mompiou, F.

    2017-11-01

    Recrystallized zirconium alloys are used as nuclear fuel cladding tubes of Pressurized Water Reactors. During operation, these alloys are submitted to fast neutron irradiation which leads to their in-reactor deformation and to a change of their mechanical properties. These phenomena are directly related to the microstructure evolution under irradiation and especially to the formation of -type dislocation loops. In the present work, the radiation damage evolution in recrystallized Zircaloy-4 has been studied using charged particles irradiation. The loop nucleation and growth kinetics, and also the helical climb of linear dislocations, were observed in-situ using a High Voltage Electron Microscope (HVEM) under 1 MeV electron irradiation at 673 and 723 K. In addition, 600 keV Zr+ ion irradiations were conducted at the same temperature. Transmission Electron Microscopy (TEM) characterizations have been performed after both types of irradiations, and show dislocation loops with a Burgers vector belonging to planes close to { 10 1 bar 0 } first order prismatic planes. The nature of the loops has been characterized. Only interstitial dislocation loops have been observed after ion irradiation at 723 K. However, after electron irradiation conducted at 673 and 723 K, both interstitial and vacancy loops were observed, the proportion of interstitial loops increasing as the temperature is increased. The loop growth kinetics analysis shows that as the temperature increases, the loop number density decreases and the loop growth rate tends to increase. An increase of the flux leads to an increase of the loop number density and a decrease of the loop growth rate. The results are compared to previous works and discussed in the light of point defects diffusion.

  3. Hydride phase dissolution enthalpy in neutron irradiated Zircaloy-4

    International Nuclear Information System (INIS)

    Vizcaino, Pablo; Banchik, Abraham D.

    2003-01-01

    The differential calorimetric technique has been applied to measure the dissolution enthalpy, ΔH irrad δ→α , of zirconium hydrides precipitated in structural components removed from the Argentine Atucha 1 PHWR nuclear power plant after 10.3 EFPY. An average value of ΔH irrad δ→α = 5 kJ/mol H was obtained after the first calorimetric run. That value is seven times lower than the value of ΔH δ→α = 37.7 kJ/mol H recently determined in Zircaloy-4 specimens taken from similar unirradiated structural components using the same calorimetric technique, [1]. Post-irradiation thermal treatments gradually increase that low value towards the unirradiated value with increasing annealing temperature similar to that observed for TSSd irrad . Therefore the same H atom trapping mechanism during reactor operation already proposed to explain the evolution of TSSd irrad is also valid for Q irrad δ→α . As the ratio Q/ΔH is proportional to the number N H of H atoms precipitated as hydrides, the increment of Q irrad δ→α with the thermal treatment indicates that the value of N H also grows with the annealing reaching the value corresponding to the bulk H concentration when ΔH irrad δ→α ≅ 37 kJ/mol H. That is a direct indication that the post-irradiation thermal treatment releases the H atoms from their traps increasing the number of H atoms available to precipitate at the end of each calorimetric run and/or isothermal treatment. (author)

  4. Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

    1975-09-01

    This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

  5. "Cirque du Freak."

    Science.gov (United States)

    Rivett, Miriam

    2002-01-01

    Considers the marketing strategies that underpin the success of the "Cirque du Freak" series. Describes how "Cirque du Freak" is an account of events in the life of schoolboy Darren Shan. Notes that it is another reworking of the vampire narrative, a sub-genre of horror writing that has proved highly popular with both adult and…

  6. Superficial characterization and zircaloy-2 electrochemistry with hydrothermal deposit of platinum; Caracterizacion superficial y electroquimica de zircaloy-2 con deposito hidrotermal de platino

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Arganis J, C. R.; Medina A, A. L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Gris C, M. M., E-mail: aida.contreras@inin.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2011-11-15

    The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy-2 tubes that contain in their interior UO{sub 2} pellets. With the objective of mitigating the speed of crack growth by IGSCC to a minimum negative impact on the BWR operation, General Electric developed the noble metals chemical addition (NMCA), in where noble metals particles as Pt, Pd, and Rh, are deposited on the surface of the metal to catalyze the recombination of H{sub 2} and O{sub 2}. Hydrogen is also injected to have it in excess and to favor this recombination (HWC) and zinc to reduce dose. In this work was oxidized zircaloy-2 low similar conditions to the HWC, platinum was deposited starting from a solution of Na{sub 2}Pt(OH){sub 6} with 30 ppm of Pt, in refined samples and without polishing, they were characterized by scanning electron microscopy, energy dispersed spectroscopy, XPS and electrochemistry, by means of Tafel curves and cyclical polarization. On the zircaloy surface was found a ZrO{sub 2} layer that remains under the different study conditions. Under HWC conditions is the oxides formation, possibly complex oxides of zirconium, iron and tin. After the platinum deposit these oxides decrease forming the sub-oxides: Zr{sub 2}O, Zr O, Zr{sub 2}O{sub 3}. The Tafel curves indicates the reduction of the oxygen of the sample with platinum and the cyclical polarization curves show that the reactions that happen on the zircaloy electrodes are not dur to located corrosion. (Author)

  7. The Determination of Composite Elements in Zircaloy-2 by X-Ray Fluorescence and Emission Spectrometry Method

    International Nuclear Information System (INIS)

    Dian Anggraini; Rosika Kriswarini; Yusuf N

    2007-01-01

    Analysis of composing elements in zircaloy-2 has been done by Emission Spectrometry method and X-Ray Fluorescence (XRF). The aim of the analysis is to verify conformity between composing elements in zircaloy-2 and the material certificate. Spectrometry Emission method has higher sensitivity in element determination of a material than that of XRF method, so can be estimated that emission spectrometry method has higher accuracy than that of XRF method. The result of qualitative analysis by Emission Spectrometry indicate that the composing elements in zircaloy-2 were Sn, Cr and Ni. However, the qualitative analysis result by XRF method indicated that the composing elements in zircaloy 2 were Sn, Cr, Ni and Fe. Fe element can not be analysed by Emission Spectrometry method because Emission Spectrometer did not equipped with Fe detector. The quantitative analysis result of the composing elements in the material with both methods showed that Sn, Cr and Ni concentration of zircaloy 2 existed in concentration ranges of the material certificate. Result of statistical test (F and t-test) of analysis result of both methods can be used for analyzing composing elements in zircaloy 2. Emission Spectrometry method was more sensitive and accurate for determining Cr and Ni element in zircaloy 2 than that of emission Spectrometry method but both methods had same accuracy. The precision of measurement of Sn, Cr and Ni element using XRF method was better than that of Emission spectrometry method. (author)

  8. Reaction behavior between B{sub 4}C, 304 grade of stainless steel and Zircaloy at 1473 K

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Ryosuke [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Ueda, Shigeru, E-mail: tie@tagen.tohokku.ac.jp [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Kim, Sun-Joong [Dept. of Materials Science and Engineering, Chosun University, 309, Pilmun-daero, Dong-gu, Gwangju (Korea, Republic of); Gao, Xu; Kitamura, Shin-ya [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan)

    2016-08-15

    For a better understanding of the decommissioning of the Fukushima-daiichi nuclear power plant, the melting behavior of the control blade and the channel box should be clarified. In Fukushima nuclear reactor, the channel box was made of Zircaloy-4, and the control rode is made of B{sub 4}C together with stainless steel cladding and sheath. In the study, the interaction among B{sub 4}C, stainless steel (SUS), and Zircaloy-4 was investigated at 1473 K in either argon or air atmosphere. In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted at 1473 K by the diffusion of C and B. In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt firstly. Then, the oxidized Zircaloy contacted with this melt and fused. Moreover, the progress of core melting process during severe accident under different atmospheres was firstly discussed. - Highlights: • The interaction among the system of B{sub 4}C, grade 304 stainless steel and Zircaloy-4 were studied at 1473 K in Ar and air. • In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted by the diffusion of C and B. • In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt. Then, the oxidized Zircaloy contacted with this melt and fused.

  9. Behavior and properties of Zircaloys in power reactors: A short review of pertinent aspects in LWR fuel

    International Nuclear Information System (INIS)

    Garzarolli, F.; Stehle, H.; Steinberg, E.

    1996-01-01

    Zircaloy-2 and -4, developed mainly in the US, have been used in Germany for fuel rod claddings and in-core structural components from the beginning of reactor technology. Extensive studies of the material properties of the Zircaloys have been performed in Siemens laboratories since 1957. Zircaloy-2 and -4 turned out to be very reliable materials that fulfilled all requirements for normal operation and likewise the requirements for postulated accidental conditions and for intermediate storage for many years. Optimization of Zircaloy-2 and -4 during recent years includes both optimization of microstructure and of chemical composition. BWRs and PWRs need differently optimized materials. Today's more demanding operation conditions and discharge burnups required a further optimization of the Zircaloys and for hot PWRs even the development of more corrosion-resistant Zr alloys. A significant improvement of PWR corrosion behavior can be achieved with Zr alloys using the alloying elements of Zircaloy with somewhat modified concentrations. Sn should be below or at least in the lower range of the ASTM specification range for Zircaloy-4, Fe and Cr should be somewhat higher, and Si should be specified as an alloying element rather than as an impurity

  10. Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

    International Nuclear Information System (INIS)

    Hagen, S.

    1984-01-01

    Out-of-pile experiments have been performed to investigate the escalation in temperature of Zircaloy-clad fuel rods during heatup in steam due to the exothermal Zircaloy steam reaction. In these tests single Zircaloy/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring rods--were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 2200 0 C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten Zircaloy. The escalation process was influenced by the temperature behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

  11. The effect of oxide microstructure on kinetic transition in out-of-pile steam corrosion test for Zircaloy-2 and Nb-added Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Nanikawa, Shuichi [Japan Nuclear Fuel Co. Ltd., Yokosuka, Kanagawa (Japan); Etoh, Yoshinori [Japan Nuclear Fuel Co. Ltd., Yokohama, Kanagawa (Japan)

    2001-06-01

    In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS. Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly. The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of {alpha}-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition. The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain. (author)

  12. Linear Friction Welding of Dissimilar Materials 316L Stainless Steel to Zircaloy-4

    Science.gov (United States)

    Wanjara, P.; Naik, B. S.; Yang, Q.; Cao, X.; Gholipour, J.; Chen, D. L.

    2018-02-01

    In the nuclear industry, there are a number of applications where the transition of stainless steel to Zircaloy is of technological importance. However, due to the differences in their properties there are considerable challenges associated with developing a joining process that will sufficiently limit the heat input and welding time—so as to minimize the extent of interaction at the joint interface and the resulting formation of intermetallic compounds—but still render a functional metallurgical bond between these two alloys. As such, linear friction welding, a solid-state joining technology, was selected in the present study to assess the feasibility of welding 316L stainless steel to Zircaloy-4. The dissimilar alloy welds were examined to evaluate their microstructural characteristics, microhardness evolution across the joint interface, static tensile properties, and fatigue behavior. Microstructural observations revealed a central intermixed region and, on the Zircaloy-4 side, dynamically recrystallized and thermomechanically affected zones were present. By contrast, deformation on the 316L stainless steel side was limited. In the intermixed region a drastic change in the composition was observed along with a local increase in hardness, which was attributed to the presence of intermetallic compounds, such as FeZr3 and Cr2Zr. The average yield (316 MPa) and ultimate tensile (421 MPa) strengths met the minimum strength properties of Zircaloy-4, but the elongation was relatively low ( 2 pct). The tensile and fatigue fracture of the welds always occurred at the interface in the mode of partial cohesive failure.

  13. Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

    International Nuclear Information System (INIS)

    Chung, H.M.; Thomas, G.R.

    1983-02-01

    Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850 0 C or extrapolated from the low-temperature data obtained at 0 C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500 0 C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis

  14. Stress corrosion cracking of zircaloy. The use of laboratory data to predict in-reactor behaviour

    International Nuclear Information System (INIS)

    Miller, A.K.; Ocken, H.

    1981-01-01

    Pellet-cladding interaction (PCI) can lead to failure of the Zircaloy tubing used as cladding in water-cooled reactors. Many investigations have shown that the mechanism directly responsible for such fuel rod failures is stress corrosion cracking (SCC) of Zircaloy tubing. Laboratory studies have yielded extensive data on the time-to-failure (tsub(f)) behaviour of Zircaloy tubing specimens as a function of such important variables as the applied hoop stress (σ sub(h)), the iodine concentration (I 2 ), the temperature (T) and the fluence (F). These data have been used to predict the response of Zircaloy tubing exposed in-reactor. A typical approach is to fit laboratory data to obtain an empirical equation for tsub(f) in terms of the variables identified above. The question can then be posed as to whether it is appropriate to use such an empirical expression for predicting in-reactor behaviour. This paper describes the approach which has been taken in modelling the SCC process. It first reviews the experimental observations upon which the model is based. A summary of the key features of the model is then presented. The model's capabilities, emphasizing those predictions that are independent of data used to evaluate empirical constants, are briefly discussed. Finally, it is shown how the model can be used to predict important differences between the response of tubing specimens exposed in the laboratory and the response of large quantities of tubing exposed in-reactor

  15. Investigation of the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4

    International Nuclear Information System (INIS)

    Soares, M.I.

    1981-12-01

    To investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes, deformation tests under pressure of samples hydrided in autoclave and of samples containing iodine were carried out, in order to simulate the fission product. The same tests were carried out in samples without hydride and iodine contents that were used as reference samples in the temperature range of 650 0 C-950 0 C. The hydrided samples and the samples containing iodine tested at 650 0 C and 750 0 C showed a higher ductility than the samples of reference. The hydrided samples tested at 850 0 C and 950 0 C showed a higher embritlement than the samples of reference and than the samples containing iodine tested at the same temperatures. A mechanical test has been developed to investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes. The mechanical test were carried out at room temperature. At room temperature the hydrition decreased the ductility of zircaloy-4. At room temperature the sample containing iodine showed a higher ductility than the sample without iodine. The combined action of hydrogen and iodine at room temperature enhanced the embrittlment of the samples zircaloy-4. (Author) [pt

  16. The effect of second-phase particles on the corrosion and struture of Zircaloy-4

    International Nuclear Information System (INIS)

    Cortie, M.B.

    1982-10-01

    The effect of heat treatment and second-phase particles on the corrosion resistance and microstructure of Zircaloy-4 has been examined. In particular the effect of precipitates on the rate and mechanism of high-temperature, high-pressure water or steam corrosion is discussed. Measurements of corrosion rate are presented for specimens which have received various heat treatments. The heat treatments studied included a fast cool from the beta field, prolonged annealing at temperatures ranging from 500 degrees Celsius to 1 100 degrees Celsius as well as combinations of the above. The fabrication of a small quantity of Zircaloy-4 strip was undertaken and the methods used and observations made are recorded. The wide range of microstructures produced in Zircaloy-4 by the heat treatments and fabrication procedures utilized are described and discussed with optical or electron microscope photographs showing the important features. Qualitative and semi-quantitative chemical analyses of the second-phase particles were carried out by both the scanning electron microscope and Auger spectroscopy. Evidence for the existence of a tin-rich precipitate in Zircaloy-4 is presented and discussed

  17. Air Oxidation Behaviors of Zircaloy-4 Cladding During a LOCA In Spent Fuel Pool

    International Nuclear Information System (INIS)

    Bang, Je Geon; Chun, Tae Hyun; Kim, Sun Ki; Koo, Yang Hyun

    2014-01-01

    It is well known that air oxidation induces a serious degradation of the Zircaloy cladding material, compared with steam oxidation. From the oxidant point of view, in comparison with steam, chemical heat release during oxidation in air is higher by 80%, which may lead to a more rapid degradation of the Zircaloy cladding, and further evolution of the accident.. Additionally, the oxidation kinetics in air is much faster than in steam due to the formation of non-protective oxide layer. From the safety point of view, the barrier effect of the cladding against release of fission products is lost much earlier in air compared to steam. The objective of this study is to investigate the oxidation behaviors of fuel cladding in two different conditions such as isothermal and transient condition and to generate its kinetic data under an accident condition in the spent fuel pool. In this study, the oxidation behaviors and its kinetics of the Zircaloy-4 were investigated in air environment for various air ingress scenarios in the temperature range 600 .deg. C-1,400 .deg. C by thermo-gravimetric analysis. In this study, the oxidation behaviors of the Zircaloy-4 for both isothermal condition and transient condition were investigated in air environment. In comparison with isothermal condition, the retardation of oxidation rate in transient condition was observed at both 1,200 .deg. C and 1,400 .deg. C. This seems to be ascribed to the effect of thin oxide formed during a heating

  18. Effects of deposited pyrolytic carbon on some mechanical properties of zircaloy-4 tubes. Vol. 3

    Energy Technology Data Exchange (ETDEWEB)

    Shrkawy, S W; Abdel-razek, I D; El-Sayed, H A [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Zircaloy cladding tubes are not compatible with the uranium fuel pellets as they suffer from failure due to pelletclad interaction (PCI). A carbon coating, as used in the canadian CANLUB fuel elements, is thought to improve the cladding performance with respect to the PCI problem. In this paper pyrolytic carbon coating was deposited on zircaloy-4 cladding tubes by the thermal cracking of commercial butant gas at the temperature range 250-450 degree C. In order to evaluate the effect of gaseous species on the mechanical properties of the tubes tensile and microhardness testing measurements were performed on samples prepared from the coated tubes. The fractured surface of the tensile zircaloy tubes and the deposited carbon coating, both, were examined by the SEM. The results of the tensile tests of zircaloy-4 tubes indicated that the coating process has insignificant effect on the ultimate strength of the tubes tested. The values of Vickers hardness numbers were not significantly changed across the tubes thickness. The microstructure of deposited carbon, due to the cracking process, was granular in all the temperature range (250-450 degree C) studied. 9 figs., 1 tab.

  19. Experimental studies on the crystallographic and plastic anisotropies of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Costa Viana, C.S. da

    1982-01-01

    The crystallographic and plastic anisotropies of a zircaloy-4 tubing using direct pole figures and experimental yield loci are analyzed. Tensile and plane-strain compression tests were used to assess the mecahnical behaviour. The results are discussed with respect to the dimensional stability and mechanical behaviour expected for the tube in its use in the core of pressurized water cooled reactors. (Author) [pt

  20. Microstructure in welding zone of a zircaloy 4 tube welded by TIG process

    International Nuclear Information System (INIS)

    Bolfarini, C.; Domingues Filho, H.

    1982-01-01

    The details concerned with the welding of seamless zircaloy 4 tubes for nuclear application and the earlier welding tests made in the tubes that will be used for the construction of the Argonautas' Reactor fuel element, are described. Based on the references the microestructure changes in the heat affected zone were analyzed in respect to the material's performance in operation. (Author) [pt

  1. Mechanical behavior of zircaloy-4 tubes under complexe state of stress

    International Nuclear Information System (INIS)

    Costa Viana, C.S. da

    1980-01-01

    The use of zircaloy-4 tubing as cladding material for fuel elements is reviewed with respect to its microstructural, textural and loading conditions. Its anisotropic plastic behaviour is studied through the experimental determination of its yield locus by mechanical testing and Knoop hardness and compared to Hill's anisotropic yield criterion. (Author) [pt

  2. Effect of cyclic loading on the viscoplastic behaviour of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Bouffioux, P.; Gabriel, B.; Soniak, A.; Mardon, J.P.

    1995-06-01

    Most of the electricity being generated by nuclear energy load follow and remote control have become normal operating modes in the French PWR. In addition, EDF is developing a strategy of fuel sub-assembly burnup extension. Those operating conditions will lead to cyclic straining of the Zircaloy cladding tube which could induce damages. Therefore, EDF, CEA, and FRAMATOME has started a joint R and D cooperative program in order to investigate the mechanical behaviour of Zircaloy cladding tubes under cyclic loading. This paper is dealing with the effect of a pre-cyclic loading on the plasticity properties of Zircaloy 4 cladding tubes. Load controlled cyclic tests were carried out at 350 deg. C and 0.5 Hz in both axial and hoop directions. The Woehler curves were determined. Sequential tests combining pre-cyclic loading to 50 and 75 % fraction life with tension were then performed. It has ben noticed that the pre-cycling loading does not change the plastic flow curve of the Zircaloy 4 cladding tubes and therefore does not induce observable macroscopic damage. It has been concluded that a linear cumulative damage rule like ΣΔN(σ)/N r(σ) is very conservative. (author)

  3. The effect of zinc addition on PWR corrosion product deposition on zircaloy-4

    International Nuclear Information System (INIS)

    Walters, W.S.; Page, J.D.; Gaffka, A.P.; Kingsbury, A.F.; Foster, J.; Anderson, A.; Wickenden, D.; Henshaw, J.; Zmitko, M.; Masarik, V.; Svarc, V.

    2002-01-01

    During the period 1995 to 2001 a programme of loop irradiation tests have been performed to confirm the effectiveness of zinc additions on PWR circuit chemistry and corrosion. The programme included two loop irradiation experiments, and subsequent PIE; the experiments were a baseline test (no added zinc) and a test with added zinc (10 ppb). This paper addresses the findings regarding corrosion product deposition and activation on irradiated Zircaloy-4 surfaces. The findings are relevant to overall corrosion of the reactor primary circuit, the use of zinc as a corrosion inhibitor, and activation and transport of corrosion products. The irradiation experience provides information on the equilibration of the loop chemistry, with deliberate injection of zinc. The PIE used novel and innovative techniques (described below) to obtain samples of the oxide from the irradiated Zircaloy. The results of the PIE, under normal chemistry and zinc chemistry, indicate the effect of zinc on the deposition and activation of corrosion products on Zircaloy. It was found that corrosion product deposition on Zircaloy is enhanced by the addition of zinc (but corrosion product deposition on other materials was reduced in the presence of zinc). Chemical analysis and radioisotope gamma counting results are presented, to interpret the findings. A computer model has also been used to simulate the corrosion product deposition and activation, to assist in the interpretation of the results. (authors)

  4. Release of indigenous gases from LWR fuel and the reaction kinetics with Zircaloy cladding

    International Nuclear Information System (INIS)

    Beyer, C.E.; Hann, C.R.

    1977-04-01

    The objective of this study was to evaluate the open literature data to estimate: the rate of gaseous impurity release from oxide fuel, the amount and composition of the gaseous impurities, and their subsequent rate of reaction with the fuel or Zircaloy

  5. Irradiation-induced growth of zircaloy and its effects on the mechanical design of fuel assemblies

    International Nuclear Information System (INIS)

    Yao Pu

    1991-01-01

    Zircaloy growth could be induced due to irradiation. The ammount of growth is described as a function of texture, irradiation temperature, fast neutron fluence and the reduction of cold work, and it should be given great attention in the mechanical design of fuel assemblies

  6. Fracture properties of hydrided Zircaloy-4 cladding in recrystallization and stress-relief anneal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Hsiao-Hung, E-mail: hhhsu@iner.gov.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 325, Taiwan (China); Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China); Tsay, Leu-Wen [Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China)

    2012-03-15

    In this work, the stress-relieved (SRA) and recrystallized (RXA) Zircaloy-4 cladding specimens were hydrogen-charged to the target concentration of 300 wppm and then manufactured into X-specimens for fracture toughness test. The hydrogen embrittlement susceptibility of Zircaloy-4 cladding specimens in both SRA and RXA conditions were investigated. At the hydrogen concentration level of 300 wppm, J-integral values for RXA cladding were higher than those for SRA cladding at both 25 Degree-Sign C and 300 Degree-Sign C. The formation of brittle zirconium hydrides had a significant impact on the fracture toughness of Zircaloy-4 cladding in both SRA and RXA states, especially at 25 Degree-Sign C. Among all the tests, SRA cladding tested at 25 Degree-Sign C exhibited a great loss of the fracture toughness. The micrographic and fractographic observations further demonstrated that the fracture toughness of Zircaloy-4 cladding would be improved by the coarse grains in RXA cladding, but degraded by zirconium hydrides precipitated along the grain boundary.

  7. Analysis of the tensile behaviour of zircaloy-4 in the region of dynamic strain aging

    International Nuclear Information System (INIS)

    Dellaretti Filho, O.

    1974-01-01

    An analysis of the tensile behavior of Zircaloy 4, centering around the influence of dynamic strain aging and strain rate history, is presented. This analysis is based on techniques introduced by Jaoul-Crussard and Reed-Hill. An attempt is also made to assess the experimental errors that influence these methods. (author)

  8. Contribution to study on recovery and recrystallization of cold rolling zircaloy-4

    International Nuclear Information System (INIS)

    Persiano, A.I.C.

    1977-01-01

    Recovery and recrystallization of work-hardened (40-60% - Cold rolling) Zircaloy-4 were studied between 200 and 600 0 C with times varying from 15 to 240 minutes, from electrical resistance and hardness measurements. Activation energy calculation for the recovery and recrystallization processes using the samples work-hardened 60% gave 0,7 and 2,1 eV. (author)

  9. Status of Zircaloy deformation and oxidation research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Chapman, R.H.; Cathcart, J.V.; Hobson, D.O.

    1976-01-01

    The U.S. Nuclear Regulatory Commission sponsors a broad range of research on the response of nuclear fuel assemblies to normal, off-normal, and accident conditions in light-water reactors. The paper reviews the current status of three Zircaloy cladding research programs in progress at the Oak Ridge National Laboratory and presents some preliminary results from each

  10. Oxiding and hydriding properties of Zr-1Nb cladding material in comparison with zircaloys

    Energy Technology Data Exchange (ETDEWEB)

    Vrtilkova, V; Molin, L [Nuclear Fuel Inst., Zbraslav (Czech Republic); Valach, M [Nuclear Research Inst., Rez plc (Czech Republic)

    1997-02-01

    This paper presents an overview of experimental research related to the Zr-1Nb corrosion behaviour in water and steam environment performed in the Czech Republic. Presented work is focused on the differences between Zr1Nb and Zircaloy corrosion performance. The effects of steam pressure, temperature transients and preoxidation are discussed. (author). 14 refs, 15 figs.

  11. Irradiation capsule design capable of continuously monitoring the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Thoms, K.R.; Dodd, C.V.; van der Kaa, T.; Hobson, D.O.

    1978-01-01

    An irradiation capsule which permits continuous monitoring of the creepdown of Zircaloy tubing has been designed and given preliminary tests. This design effort is the major element of a cooperative research program between the United States Nuclear Regulatory Commission and the Netherlands Energy Research Foundation (ECN) and is a part of the NRC-sponsored Zircaloy creepdown program. The purpose of the Zircaloy creepdown program is to provide data on the deformation characteristics of Zircaloy tubes, typical of LWR fuel element cladding, under combined axial and tangential compressive stresses. These data will be used to verify and improve the material behavior codes that are used for the description of fuel pin behavior. The first capsule of this series contains a mockup test specimen which was machined with three different diameters, nominally 10.92-mm, 10.54-mm and 11.30-mm (.430-in., .415-in., and .445-in.). This test specimen can be moved axially thereby varying the lift-off and serving as a calibration device for the eddy-current deformation monitoring probes. Fabrication of this capsule has been completed and during out-or-reactor checkout we were able to obtain a resolution of better than 0.01-mm (0.0004-in.). The capsule is scheduled for installation in the HFR on February 8, 1978, for a 26 day irradiation test. The first pressurized capsule, and therefore the first one to monitor in-reactor cladding deformation, will be installed in the HFR on May 3, 1978

  12. beta. -Amyloid gene dosage in Alzheimer's disease

    Energy Technology Data Exchange (ETDEWEB)

    Murdoch, G H; Manuelidis, L; Kim, J H; Manuelidis, E E

    1988-01-11

    The 4-5 kd amyloid ..beta..-peptide is a major constituent of the characteristic amyloid plaque of Alzheimer's disease. It has been reported that some cases of sporatic Alzheimer's disease are associated with at least a partial duplication of chromosome 21 containing the gene corresponding to the 695 residue precursor of this peptide. To contribute to an understanding of the frequency to such a duplication event in the overall Alzheimer's population, the authors have determined the gene dosage of the ..beta..-amyloid gene in this collection of cases. All cases had a clinical diagnosis of Alzheimer's confirmed neuropathologically. Each Alzheimer's case had an apparent normal diploid ..beta..-amyloid gene dosage, while control Down's cases had the expected triploid dosage. Thus partial duplication of chromosome 21 may be a rare finding in Alzheimer's disease. Similar conclusions were just reported in several studies of the Harvard Alzheimer collection.

  13. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  14. Hydrides formation In Zircaloy-4 irradiated with neutrons

    International Nuclear Information System (INIS)

    Vizcaino, P; Flores, A V; Vicente Alvarez, M A; Banchik, A.D; Tolley, A; Condo, A; Santisteban, J R

    2012-01-01

    Under reactor operating conditions zirconium components go through transformations which affect their original properties. Two phenomena of significant consequences for the integrity of the components are hydrogen uptake and radiation damage, since both contribute to the material fragilization. In the case of the Atucha I nuclear power reactor, the cooling channels, Zircaloy-4 tubular structural components about 6 meters long, were designed to withstand the entire lifetime of the reactor. Inside them, fuel elements 5.3 meters long are located. The fuel elements are cooled by a heavy water flow which circulates from the bottom (250 o ) to the top of the reactor (305 o C). The channels are affected by a fast neutron flux (En>1 Mev), increasing from a nominal value of 1.35 x 10 13 neutrons/cm 2 sec at the bottom to 1.69 x 10 13 neutrons/cm 2 sec at the top, reaching a maximum value of 3.76 x 10 13 neutrons/cm 2 sec at the center of the channels. However, due to the reactor operating conditions, they are replaced after about 10 effective full power years, time at which they reach 10 22 neutrons/cm 2 at the most neutronically active regions of the reactor. Studies on cooling channels are meaningful from many points of view. The channels are structural components which do not work under internal pressure or any other type of structural stress. The typical temperature of the cladding tubes in the reactor is about 350 o C, at which many types of irradiation defects are annealed [1]. The temperature range of the cooling channels lies between 200 o C-235 o C (outer foil of the channels) and 260 o C-300 o C (internal tube), a difference which makes the defect recovery kinetics slower. In the present context, following the program developed in the research contract 15810, we continue with the work started on the effects of the radiation on the hydride formation focusing on the dislocation loops in the zirconium matrix and its possible role as preferential sites for hydride

  15. Ecologie du phytoplancton du lac Kivu

    Directory of Open Access Journals (Sweden)

    Sarmento, H.

    2008-01-01

    Full Text Available Speciation within the African Coffee Pathogen. Cet article analyse s'il est avantageux d'utiliser le compost au lieu de l'engrais minéral pour produire la laitue dans la zone urbaine et péri-urbaine de Yaoundé. Les résultats de terrain montrent l'obtention de rendements et profits plus élevés lorsqu'on utilise le compost. Les résultats de la fonction de production Cobb-Douglas prouvent que l'utilisation du compost est statistiquement significative pour expliquer la variation de rendement de la laitue et que le compost est l'intrant le plus productif. D'autres résultats montrent que le compost fournit la matière organique utile au sol et que les besoins d'irrigation en eau de la culture sont réduits grâce à l'utilisation du compost. Par conséquent, malgré le fait que l'application du compost demande une main-d'oeuvre beaucoup plus élevée, son utilisation est généralement bénéfique pour les agriculteurs vivant aux alentours de Yaoundé. Les programmes de vulgarisation de cet intrant pour encourager son adoption devraient donc figurer parmi les points prioritaires dans la politique agricole du gouvernement camerounais.

  16. Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

    Science.gov (United States)

    Yan, Yong; Qian, Shuo; Garrison, Ben; Smith, Tyler; Kim, Peter

    2018-04-01

    A nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0 wt. % at 1100 °C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.

  17. Intelligent system for improving dosage control

    Directory of Open Access Journals (Sweden)

    Fabio Cosme Rodrigues dos Santos

    2017-02-01

    Full Text Available Coagulation is one of the most important processes in a drinking-water treatment plant, and it is applied to destabilize impurities in water for the subsequent flocculation stage. Several techniques are currently used in the water industry to determine the best dosage of the coagulant, such as the jar-test method, zeta potential measurements, artificial intelligence methods, comprising neural networks, fuzzy and expert systems, and the combination of the above-mentioned techniques to help operators and engineers in the water treatment process. Current paper presents an artificial neural network approach to evaluate optimum coagulant dosage for various scenarios in raw water quality, using parameters such as raw water color, raw water turbidity, clarified and filtered water turbidity and a calculated Dose Rate to provide the best performance in the filtration process. Another feature in current approach is the use of a backpropagation neural network method to estimate the best coagulant dosage simultaneously at two points of the water treatment plant. Simulation results were compared to the current dosage rate and showed that the proposed system may reduce costs of raw material in water treatment plant.

  18. Spectrophotometric Determination of Trimipramine in Tablet Dosage ...

    African Journals Online (AJOL)

    Purpose: To develop and validate simple, rapid and sensitive spectrophotometric procedures for determination of trimipramine in tablet dosage form. Methods: The methods were based on the interaction of trimipramine as n-electron donor with the ο-acceptor, iodine and various π-acceptors, namely: chloranil (CH), ...

  19. A brief history of dosage compensation

    Indian Academy of Sciences (India)

    depression of X-linked gene activity in the female, as well as by hyperexpression of the ... to the Harvey lecture, Muller had presented important ideas relative to dosage ... at Columbia. I do recall a talk by the popular physical anthro- pologist ...

  20. Chemical interactions between as-received and pre-oxidized Zircaloy-4 and Inconel-718 at high temperatures

    International Nuclear Information System (INIS)

    Hofmann, P.; Markiewicz, M.

    1994-06-01

    Isothermal reaction experiments were performed in the temperature range of 1000 - 1300 C in order to determine the chemical interactions between Zircaloy-4 fuel rod cladding and Inconel-718 spacer grids of Pressurized Water Reactors (PWR) under severe accident conditions. It was not possible to apply even higher temperatures since fast and complete liquefaction of the components occurred as a result of eutectic interactions during heatup. The liquid reaction products formed enhance and accelerate the degradation of the material couples and the fuel elements, respectively. Only small amounts of Inconel are necessary to liquefy large amounts of Zircaloy. Thin oxide layers on the Zircaloy surface delay the beginning of the chemical interactions with Inconel but cannot prevent them. In this work the reaction kinetics have been determined for the system: as-received and pre-oxidized Zircaloy-4/Inconel 718. The interactions can be described by parabolic rate laws; the Arrhenius equations for the various interactions are given. (orig.) [de

  1. Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy; Desenvolvimento de processos de reciclagem de cavacos de zircaloy via refusao em forno eletrico a arco e metalurgia do po

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Luiz Alberto Tavares

    2014-09-01

    PWR reactors employ, as nuclear fuel, UO{sub 2} pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

  2. Du Pont de Nemours

    NARCIS (Netherlands)

    Ros JPM; LAE

    1994-01-01

    Dit rapport over Du Pont de Nemours (produktie van o.a. chemische stoffen) is gepubliceerd binnen het Samenwerkingsproject Procesbeschrijvingen Industrie Nederland (SPIN). In het kader van dit project is informatie verzameld over industriele bedrijven of industriele processen ter ondersteuning

  3. Les Cahiers du CREAD

    African Journals Online (AJOL)

    Admin

    6 juil. 2007 ... La problématique du développement du secteur de l'artisanat en. Algérie a été très peu abordée par les chercheurs, qu'ils soient universitaires ou .... La loi a institué une taxe d'apprentissage dont le taux a été fixé à. 1% de la ...

  4. Les outils du CERN

    CERN Multimedia

    1999-01-01

    C'est le plus grand centre mondial de recherche en physique des particules. Les outils du Laboratoire, accélérateurs et détecteurs de particules, figurent parmi les instruments scientifiques les plus complexes au monde. Des prix Nobels ont d'ailleurs été attribués aux physiciens du CERN pour leurs développements.

  5. Bulletin du CRDI #127

    International Development Research Centre (IDRC) Digital Library (Canada)

    La mise à l'échelle de la recherche et de l'innovation en vue de créer un impact social constitue une priorité pour la communauté du développement. Toutefois ... Nous avons renouvelé notre soutien à la recherche auprès du gouvernement de l'Inde ... Des femmes étudient à l'École supérieure d'infotronique d'Haïti.

  6. Comparison of the air oxidation behaviors of Zircaloy-4 implanted with yttrium and cerium ions at 500 deg. C

    International Nuclear Information System (INIS)

    Chen, X.W.; Bai, X.D.; Xu, J.; Zhou, Q.G.; Chen, B.S.

    2002-01-01

    As a valuable process for surface modification of materials, ion implantation is eminent to improve mechanical properties, electrochemical corrosion resistance and oxidation behaviors of varieties of materials. To investigate and compare the oxidation behaviors of Zircaloy-4, implantation of yttrium ion and cerium ion were respectively employed by using an MEVVA source at the energy of 40 keV with a dose ranging from 1x10 16 to 1x10 17 ions/cm 2 . Subsequently, weight gain curves of the different specimens including as-received Zircaloy-4 and Zircaloy-4 specimens implanted with the different ions were measured after oxidation in air at 500 deg. C for 100 min. It was obviously found that a significant improvement was achieved in the oxidation behaviors of implanted Zircaloy-4 compared with that of the as-received Zircaloy-4, and the oxidation behavior of cerium-implanted Zircaloy-4 was somewhat better than that of yttrium-implanted specimen. To obtain the valence and the composition of the oxides in the scale, X-ray photoemission spectroscopy was used in the present study. Glancing angle X-ray diffraction, employed to analyze the phase transformation in the oxide films, showed that the addition of yttrium transformed the phase from monoclinic zirconia to tetragonal zirconia, yet the addition of cerium transformed the phase from monoclinic zirconia to hexagonal zirconia. In the end, the mechanism of the improvement of the oxidation behavior was discussed

  7. Chemical interactions between as-received and pre-oxidized Zircaloy-4 and stainless steel at high temperatures

    International Nuclear Information System (INIS)

    Hofmann, P.

    1994-05-01

    The chemical reaction behavior between Zircaloy-4 and 1.4919 (AISI 316) stainless steel, which are used in absorber assemblies of Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR), has been studied in the temperature range 1000 - 1400 C. Zircaloy was used in the as-received, pre-oxidized and oxygen-containing condition. The maximum temperature was limited by the fast and complete liquefaction of the reaction couple as a result of eutectic chemical interactions. Liquefaction of the components occurs below their melting point. The effect of oxygen dissolved in Zircaloy plays an important role in the interaction; oxide layers on the Zircaloy surface delay the chemical interactions with stainless steel but cannot prevent them. Oxygen dissolved in Zircaloy reduces the reaction rates and shift the liquefaction temperature to slightly higher levels. The interaction experiments at the examined temperatures with or without pre-oxidized Zircaloy can be described by parabolic rate laws. The Arrhenius equations for the various conditions of interactions are given. (orig.) [de

  8. Identification of the zirconium hydrides metallography in zircaloy-2; Contribucion al estudio por metalografia de los hidruros de circonio en Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gonzalez, F

    1968-07-01

    Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs.

  9. Optimizing the dosage of stabilizing chemical

    OpenAIRE

    Harjula, Tomi

    2013-01-01

    A chemical company provides chemical treatment at customer mill in paper industry. This thesis work was done to determine the optimum dosage of stabilizing chemical. The theoretical framework explains the basics of paper brightness and bleaching and how these topics are connected to each other. The knowledge gained is very valuable and can possibly be used in the future in other similar applications as well. This thesis work contains confidential back ground information. Key ...

  10. Elucidating the iodine stress corrosion cracking (SCC) process for zircaloy tubing

    International Nuclear Information System (INIS)

    Nagai, M.; Shimada, S.; Nishimura, S.; Amano, K.

    1984-01-01

    Several experimental investigations were made to enhance understanding of the iodine stress corrosion cracking (SCC) process for Zircaloy: (1) oxide penetration process, (2) crack initiation process, and (3) crack propagation process. Concerning the effect of the oxide layer produced by conventional steam-autoclaving, no significant difference was found between results for autoclaved and as-pickled samples. Tests with 15 species of metal iodides revealed that only those metal iodides which react thermodynamically with zirconium to produce zirconium tetraiodide (ZrI 4 ) caused SCC of Zircaloy. Detailed SEM examinations were made on the SCC fracture surface of irradiated specimens. The crack propagation rate was expressed with a da/dt=C Ksup(n) type equation by combining results of tests and calculations with a finite element method. (author)

  11. Superior corrosion resistance properties of TiN-based coatings on Zircaloy tubes in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Khatkhatay, Fauzia [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Jiao, Liang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jian, Jie [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Zhang, Wenrui [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jiao, Zhijie [Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109-2104 (United States); Gan, Jian; Zhang, Hongbin [Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Zhang, Xinghang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Department of Mechanical Engineering, Texas A and M University, College Station, TX 77843-3123 (United States); Wang, Haiyan, E-mail: wangh@ece.tamu.edu [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States)

    2014-08-01

    Thin films of TiN and Ti{sub 0.35}Al{sub 0.65}N nanocomposite were deposited on polished Zircaloy-4 tubes. After exposure to supercritical water for 48 h, the coated tubes are remarkably intact, while the bare uncoated tube shows severe oxidation and breakaway corrosion. X-ray diffraction patterns, secondary electron images, backscattered electron images, and energy dispersive X-ray spectroscopy data from the tube surfaces and cross-sections show that a protective oxide, formed on the film surface, effectively prevents further oxidation and corrosion to the Zircaloy-4 tubes. This result demonstrates the effectiveness of thin film ceramics as protective coatings under extreme environments.

  12. Temperature estimates from the Zircaloy oxidation kinetics in the α plus β phase region

    International Nuclear Information System (INIS)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of Zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of Zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near the thermocouple locations

  13. Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2012-03-15

    The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

  14. Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-04-15

    Zirconium and zircaloy-2 alloy, as constructive materials, have found wide application in reactor technology, especially in heavy water systems for two reasons: a) low neutron absorption cross section, b) good mechanical properties. The thickness of the zirconium and zircaloy-2 for different applications varies from several tenths of a millimeter to about ten millimeters. Therefore, to calculate reactor systems it is desirable to know the effective neutron absorption cross section for the range of thicknesses mention above. The thermal neutron cross sections for these materials are low and no appreciable variation of the effective neutron cross section occurs even for the largest thicknesses. However, this is not true for effective resonance absorption. On the other hand, due to the lack of detailed knowledge of the zirconium resonances, calculations of the effective resonance integrals cannot be performed. Therefore it is necessary to measure the effective total and resonance absorption cross section for zirconium (author)

  15. Effect of current density on the anodic behaviour of zircaloy-4 and niobium: a comparative study

    International Nuclear Information System (INIS)

    Raghunath Reddy, G.; Lavanya, A.; Ch Anjaneyulu

    2004-01-01

    The kinetics of anodic oxidation of zircaloy-4 and niobium have been studied at current densities ranging from 2 to 14 mA.cm -2 at room temperature in order to investigate the dependence of ionic current density on the field across the oxide film. Thickness of the anodic films were estimated from capacitance data. The formation rate, current efficiency and differential field were found to increase with increase in the ionic current density for both zircaloy-4 and niobium. Plots of the logarithm of formation rate vs. logarithm of the current density are fairly linear. From linear plots of logarithm of ionic current density vs. differential field, and applying the Cabrera-Mott theory, the half-jump distance and the height of the energy barrier are deduced and compared. (author)

  16. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho

    2016-01-01

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  17. Investigations of the interaction between ballooning Zircaloy cladding and emergency core cooling

    International Nuclear Information System (INIS)

    Wiehr, K.; Barth, S.; Erbacher, F.; Hame, W.; Harten, U.; Just, W.; Megerle, A.; Mueller, S.; Neitzel, H.J.; Reimann; Schaeffner, P.; Schmidt, H.

    1975-01-01

    The development of fabrication methods for the production of fuel rod simulators has been largely terminated. For welding of Zircaloy-4 and inconel 600 explosive welding has proved to be promissory in preliminary tests. A prototype fuel rod simulator was tested at full power. Its performance was faultless and the fuel rod and ring pellets could be easily dismantled and reused after the experiment. Planning of the test rig and electricity supply were terminated. Most of the assembly work has been finished. For electric heating of the fuel rod simulators a special device was built and tested which allows to program the power control. The radiographic system recording ballooning of the Zircaloy clad was erected outside the test space and put into operation. First trial pictures yielded good results. (orig.) [de

  18. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  19. A new strain gage method for measuring the contractile strain ratio of Zircaloy tubing

    International Nuclear Information System (INIS)

    Hwang, S.K.; Sabol, G.P.

    1988-01-01

    An improved strain gage method for determining the contractile strain ratio (CSR) of Zircaloy tubing was developed. The new method consists of a number of load-unload cyclings at approximately 0.2% plastic strain interval. With this method the CSR of Zircaloy-4 tubing could be determined accurately because it was possible to separate the plastic strains from the elastic strain involvement. The CSR values determined by use of the new method were in good agreement with those calculated from conventional post-test manual measurements. The CSR of the tubing was found to decrease with the amount of deformation during testing because of uneven plastic flow in the gage section. A new technique of inscribing gage marks by use of a YAG laser is discussed. (orig.)

  20. Plastic strain accumulation during asymmetric cyclic loading of Zircaloy-2 at room temperature

    International Nuclear Information System (INIS)

    Rajpurohit, R.S.; Santhi Srinivas, N.C.; Singh, Vakil

    2016-01-01

    Asymmetric cyclic loading leads to accumulation of cyclic plastic strain and reduces the fatigue life of components. This phenomenon is known as ratcheting fatigue. Zircaloy-2 is a important structural material in nuclear reactors and used as pressure tubes and fuel cladding in pressurized light and heavy water nuclear reactors. Due to power fluctuations, these components experience plastic strain cycles in the reactor and their life is reduced due to strain cycles. Power fluctuations also cause asymmetric straining of the material and leads to accumulation of plastic strain. The present investigation deals with the effect of the magnitude of mean stress, stress amplitude and stress rate on hardening/softening behavior of Zircaloy-2 under asymmetric cyclic loading, at room temperature. It was observed that plastic strain accumulation increased with mean stress and stress amplitude; however, it decreased with stress rate. (author)

  1. Stress corrosion of Zircaloy-4. Fracture mechanics study of the intergranular - transgranular transition

    International Nuclear Information System (INIS)

    Farina, Silvia B.; Duffo, Gustavo S.

    2003-01-01

    Stress corrosion cracking susceptibility of Zircaloy-4 wires was studied in 1M NaCl, 1M KBr and 1M KI aqueous solutions, and in iodine alcoholic solutions. In all cases, intergranular attack preceded transgranular propagation. It is generally accepted that the intergranular-transgranular transition occurs when a critical value of the stress intensity factor is reached. In the present work it was confirmed that the transition from intergranular to transgranular propagation cracking in Zircaloy-4 wires also occurs when a critical value of the stress intensity factor is reached. This critical stress intensity factor in wire samples is independent of the solution tested and close to 10 MPa.m-1/2. This value is in good agreement with those reported in the literature measured by different techniques. (author)

  2. Zircaloy cladding corrosion degradation in a Tuff repository: initial experimental plan

    International Nuclear Information System (INIS)

    Smith, H.D.

    1984-07-01

    The projected environmental history of a Tuff repository sited in an unsaturated hydrologic setting is evaluated to identify the potentially most severe corrosion conditions for Zircaloy spent fuel cladding. Three distinct corrosion periods are identified over the projected history. In two of those, liquid water may be present which is believed to produce the most severe corrosive environment for Zircaloy spent fuel cladding. In the time interval 100 to 1000 years after emplacement in the repository, the most severe condition is exposure to 170 0 C water at about 100 psi in an unbreached canister. This condition will be reproduced experimentally in an autoclave. For times after 1000 years, the most severe condition is exposure to 90 0 C water that is equilibrated with the tuff and invades breached canisters. This condition will be reproduced with a water bath system

  3. The influence of hydride on fracture toughness of recrystallized Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Hsiao-Hung, E-mail: 175877@mail.csc.com.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China); China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chiang, Ming-Feng [China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chen, Yen-Chen [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2014-04-01

    In this work, RXA cladding tubes were hydrogen-charged to target hydrogen content levels between 150 and 800 wppm (part per million by weight). The strings of zirconium hydrides observed in the cross sections are mostly oriented in the circumferential direction. The fracture toughness of hydrided RXA Zircaloy-4 cladding was measured to evaluate its hydride embrittlement susceptibility. With increasing hydrogen content, the fracture toughness of hydrided RXA cladding decreases at both 25 °C and 300 °C. Moreover, highly localized hydrides (forming a hydride rim) aggravate the degradation of the fracture properties of RXA Zircaloy-4 cladding at both 25 °C and 300 °C. Brittle features in the form of quasi-cleavages and secondary cracks were observed on the fracture surface of the hydride rim, even for RXA cladding tested at 300 °C.

  4. Experiments on ballooning in pressurized and transiently heated Zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Markiewicz, M.E.; Erbacher, F.J.

    1988-02-01

    Single-rod burst tests were performed with Atucha I Zircaloy-4 cladding tubes in the REBEKA burst equipment of KfK. The objective was to investigate the ballooning and burst behavior of argentine cladding tubes obtained from NRG, Germany and CONVAR, Argentina. The burst data were compared with those of cladding tubes used in german PWR's. It was found that the burst data e.g. burst temperature, circumferential burst strain and its response to azimuthal temperature differences are identical for the Argentine and German tubing quality. The burst data are in good agreement with those of German PWR-Zircaloy tubes. Thus, the fuel rod behavior codes developed for German PWR's can also be used for the Argentine reactor Atucha I. (orig.) [de

  5. Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Malang, S.; Neitzel, H.J.

    1983-01-01

    Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

  6. The steam pressure effect on high temperature corrosion of zircaloy-4

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, G. H.

    1998-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The degree of high temperature oxidation of zircaloy-4 was measured at three different conditions, high pressure steam, high pressure Ar gas with small amount of steam, and 1 atm steam. All the measurements were done at 750 deg C. The oxide thickness is much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. No effect was observed in the case of high pressure Ar containing small amount of steam. Many cracks exist on the surface of specimens oxidized at high pressure steam, which come from the enhanced tetragonal to monoclinic phase transformation due to high pressure steam. The enhanced oxidation seems to oxide cracking

  7. Effect of ageing time and temperature on the strain ageing behaviour of quenched zircaloy-4

    International Nuclear Information System (INIS)

    Rheem, K.S.; Park, W.K.; Yook, C.C.

    1977-01-01

    The strain ageing behaviour of quenched Zircaloy-4 has been studied as a function of ageing time and temperature in the temperature range 523-588 K for a short-ageing time of 1 to 52 seconds. A the test conditions, the strain ageing stress increased with ageing time and temperature at a strain rate of 5.55x10 -4 sec -1 . Applying stress on the quenched Zircaloy-4, the strain ageing effect indicated following two states: an initial stage having an activation energy of 0.39ev considered to be due to Snoek type ordering of interstitial oxygen atoms in the stress field of a dislocaiton and a second stage havingan activation energy of 0.60 ev, due to mainly long range diffusion of oxygen atoms. (author)

  8. Microbial quality of some herbal solid dosage forms

    African Journals Online (AJOL)

    STORAGESEVER

    2010-03-15

    Mar 15, 2010 ... Key words: Microbial quality, herbal, contamination, solid dosage form ... The type of dosage form, packaging, manufacturing and expiration dates of subject solid herbal .... According to WHO report (2002), Salmonella food.

  9. A fracture mechanics model for iodine stress corrosion crack propagation in Zircaloy tubing

    International Nuclear Information System (INIS)

    Crescimanno, P.J.; Campbell, W.R.; Goldberg, I.

    1984-01-01

    A fracture mechanics model is presented for iodine-induced stress corrosion cracking in Zircaloy tubing. The model utilizes a power law to relate crack extension velocity to stress intensity factor, a hyperbolic tangent function for the influence of iodine concentration, and an exponential function for the influence of temperature and material strength. Comparisons of predicted to measured failure times show that predicted times are within a factor of two of the measured times for a majority of the specimens considered

  10. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Price, E G [ed.

    1994-09-01

    Under the auspices of the IAEA, a consultants` meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena.

  11. The characteristics of surface oxidation and corrosion resistance of nitrogen implanted zircaloy-4

    International Nuclear Information System (INIS)

    Tang, G.; Choi, B.H.; Kim, W.; Jung, K.S.; Kwon, H.S.; Lee, S.J.; Lee, J.H.; Song, T.Y.; Shon, D.H.; Han, J.G.

    1997-01-01

    This work is concerned with the development and application of ion implantation techniques for improving the corrosion resistance of zircaloy-4. The corrosion resistance in nitrogen implanted zircaloy-4 under a 120 keV nitrogen ion beam at an ion dose of 3 x 10 17 cm -2 depends on the implantation temperature. The characteristics of surface oxidation and corrosion resistance were analyzed with the change of implantation temperature. It is shown that as implantation temperature rises from 100 to 724 C, the colour of specimen surface changes from its original colour to light yellow at 100 C, golden at 175 C, pink at 300 C, blue at 440 C and dark blue at 550 C. As the implantation temperature goes above 640 C, the colour of surface changes to light black, and the surface becomes a little rough. The corrosion resistance of zircaloy-4 implanted with nitrogen is sensitive to the implantation temperature. The pitting potential of specimens increases from 176 to 900 mV (SCE) as the implantation temperature increases from 100 to 300 C, and decreases from 900 to 90 mV(SCE) as the implantation temperature increases from 300 to 640 C. The microstructure, the distribution of oxygen, nitrogen and carbon elements, the oxide grain size and the feature of the precipitation in the implanted surface were investigated by optical microscope, TEM, EDS, XRD and AES. The experimental results reveal that the ZrO 2 is distributed mainly on the outer surface. The ZrN is distributed under the ZrO 2 layer. The characteristics of the distribution of ZrO 2 and ZrN in the nitrogen-implanted zircaloy-4 is influenced by the implantation temperature of the sample, and in turn the corrosion resistance is influenced. (orig.)

  12. Programme of CABRI start-up measurements with the Zircaloy loop

    International Nuclear Information System (INIS)

    Kussmaul, G.; Rongier, P.

    1981-06-01

    After installation and operational tests of the CABRI Zircaloy loop, a start-up test programme will be carried out to determine the new coupling value between the driver core and the test pin and the reactivity dependent driver core energy release for transients from different power levels and modified injection rates. The purpose of the tests and the test programme itself are described in the report

  13. Experimental determination of resonance absorption cross sections for Zircaloy-2 and zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Markovic, V [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1968-05-15

    The integral absorption cross section for the neutron spectrum and the thermal absorption cross section for zircaloy-2 have been determined using the pile oscillator technique. Using both values and a measured ratio of the epithermal to the thermal flux, the effective resonance integrals were obtained. After subtraction of the contributions for alloy and impurity elements, the effective resonance integrals for zirconium were evaluated. An extrapolated value of 0.91{+-}0.10 was obtained for the dilute integral. (author)

  14. The effects of irradiation and temperature on the growth of Zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Kendoush, A.A.

    1987-01-01

    The growth strain was measured after irradiation for 16 Zircaloy-4 tubes of the recrystallised and stress relieved types. The operating temperature during irradiation ranged between 317 and 344 0 C. The average fast neutron fluence was 9.6x10 20 n/cm 2 . Experimental results indicated the dependence of the growth on the irradiation temperature. The stress relieved result was compared with data of the literature. (orig.)

  15. Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

    International Nuclear Information System (INIS)

    Minato, K.; Hering, W.; Hagen, S.

    1991-07-01

    Out-of-pile bundle experiments (zircaloy 4) are performed in the CORA facility to investigate the behavior of PWR fuel elements during severe fuel damage (SFD) accidents. Within the international cooperation the most significant phenomena such as cladding deformation, oxidation (especially the zirconium/steam reaction), melt formation, melt release, and relocation which were found in all tests have been analyzed. (orig./MM) [de

  16. Determination of lower bound crystallographic yield loci of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Costa Viana, C.S. da

    1980-01-01

    The use of zircaloy-4 tubing in fuel elements of water cooled reactors is discussed with respect to its mechanisms of deformation and also its resulting anisotropic plastic behaviour. A method for obtaining lower bound crystallographic yield loci of α-Zr is presented and applied to individual crystal orientations and to a real texture described by the main components observed on a direct pole figure. (Author) [pt

  17. The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

    Science.gov (United States)

    Cockeram, B. V.; Kammenzind, B. F.

    2018-02-01

    Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

  18. Observations on the ductility of zircaloy-2 under simultaneous tension and bending

    International Nuclear Information System (INIS)

    Pettersson, K.

    1975-01-01

    The ductility of Zircaloy-2 in creep-fatigue interaction tests has been found to exceed the ductility in separate tensile tests. It was shown that the increase of ductility was due to either the suppression of the localized shear band instability causing final failure in a tensile test, or because the hydrostatic tension-shear stress ratio in the creep-fatigue test is lower than in the tensile test. Possible applications of the ductility increase in forming operations are suggested. (author)

  19. Zircaloy-4 and M5 high temperature oxidation and nitriding in air

    Energy Technology Data Exchange (ETDEWEB)

    Duriez, C. [Institut de Radioprotection et Surete Nucleaire, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Dupont, T.; Schmet, B.; Enoch, F. [Universite Technologique de Troyes, BP 2060, 10010 Troyes (France)

    2008-10-15

    For the purpose of nuclear power plant severe accident analysis, degradation of Zircaloy-4 and M5 cladding tubes in air at high temperature was investigated by thermo-gravimetric analysis, in isothermal conditions, in a 600-1200 deg. C temperature range. Alloys were investigated either in a 'as received' bare state, or after steam pre-oxidation at 500 {sup o}C to simulate in-reactor corrosion. At the beginning of air exposure, the oxidation rate obeys a parabolic law, characteristic of solid-state diffusion limited regime. Parabolic rate constants compare, for Zircaloy-4 as well as for M5, with recently assessed correlations for high temperature Zircaloy-4 steam-oxidation. A thick layer of dense protective zirconia having a columnar structure forms during this diffusion-limited regime. Then, a kinetic transition (breakaway type) occurs, due to radial cracking along the columnar grain boundaries of this protective dense oxide scale. The breakaway is observed for a scale thickness that strongly increases with temperature. At the lowest temperatures, the M5 alloy appears to be breakaway-resistant, showing a delayed transition compared to Zircaloy-4. However, for both alloys, a pre-existing corrosion scale favours the transition, which occurs much earlier. The post transition kinetic regime is linear only for the lowest temperatures investigated. From 800 deg. C, a continuously accelerated regime is observed and is associated with formation of a strongly porous non-protective oxide. A mechanism of nitrogen-assisted oxide growth, involving formation and re-oxidation of ZrN particles, as well as nitrogen associated zirconia phase transformations, is proposed to be responsible for this accelerated degradation.

  20. A unified model to describe the anisotropic viscoplastic behavior of Zircaloy-4 cladding tubes

    International Nuclear Information System (INIS)

    Delobelle, P.; Robinet, P.; Bouffioux, P.; Geyer, P.; Pichon, I. Le

    1996-01-01

    This paper presents the constitutive equations of a unified viscoplastic model and its validation with experimental data. The mechanical tests were carried out in a temperature range of 20 to 400 C on both cold-worked stress-relieved and fully annealed Zircaloy-4 tubes. Although their geometry (14.3 by 1.2 mm) is different, the crystallographic texture was close to that expected in the cladding tubes. To characterize the anisotropy, mechanical tests were performed under both monotonic and cyclic uni- and bi-directional loadings, i.e., tension-compression, tension-torsion, and tension-internal pressure tests. The results obtained at ambient temperatures and the independence of the ratio R p = var-epsilon θθ p /var-epsilon zz p , with respect to temperature would seem to indicate that the set of anisotropy coefficients does not depend on temperature. Zircaloy-4 material also has a slight supplementary hardening during out-of-phase cyclic loading. The authors propose to extend the formulation of a unified viscoplastic model, developed and identified elsewhere for other initially isotropic materials, to the case of Zircaloy-4. Generally speaking, anisotropy is introduced through fourth order tensors affecting the flow directions, the linear kinematical hardening components, as well as the dynamic and static recoveries of the forementioned hardening variables. The ability of the model to describe all the mechanical properties of the material is shown. The application of the model to simulate mechanical tests (tension, creep, and relaxation) performed on true CWSR Zircaloy-4 cladding tubes with low tin content is also presented

  1. Influence of surface treatment on the oxidation behavior of zirconium and zircaloy-4

    International Nuclear Information System (INIS)

    Costa, I.; Ramanathan, L.V.

    1986-01-01

    The influence of fluoride concentration in surface treatment solutions on the oxidation behavior of Zr and Zircaloy-4 in the temperature range 350-760 0 C have been studied by means of thermogravimetric analysis. Two solutions containing different concentrations of hydrofluoric acid have been used for surface treatments, following which surface roughness measurements were also carried out. The influence of fluoride ion concentration on oxidation behavior has been found to be significant at higher temperatures. (Author) [pt

  2. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Gilbert, E.R.

    1983-09-01

    The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

  3. Effects of oxidation in the mechanical behavior of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos.

    1981-07-01

    The kinetics of oxidation of zircaloy-4 is isothermally studied utilizing discontinous gravimetric method under two different oxidizing conditions, using gaseous oxigen and steam. The total weight gain during oxidation occurs in two different way: formation of oxide and solid solution. A mechanical test for studying the effect of embrittlement due to the absorption of oxygen in small zircalloy tubes have been developed. (Author) [pt

  4. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  5. Effect of annealing temperature on the mechanical properties of Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Beauregard, R.J.; Clevinger, G.S.; Murty, K.L.

    1977-01-01

    The mechanical properties of Zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced Zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. The burst strength of the cladding at 650F decreased with the annealing temperature reaching a saturation value at approximately 1000F. The total circumferential elongation increased with the annealing temperature reaching a maximum at approximately 1000F and decreasing at higher temperatures. Hoop creep characteristics of Zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. The R-parameter was essentially independent of the annealing temperature while the P-parameter increased with annealing temperature. The mechanical anisotropy parameters were also studied as a function of the test temperature from ambient to approximately 800F using continuously monitored high precision extensometry. (Auth.)

  6. Brazing characteristics of a Zr-Ti-Cu-Fe eutectic alloy filler metal for Zircaloy-4

    Science.gov (United States)

    Lee, Jung G.; Lim, C. H.; Kim, K. H.; Park, S. S.; Lee, M. K.; Rhee, C. K.

    2013-10-01

    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.%) showing a low melting temperature range from 832 °C to 853 °C was designed by a partial substitution of Zr with Ti based on a Zr-Cu-Fe ternary eutectic system. By applying an alloy ribbon with the determined composition, a highly reliable joint was obtained with a homogeneous formation of predominantly grown α-Zr phases owing to a complete isothermal solidification, exhibiting strength higher than that of Zircaloy-4. The homogenization of the joint was rate-controlled by the diffusion of the filler elements (Ti, Cu, and Fe) into the Zircaloy-4 base metal, and the detrimental segregation of the Zr2Fe phase in the central zone was completely eliminated by an isothermal holding at a brazing temperature of 920 °C for 10 min.

  7. Examination of Zircaloy-clad spent fuel after extended pool storage

    International Nuclear Information System (INIS)

    Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

    1981-09-01

    This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed

  8. Pool boiling CHF enhancement by micro/nanoscale modification of zircaloy-4 surface

    International Nuclear Information System (INIS)

    Ahn, Ho Seon; Lee, Chan; Kim, Hyungdae; Jo, HangJin; Kang, SoonHo; Kim, Joonwon; Shin, Jeongseob; Kim, Moo Hwan

    2010-01-01

    Consideration of the critical heat flux (CHF) requires difficult compromises between economy and safety in many types of thermal systems, including nuclear power plants. Much research has been directed towards enhancing the CHF, and many recent studies have revealed that the significant CHF enhancement in nanofluids is due to surface deposition of nanoparticles. The surface deposition of nanoparticles influenced various surface characteristics. This fact indicated that the surface wettability is a key parameter for CHF enhancement and so is the surface morphology. In this study, surface wettability of zircaloy-4 used as cladding material of fuel rods in nuclear power plants was modified using surface treatment technique (i.e. anodization). Pool boiling experiments of distilled water on the prepared surfaces was conducted at atmospheric and saturated conditions to examine effects of the surface modification on CHF. The experimental results showed that CHF of zircaloy-4 can be significantly enhanced by the improvement in surface wettability using the surface modification, but only the wettability effect cannot explain the CHF increase on the treated zircaloy-4 surfaces completely. It was found that below a critical value of contact angle (10 o ), micro/nanostructures created by the surface treatment increased spreadability of liquid on the surface, which could lead to further increase in CHF even beyond the prediction caused only by the wettability improvement. These micro/nanostructures with multiscale on heated surface induced more significant CHF enhancement than it based on the wettability effect, due to liquid spreadability.

  9. Texture Of Zircaloy-4 Result Of Beta-Quenching, Cold Rolling And Recrystallization

    International Nuclear Information System (INIS)

    Futichah; Sulistioso

    1998-01-01

    Differences of crystallographic texture of zircaloy-4 plate depends on cold working and heat treatment.To determine the change of zircaloy-4 textures, the solid solution treatment process at beta phase which was followed by quenching on water was employed for this sample. The next step was cold rolling until deformation epsilon = 1.62. The specimens were recrystallized at 750 o C, for 2 hours. The result of beta-quench gave a spread and different orientations and the main orientation occurred at (0001)[1010] and (0001)[1120]. Result of cold rolling with epsilon = 1.39 and epsilon 1.62 is the deformation texture at the main orientation of (0001)[1010] with the angle of inclination was around 38 o. However, the result of Recrystallization process on 750 o C for 2 hours gave annealing textures with orientations of (0001)[1120]. It means that the recrystallization process of zircaloy-4 plate can not remove the deformation textures, but can change the crystallographic orientation

  10. Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

    International Nuclear Information System (INIS)

    Roy, R.B.

    1963-12-01

    Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {1010} 1/3 type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {1010} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is ∼ 65 ergs/cm 2 . Calculations show that the equilibrium separation of partials is ∼ 60 A and a stress as high as 19x10 -3 μ acting along {0010} direction is needed to separate them. It has been suggested that O 2 and N 2 in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening

  11. Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

    International Nuclear Information System (INIS)

    Stout, R.B.

    2001-01-01

    An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

  12. Oxidation of Zircaloy-4 under limited steam supply at 1000 and 13000C

    International Nuclear Information System (INIS)

    Uetsuka, H.

    1984-12-01

    With the view of examining the oxidation behavior of Zircaloy-4 under limited steam supply occurring in severe accidents of LWRs, Zircaloy-4 cladding specimens were examined at the isothermal oxidation temperatures of 1000 and 1300 0 C under a steam atmosphere, flowing at a reduced and constant rate in the range of 3proportional170 mg/cm 2 xmin. The effect of steam starvation, which was restricted to very low levels of steam supply rate, was observed at the two examined temperatures. And the critical supply rate of steam starvation was evaluated to be 13 and 20 mg/cm 2 xmin for the oxidation at 1000 and 1300 0 C, respectively. Variation of the oxidation duration between 2 and 60 min at 1000 0 C allowed to compare the reaction kinetics for three different rates of steam supply. The short-term results confirmed the reduced reaction rates for the lower steam supplies. At the longer times, however, a clear trend towards linear kinetics was observed for the lower supplies. This can be interpreted as the result of earlier breakaway transition under limited steam supply. In the test at 1300 0 C, an acceleration of the oxidation rate was measured for the specified steam supply rate between 20 and 60 mg/cm 2 xmin. This related strongly with high hydrogen concentration in the atmosphere. Hydrogen blanketing - the retarding effect of hydrogen on Zircaloy oxidation - was not identified in the examined temperature range. (orig./HP) [de

  13. Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B. [California Univ., Livermore, CA (United States). Lawrence Livermore National Lab

    2001-07-01

    An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

  14. Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R B

    1963-12-15

    Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {l_brace}1010{r_brace} 1/3 <1210> type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {l_brace}1010{r_brace} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is {approx} 65 ergs/cm{sup 2}. Calculations show that the equilibrium separation of partials is {approx} 60 A and a stress as high as 19x10{sup -3} {mu} acting along {l_brace}0010{r_brace} direction is needed to separate them. It has been suggested that O{sub 2} and N{sub 2} in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening.

  15. The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Swam, L.F.P.; Shann, S.H.

    1991-01-01

    This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

  16. Influence of impurities on the ignition, combustion and explosion properties of Zircaloy filings

    International Nuclear Information System (INIS)

    Muenzel, H.; Praetorius, R.

    1990-11-01

    The influence of solid substances (e.g. UO 2 , MoO 3 , KNO 3 ) and liquids (e.g. water, nitric acid) on the behavior of Zircaloy filings was investigated. The addition of solid substances as well as liquids increases the ignition temperature. Samples with more than 50% water cannot be ignited (except with KCl solutions). With solid impurities added two modes of combustion are observed with propagation velocities of about 1 and >4 cm/s, respectively. The velocity depends on the heat capacity of the sample. In the presence of water the velocity increases by about two orders of magnitude. The maximum pressure observed in dust explosions in the presence of solid impurities depends on the heat capacity and the amount of Zircaloy burnt but not on the chemical properties of the added substances. The maximum pressure can be higher than 20 bar if water or nitric acid are added. With the proposed models and few additional experimental measurements it is possible to predict the behavior of other Zircaloy filings. (orig.) With 32 refs., 20 tabs., 91 figs [de

  17. Variation in the strain anisotropy of Zircaloy with temperature and strain

    International Nuclear Information System (INIS)

    Hindle, E.D.; Worswick, D.

    1984-04-01

    Strain anisotropy was investigated at temperatures in the range 293 to 1117K in circular tensile specimens prepared from rolled Zircaloy-2 plate so that their tensile axes were parallel to and transverse to the rolling direction. The strain anisotropy factor for both types of specimen increased markedly in the high alpha phase region above 923K reaching a maximum at circa 1070K. Above this temperature in the alpha-plus-beta phase region the strain anisotropy decreased rapidly as the proportion of beta phase increased and was almost non-existent at 1173K. The strain anisotropy was markedly strain dependent, particularly in the high alpha phase region. The study was extended to Zircaloy-4 pressurized water reactor (PWR) 17 x 17 type fuel rod tubing specimens which were strained under biaxial conditions using cooling conditions which promoted uniform diametral strain over most of their lengths (circa 250 mm). In these circumstances the strain anisotropy is manifest by a reduction in length. Measurement of this change along with that in diameter and wall thickness produced data from which the strain anisotropy factor was calculated. The results, although influenced by additional factors discussed in the paper, were similar to those observed in the uniaxial Zircaloy-2 tensile tests. (author)

  18. The steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Bera, P.C.; Clendening, W.R.

    1978-11-01

    The steady-state creep rates of as-received Zircaloy-4 fuel cladding have been determined in the α-Zr phase (940 -6 and 10 -3 s -1 were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law - Arrhenius equation, the creep rate for α-phase Zircaloy-4 is given by: epsilon sub(ss) = 2000σ sup(5.32) exp (-284 600/kT) s -1 and for the β-phase Zircaloy-4 is given by: epsilon sub(ss) = 8.1σ sup(3.79) exp (-142 300/kT) s -1 . For both the α-Zr and β-Zr phases, the activation energies for creep are in agreement with those for self-diffusion of zirconium and the rate-controlling mechanism is attributed to dislocation climb. Because of the scarcity of data, it is not possible to determine the rate equation unambiguously, nor to identify the mechanism for creep in the mixed α + β phase region. (author)

  19. Variation in the strain anisotropy of Zircaloy with temperature and strain

    International Nuclear Information System (INIS)

    Hindle, E.D.; Worswick, D.

    1984-01-01

    The strong crystallographic texture which is developed during the fabrication of zirconium-based alloys causes pronounced anisotropy in their mechanical properties, particularly deformation. The tendency for circular-section tension specimens with a high concentration of basal poles in one direction to become elliptical when deformed in tension has been used in this study to provide quantitative data on the effects of both strain and temperature on strain anisotropy. Tension tests were carried out over a temperature range of 293 to 1193 K on specimens machined from Zircaloy-2 plate. The strain anisotropy was found to increase markedly at temperatures over 923 K, reaching a maximum in the region of 1070 K. The strain anisotropy increased with increasing strain in this temperature region. The study was extended to Zircaloy-4 pressurized-water reactor fuel cladding by carrying out tube swelling tests and evaluating the axial deformation produced. Although scatter in the test results was higher than that exhibited in the tension tests, the general trend in the data was similar. The effects of the strain anisotropy observed are discussed in relation to the effects of temperature on the ductility of Zircaloy fuel cladding tubes during postulated largebreak loss-of-coolant accidents

  20. A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

    International Nuclear Information System (INIS)

    Polley, M.V.; Evans, H.E.

    1993-08-01

    High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

  1. Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

    International Nuclear Information System (INIS)

    Griger, Agnes

    2004-01-01

    The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

  2. MICROSTRUCTURE AND MECHANICAL STRENGTH OF SURFACE ODS TREATED ZIRCALOY-4 SHEET USING LASER BEAM SCANNING

    Directory of Open Access Journals (Sweden)

    HYUN-GIL KIM

    2014-08-01

    Full Text Available The surface modification of engineering materials by laser beam scanning (LBS allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and Y2O3 particles of 10 μm were selected for ODS treatment using LBS. Through the LBS method, the Y2O3 particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at 500°C was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive Y2O3 particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

  3. Charpy impact test of oxidized and hydrogenated zircaloy using a thin strip specimen

    International Nuclear Information System (INIS)

    Otsuka, Teppei; Hashizume, Kenichi; Sugisaki, Masayasu

    2004-01-01

    The impact properties of an oxidized and a hydrogenated Zircaloy have been studied with an instrumented Charpy machine by using a strip Charpy V-notch specimen (1 mm thick by 4mm wide). Fracture processes such as crack initiation and propagation were examined using load-displacement curves obtained in this study. In the case of the hydrogenated specimen containing preferentially oriented hydrides, an appreciable decrease in the absorbed energy was observed in the crack propagation rather than in the crack initiation. From results of fractographs of the specimen, it was suggested that the reduction of the crack propagation energy of hydrogenated specimen could be attributed to the change of the stress state in the Zircaloy matrix, which was caused by the fracture of hydride in the inner part of specimen. In the case of the specimen oxidized at 973k for 60 min, on which an oxide layer (4 μm in thickness) and oxygen incursion layer (4μm) were formed, the surface layers affected the crack initiation process. The growing oxygen incursion layer, in particular, resulted in the constraint of plastic deformation of the Zircaloy matrix not only in the crack initiation but also in the crack propagation as its thickness increased. (author)

  4. Effect of annealing temperature on the mechanical properties of zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Beauregard, R.J.; Clevinger, G.S.; Murty, K.L.

    1977-01-01

    The mechanical properties of zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. Hoop creep characteristics of zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The critical annealing temperature at which a minimum creep strain occurs decreases as the applied stress increases. An additional test at 700F and 30 ksi hoop stress was conducted to demonstrate that the critical annealing temperature is essentially independent of the test temperature. Plausible explanations based on differing substructures developed in cold-worked stress-relieved material are forwarded. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. R-parameters were determined from in situ transverse strain gage measurements in uniaxial tensile tests. P-parameters were calculated from uniaxial test data (R and yield stress) and hoop yield stress determined in biaxial, closed-end internal pressurization tests

  5. The effect of stimulated fission products on the structure and the mechanical properties of zircaloy

    International Nuclear Information System (INIS)

    Holub, F.

    1982-01-01

    The objective of investigation was to study the long-term effects of individual simulated fission products on the mechanical properties and the structure of Zircaloy. Tensile Test specimens of Zircaloy were annealed with important simulated fission products at 350 0 C up to 10,000 hours and at higher temperatures (500, 700 0 C) up to 2,000 hours. The principal methods of investigation on annealed Zircaloy specimens were tension tests at room temperature and at 400 0 C, scanning electron microscopy and microprobe technique, X-ray diffraction, X-ray fluorescence, optical metallography. The action of fission products at normal temperatures of reactor operation will give rise to a small enhancement of strength and a small drop of ductility of the fuel cladding material only. At high fuel pin temperatures which may be realized under abnormal operation conditions, some of the fission products potentially will produce detrimental consequences on the integrity of fuel pins. The most effective fission products will be: lanthanum oxide, followed by the earth alkaline oxides and the other rare earth oxides, molybdenum, iodine and cadmium

  6. Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

    International Nuclear Information System (INIS)

    Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

    1983-01-01

    To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

  7. For the world's best cladding tubes, ten years of progress by Zircaloy Special Committee of JAPCO

    International Nuclear Information System (INIS)

    Mishima, Yoshitsugu

    1982-01-01

    The zircaloy special committee was organized in 1971 for the purpose of planning the trial use of two nuclear fuel assemblies for which Japan-made cladding tubes were to be used, for a BWR. Now, seven years later, these two fuel assemblies have completed their service life, and have been submitted to post-irradiation examination after cooling for a year. Zircaloy tubes have been produced by Sumitomo Metal Industries, Ltd., and Kobe Steel, Ltd., and more than ten years have elapsed since wholly Japan-made zircaloy cladding tubes were used for reloading fuel elements for the Japan Power Demonstration Reactor. In this report, the history, progress and significance of the works performed by the committee are summarized. The LWR fuel elements made in Japan have attained the highest performance in the world as the leak has been scarce, and the works of the committee is one of the pioneering activities in the development of LWR fuel technology. The situation for starting the committee, the activity of the committee during ten years, the significance and outcome of the committee activity are reported. (Kako, I.)

  8. Les mots du jazz

    OpenAIRE

    Roueff, Olivier

    2007-01-01

    L’ouvrage d’André Schaeffner constitue la première analyse savante du jazz (1926). Il a marqué une étape importante dans le processus de réinvention du jazz en France en contribuant notamment, par sa réception et les polémiques qu’il a suscitées, à transformer l’identification du jazz d’une musique « américaine » à une musique « noire-américaine » (c’est-à-dire aux « racines » africaines). Les analyses proposées dans cet ouvrage, alors qu’elles désignaient des musiques que la critique de jazz...

  9. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

    1983-12-01

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

  10. fibrosarcome du larynx

    African Journals Online (AJOL)

    pie du lit tumoral est employée comme complément thé- rapeutique [9] alors que la chimiothérapie est générale- ment indiquée dans les formes métastatiques. Le pronos- tic dépend essentiellement du degré de différentiation his- tologique. En fait, le fibrosarcome bien différencié est caractérisé par la fréquence de récidive ...

  11. du Chott Marouane

    African Journals Online (AJOL)

    plancton de 90 µm de vide de maille. Ils ont été conservés dans du formol à 5%. L'identification de l'espèce est basée sur des critères morphologiques [20]: la forme des furcas, les lobes frontaux des antennes des mâles, de l'organe copulateur (pénis) et du sac ovigère. Le comptage des soies furcales a été réalisé. L'étude ...

  12. Automatisation en flux continu du dosage enzymatique de l'acide malique des vins

    Directory of Open Access Journals (Sweden)

    Aline Lonvaud-Funel

    1980-12-01

    The authors report a continuous flow technic for enzymatic analysis of malic acid in wines. In order to minimize the effect of interfering compounds, the NADH content is read at 570 nm. This method preserve the specificity and sensibility of the enzymatic manual assay.

  13. Influence de la nature et du dosage en fibres sur le comportement ...

    African Journals Online (AJOL)

    Influence of fiber type and volume contents on the physical and mechanical behaviour of ... obtained compared to reference concrete. ... Key-Words: sandcrete - granulated slag- polypropylene fibers - metallic fibers- mechanical behavior.

  14. Plutonium titration by controlled potential coulometry; Dosage du plutonium par coulometrie a potentiel impose

    Energy Technology Data Exchange (ETDEWEB)

    Leguay, N.

    2011-07-01

    The LAMMAN (Nuclear Materials Metrology Laboratory) is the support laboratory of the CETAMA (Analytical Method Committee), whose two main activities are developing analytic methods, and making and characterizing reference materials. The LAMMAN chose to develop the controlled potential coulometry because it is a very accurate analytical technique which allows the connection between the quantity of element electrolysed to the quantity of electricity measured thanks to the Faraday's law: it does not require the use of a chemical standard. This method was first used for the plutonium titration and was developed in the Materials Analysis and Metrology Laboratory (LAMM), for upgrading its performances and developing it to the titration of other actinides. The equipment and the material used were developed to allow the work in confined atmosphere (in a glove box), with all the restrictions involved. Plutonium standard solutions are used to qualify the method, and in particular to do titrations with an uncertainty better than 0.1 %. The present study allowed making a bibliographic research about controlled potential coulometry applied to the actinides (plutonium, uranium, neptunium, americium and curium). A full procedure was written to set all the steps of plutonium titration, from the preparation of samples to equipments storage. A method validation was done to check the full procedure, and the experimental conditions: working range, uncertainty, performance... Coulometric titration of the plutonium from pure solution (without interfering elements) was developed to the coulometric titration of the plutonium in presence of uranium, which allows to do accurate analyses for the analyses of some parts of the reprocessing of the spent nuclear fuel. The possibility of developing this method to other actinides than plutonium was highlighted thanks to voltammetric studies, like the coulometric titration of uranium with a working carbon electrode in sulphuric medium. (author)

  15. Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.

    1980-09-01

    Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

  16. les cahiers du cread

    African Journals Online (AJOL)

    Our Journal “les cahiers du cread” is a quarterly economic review publishing original findings of empirical research and theoretical debates on fields pertaining to our mission coverage (Macro Economics, Industrial Economics and Firms, Human Development & Social Economics, Agriculture & Environment). Other websites ...

  17. CONCEPTUALISATION ONTOLOGIQUE DE LA REPRÉSENTATION DU COMBAT CONCEPTUALISATION ONTOLOGIQUE DE LA REPRÉSENTATION DU COMBAT

    Directory of Open Access Journals (Sweden)

    Sylvain Rheault

    2012-10-01

    Full Text Available En adoptant une perspective existentielle, on peut tenter d'expliquer la représentation du combat au moyen de concepts comme les positions de SOI et de L'AUTRE ainsi que les statuts de l'ÊTRE et de la CHOSE, qui, une fois combinés, définissent des domaines ontologiques. Le combat devient alors, conceptuellement, l'action de forcer une conscience à passer d'un domaine à un autre. On observe qu'en modifiant l'intensité des positions (polarisation, des statuts (hiérarchisation et des actions (dosage, on peut expliquer les variations possibles des représentations du combat. Il restera à valider la pertinence de ces concepts en multipliant les analyses.En adoptant une perspective existentielle, on peut tenter d'expliquer la représentation du combat au moyen de concepts comme les positions de SOI et de L'AUTRE ainsi que les statuts de l'ÊTRE et de la CHOSE, qui, une fois combinés, définissent des domaines ontologiques. Le combat devient alors, conceptuellement, l'action de forcer une conscience à passer d'un domaine à un autre. On observe qu'en modifiant l'intensité des positions (polarisation, des statuts (hiérarchisation et des actions (dosage, on peut expliquer les variations possibles des représentations du combat. Il restera à valider la pertinence de ces concepts en multipliant les analyses.

  18. Dosage of DTPA administration by inhalation

    International Nuclear Information System (INIS)

    Koizumi, Akira; Fukuda, Satoshi; Yamada, Yuji; Iida, Haruzo; Shimo, Michikuni

    2000-01-01

    The administration of DTPA by inhalation was examined as an emergency medical treatment. In order to estimate the practical dosage to the human, an accurate model of the human air way was connected to a anesthetizer and respiration was simulated. Ca-DTPA, aerosolized by an ultra-sonic nebulizer, was administered by inhalation to the model. For the experiments, the respiratory volume (tidal volume) and the respiration rate was 12 per minute. Irrigation water from the model of larynx and mouth, and the air filter were collected and measured by chelate titration in order to determine the quantity of aerosolized DTPA and the amount deposited on the trachea and lang. The results indicated that the quantity of aerosolized DTPA varied with dilution of the DTPA solution in a ample. It was found that a 3 time dilution was the most practical and that 73 mg of DTPA per minute could be aerosolized. Furthermore, the results indicated that 46% of the aerosolized DTPA was taken in through inhalation and that 26% of DTPA was deposited in the trachea and lung. These results suggest that in practical application in the emergency medical treatment, 15 minutes of inhalation could delivered to approximately 500 mg of DTPA, and 130 mg could be delivered to the trachea and lung. It is considered that these quantity are enough amount to increase the effects of radioactive nuclides from the body, comparing with the recommended dosage for injection administration. (author)

  19. The characteristics of novel dosage forms

    Directory of Open Access Journals (Sweden)

    Milić-Aškrabić Jela

    2003-01-01

    Full Text Available The objective of pharmaceutical-technological development is to find a procedure of transforming an active substance (a drug into a drug dosage form which is not only acceptable for application, but also enables the active substance to be released following administration, pursuant to therapy objectives. The aim is that the concentration of the active substance in the action location rapidly reaches a therapeutic level and maintains an approximately constant level in the course of a particular time, according to the established therapeutic goal. The primary objective is to present the active ingredient (drug in the form and concentration/quantity that enables the corresponding therapeutic response, i.e. to control the site and rate of medicinal substance release from the drug, as well as the rate at which it reaches the membranes and surfaces to which it is absorbed, while applying a common method of administration. The procedures used to achieve this goal are becoming highly complex and demanding and are aiming at sophisticated drug delivery systems and functional packaging material. Development from the existing drug molecule, through the conventional drug dosage form, to a new system of drug "delivery" (novel delivery system, can improve the drug (active substance characteristics significantly in view of compliance (acceptability by the patient, safety and efficiency. The paper presents an overview of the most important examples of pharmaceutical forms with controlled release and advanced drug "carriers".

  20. Comparaison du filtre adaptatif RIF et du filtre a base de reseau de ...

    African Journals Online (AJOL)

    Comparaison du filtre adaptatif RIF et du filtre a base de reseau de neurones pour le filtrage du courant de reference pour la commande du filtre actif parallele. C Benachaiba, A Bassou, B Mazari ...

  1. Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

    International Nuclear Information System (INIS)

    Horwood, W.A.

    1981-07-01

    General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 90 0 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

  2. Cladding the inside surface of a 3 1/4 in. ID Zircaloy-2 pressure tube with 1S aluminum

    International Nuclear Information System (INIS)

    Watson, R.D.

    1966-09-01

    A hot-press sizing technique has been developed for cladding the inside surface of Zircaloy-2 pressure tubes with 1S aluminum. The process is performed in air with the Zircaloy-2 and aluminum at a temperature of approximately 950 o F. A controlled atmosphere is not required, either during preheating or while the cladding is being applied. Tubes 30 inches long and 3 1/4 inches ID have been coated with 1S aluminum in thicknesses ranging from 0.005 inches to more than 0.02 inches; tubes longer than 30 inches have not been attempted. The lining of aluminum is firmly attached to the Zircaloy-2 at all points in the tube but the bond strength varies considerably - from. 6500 to 28000 lbf/in 2 . This work is the subject of Canadian Patent Application No. 955,358 filed March 21, 1966. (author)

  3. Development of remote welding technology for nuclear fuel end capping (A study on the weldability of Zircaloy-4)

    Energy Technology Data Exchange (ETDEWEB)

    Kho, Jin Hyun; Sung, Ho Hyun; Hyun, Yong Kyu; Suh, Hee Kang [Korea University of Technology and Education, Cheonan (Korea)

    1998-03-01

    The integrity of nuclear fuel end cap welds is essential to the nuclear fuel performance and safety as well as the usability of power plant. The first aim of this project is to obtain experimental data on the nuclear fuel cladding materials of Zircaloy-4 with welding processes such as plasma arc, gas tungsten arc and laser beam welding. the data obtained in this study will be applicable to the nuclear fuel design, fabrication and nuclear fuel quality control. In addition, the welding processes applicable to the Zircaloy-4 welding were compared and contrasted. The weldability of Zircaloy-4 was evaluated from the metallurgical and mechanical standpoints. 88 refs., 57 figs., 16 tabs. (Author)

  4. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

    International Nuclear Information System (INIS)

    Pandarinathan, P.R.; Vasudevan, P.

    1980-01-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

  5. Le ministre du Commerce international du Canada rencontre des ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    17 juil. 2017 ... La promotion de l'entrepreneuriat, la façon dont le commerce peut profiter aux femmes et à leur famille, et la création d'emplois pour les plus vulnérables étaient au coeur de la discussion en table ronde du ministre du Commerce international du Canada, l'honorable François-Philippe Champagne, et des ...

  6. [Oral films as perspective dosage form].

    Science.gov (United States)

    Walicová, Veronika; Gajdziok, Jan

    Oral films, namely buccal mucoadhesive films and orodispersible films represent innovative formulations for administration of a wide range of drugs. Oral films show many advantageous properties and are intended for systemic drug delivery or for local treatment of the oral mucosa. In both cases, the film represents a thin layer, which could be intended to adhere to the oral mucosa by means of mucoadhesion; or to rapid dissolution and subsequent swallowing without the need of liquid intake, in the case of orodispersible films. Main constitutive excipients are film-forming polymers, which must in the case of mucoadhesive forms remain on the mucosa within the required time interval. Oral films are currently available on the pharmaceutical market and could compete with conventional oral dosage forms in the future. oral cavity oral films buccal mucoadhesive films orodispersible films film-forming polymers.

  7. Enalapril dosage in progressive chronic nephropathy

    DEFF Research Database (Denmark)

    Elung-Jensen, Thomas; Heisterberg, Jens; Sonne, Jesper

    2005-01-01

    OBJECTIVE: In chronic renal failure, clearance of enalapril is reduced. Hence, a renoprotective effect may be achieved with lower doses than conventionally used. Since marked inter-patient variation in concentrations of enalaprilat has been shown in patients with renal failure despite equivalent...... dosage of enalapril, a direct comparison of the effect of high versus low plasma concentrations of enalaprilat on the progression of renal failure was undertaken. METHODS: Forty patients with a median glomerular filtration rate (GFR) of 17 (6-35) ml/min/1.73 m2 were studied in an open-label, randomised...... intervals by the plasma clearance of 51Cr-EDTA, and the individual rates of progression of renal failure were calculated as the slope of GFR versus time plot. RESULTS: In the high-concentration group, the median enalaprilat trough concentration was 92.9 ng/ml (21.8-371.0 ng/ml) and in the low...

  8. Microstructure and crystallographic texture evolution during TIG welding of zircaloy-2 material

    International Nuclear Information System (INIS)

    Jha, S.K.; Singh, R.P.; Singh, V.K.; Ramanathan, R.; Samjdar, I.; Srivastava, D.; Tewari, R.; Dey, G.K.

    2005-01-01

    Zirconium and its alloys are extensively used as structural materials in nuclear reactors, because of better neutron economy, good corrosion resistance in water and good mechanical properties at operating temperature. Zircaloy-2 and zircaloy-4 are widely used in both pressurized water reactors (PWR) and boiling water reactors (BWR) as fuel cladding materials and as calandria tube and pressure tube materials in pressurized heavy water reactors (PHWR). The satisfactory performance and the life of the reactor components depend mainly upon their mechanical properties, corrosion properties and dimensional stability in the reactor condition, which are strong function of metallurgical parameters such as microstructure and texture. Therefore, for best performance of the reactor components these parameters are optimized during their fabrication. The microstructure and texture of the zircaloy-2 components are expected to get modified during the welding of the components. In this study the evolution of the microstructure and texture has been investigated as a function of the welding parameters. Heat input was varied the current and welding time. A variety of analytical techniques have been applied for the study on microstructure and texture of the welds. Optical microscopy and electron microscopy were used to evaluate the detailed microstructure. X-ray diffraction (XRD) was used investigate the crystallographic textures among the base metal, heat affected zone and fusion zone. Particular attention was focused on the determination of microtexture in weld by using electron backscatter diffraction (EBSD) technique. After that, an effort was put to compare the results of X-ray macro-texture and EBS-microtexture. (author)

  9. Stress corrosion cracking of Zircaloy-4 in non-aqueous iodine solutions

    International Nuclear Information System (INIS)

    Gomez Sanchez, Andrea V.

    2006-01-01

    In the present work the susceptibility to intergranular attack and stress corrosion cracking of Zircaloy-4 in different iodine alcoholic solutions was studied. The influence of different variables such as the molecular weight of the alcohols, the water content of the solutions, the alcohol type (primary, secondary or tertiary) and the temperature was evaluated. To determine the susceptibility to stress corrosion cracking the slow strain rate technique was used. Specimens of Zircaloy-4 were also exposed between 0.5 and 300 hours to the solutions without applied stress to evaluate the susceptibility to intergranular attack. The electrochemical behavior of the material in the corrosive media was studied by potentiodynamic polarization tests. It was determined that the active species responsible for the stress corrosion cracking of Zircaloy-4 in iodine alcoholic solutions is a molecular complex between the alcohol and iodine. The intergranular attack precedes the 'true' stress corrosion cracking phenomenon (which is associated to the transgranular propagation of the crack) and it is controlled by the diffusion of the active specie to the tip of the crack. Water acts as inhibitor to intergranular attack. Except for methanolic solutions, the minimum water content necessary to inhibit stress corrosion cracking was determined. This critical water content decreases when increasing the molecular weight of the alcohol. An explanation for this behavior is proposed. The susceptibility to stress corrosion cracking also depends on the type of the alcohol used as solvent. The temperature dependence of the crack propagation rate is in agreement with a thermal activated process, and the activation energy is consistent with a process controlled by the volume diffusion of the active species. (author) [es

  10. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  11. The corrosion of zircaloy 2 in anaerobic synthetic cement pore solution

    International Nuclear Information System (INIS)

    Hansson, C.M.

    1984-12-01

    Measurements have been made of the corrosion rates of Zircaloy 2 tubes in anaerobic synthetic cement pore solution of pH 12.0-13.8. The samples were tested in the as-received condition by the polarization resistance technique using a Tafal constant of 52 mV/decade and, for all pH values, corrosion rates of 3.10 -5 A/m 2 (0.03 μm/yr) were determined. These corrosion currents are at the lower limit of the experimental detection range of the technique used. Some samples were then held at a low electrochemical potential, namely -1850 mV SCE, for several days but this treatment had only a minor effect on the behaviour of the Zircaloy: the value of corrosion rate was increased by a factor of 3 and the free potential was temporarily lowered but drifted towards more positive values after the applied potential was removed. Attempts were made to remove the passive film from the surface of the samples by electrochemical reduction. For practical, experimental reasons, this was not successful and, instead, the effect of removing the film by scratching the surface was investigated. At both the free potential and at applied cathodic potentials, an anodic current was detected immediately and the surface was scratched but, in all cases, the scratched area repassivated within a few seconds and the anodic corrosion current fell accordingly. Thus, it may be concluded that active corrosion of Zircaloy 2 in anaerobic concrete will not occur and, by comparison with measurements on steel, it is likely that the passive corrosion rates will be even lower in concrete than those measured in the synthetic pore solution. (Author)

  12. Autoclave corrosion of zircaloy-4 cladding samples in LiOH solutions

    International Nuclear Information System (INIS)

    Hermann, A.

    2010-03-01

    In reactor operation, pH of the cooling water is adjusted by addition of alkaline hydroxides, and LiOH has been found to be the most suitable one. The addition of LiOH above a certain concentration level (depending on temperature) increases the corrosion rate of zirconium and its alloys. Hydrogen pick-up by the metal is also increased, and this effect is used to produce hydrided specimens for different investigations using the corrosion reaction. At the Paul Scherrer Institute several projects were accomplished to investigate the influence of hydrogen in Zircaloy cladding on its mechanical properties. In order to produce hydrided specimens for comparison and for adjusting new equipment, Zircaloy tubing samples were hydrogen charged by autoclave corrosion in lithiated water. Results of the corrosion experiments are outlined in this publication. Because of the great variety of possible experimental parameters these results are still of interest for the scientific community. Autoclave corrosion was accomplished in 0.2 M or 0.5 M LiOH solution at a constant temperature of 340 o C and a pressure of 160 bar. The corrosion rate increases from 84 mg/(dm 2 d) in 0.2 M LiOH to 153 mg/(dm 2 d) in 0.5 M LiOH. The hydrogen pick-up fraction in 0.5 M LiOH amounts to 80%. In 0.5 M LiOH, Zircaloy tubing samples can be charged with ∼ 500 ppm hydrogen in about 40 hours. In the corrosion experiments described in this report a homogeneous distribution of hydrides should be expected (except very high hydride concentrations) because no temperature gradient exists through the tubing wall. Hydrogen stringers are homogeneously distributed with circumferential orientation (stress-relieved tubing samples). (author)

  13. Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4

    International Nuclear Information System (INIS)

    Thevenet, J.

    1964-01-01

    The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the φ = 340 ingot into φ = 220 billets, cutting into lengths and hot drilling at φ = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes (φ =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [fr

  14. Portage vaginal du streptocoque du groupe B chez la femme ...

    African Journals Online (AJOL)

    Introduction: le streptocoque du groupe B est le principal agent impliqué dans les infections materno-fœtales, les septicémies et les méningites du nouveau-né à terme. L'objectif est de déterminer le taux de portage maternel du streptocoque du groupe B (SGB) à terme. Méthodes: un prélèvement vaginal a été réalisé de ...

  15. Aux origines du monde

    CERN Multimedia

    2004-01-01

    "C'est l'histoire d'une aventure humaine, scientifique, international qui a vu le jour il y a cinquante ans, aux confins de la Suisse et du département de l'Ain. Le plus grand laboratoire de physique des particules du monde, le Cern, a été fondé en 1954. Les festivités organisées à l occasion de cet anniversaire connaîtront leur point d'orgue le 16 octobre prochain, avec portes-ouvertes, accueil de personallités et inauguration d'un monumnet spécifique, le Globe de l'innovation" (2 pages)

  16. CHOEUR DU CERN

    CERN Multimedia

    CHOEUR DU CERN

    2010-01-01

    Les répétitions du chœur du CERN reprendront le mercredi 15 septembre à 20.00 heures à l’amphithéâtre principal – bâtiment 500. Au programme la préparation de notre concert de Noël avec la Missa Brevis, KV115, de Léopold Mozart et de la musique de Noël d’Europe. Les personnes qui aiment chanter, notamment des sopranes et des ténors, sont les bienvenues. Pour tout contact s’adresser à : Baudouin Bleus - (tél.CERN 767 82 44) -(baudouin.bleus@cern.ch) ou Martin Gatehouse ( martin.gatehouse@wanadoo.fr) ou Jean-Paul Diss (jean-pauldiss@wanadoo.fr).  

  17. Hepatiques du Surinam

    NARCIS (Netherlands)

    Jovet-Ast, S.

    1957-01-01

    Il n’existe pas, actuellement, de catalogue des Hépatiques du Surinam. Les Hépatiques de ce pays restent très peu connues. Cependant, certaines ont attiré l’attention des Bryologues et ont été citées dans quelques ouvrages anciens ou récents. Je ne ferai pas ici une révision complète de ces

  18. (l.) Medik du Maroc

    African Journals Online (AJOL)

    PR BOKO

    Résumé. Dipcadi serotinum (L.) Medik, est une plante de la famille des Hyacinthaceae, elle est largement utilisée comme réchauffant et aussi pour combattre la jaunisse. Cette plante trouve une large utilisation par la population de la région côtière du Maroc. À notre connaissance l'huile essentielle de cette espèce n'a ...

  19. Cyclic softening in annealed Zircaloy-2: Role of edge dislocation dipoles and vacancies

    Science.gov (United States)

    Sudhakar Rao, G.; Singh, S. R.; Krsjak, Vladimir; Singh, Vakil

    2018-04-01

    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.

  20. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Science.gov (United States)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). 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.

  1. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    International Nuclear Information System (INIS)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). 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.

  2. Statistics of the acoustic emission signals parameters from Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Oliveto, Maria E.; Lopez Pumarega, Maria I.; Ruzzante, Jose E.

    2000-01-01

    Statistic analysis of acoustic emission signals parameters: amplitude, duration and risetime was carried out. CANDU type Zircaloy-4 fuel claddings were pressurized up to rupture, one set of five normal pieces and six with defects included, acoustic emission was used on-line. Amplitude and duration frequency distributions were fitted with lognormal distribution functions, and risetime with an exponential one. Using analysis of variance, acoustic emission was appropriated to distinguish between defective and non-defective subsets. Clusters analysis applied on mean values of acoustic emission signal parameters were not effective to distinguish two sets of fuel claddings studied. (author)

  3. Corrosion kinetic of 2 and 4 zircaloys in air at high temperatures

    International Nuclear Information System (INIS)

    Goncalves, A.C.; Goncalves, Z.C.

    1986-01-01

    The corrosion results of 2 and 4 zircaloys obtained in a thermal balance between 500 and 850 0 C are discussed based on the model of 'reduction of diffusion path'. The behaviour of both alloys has shown almost similar in this interval of temperature, proving that the corrosion is processed by an identical kinetic mechanism. It is still analysed the formation of superposed layer of porous oxide and the possible influence of the oxygen partial pressure in inversion velocities between 750 and 800 0 C. (Author) [pt

  4. A new method of residual stress distribution analysis for corroded Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Godlewski, J.; Cadalbert, R.

    1992-01-01

    An X-ray diffraction method of residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400degC under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. (author)

  5. Metallographic Study of the Isothermal Transformation of Beta Phase in Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Oestberg, G

    1960-06-15

    Observations of the structure of commercial zircaloy-2 have been made in the microscope showing that the high temperature beta phase is transformed isothermally at lower temperatures into alpha plus secondary precipitate. The alpha occurs mainly as Widmanstaetten plates developed by a shear mechanism. The secondary precipitate is formed from the beta - alpha structure at the phase boundary between these phases. This precipitation of particles of secondary phase occurs on account of a eutectoid reaction, alpha also being formed. A time-temperature transformation diagram has been constructed from the observations.

  6. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  7. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  8. Process development for fabrication of zircaloy- 4 of dissolver assembly for spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Tonpe, Sunil; Saibaba, N.; Jairaj, R.N.; Ravi Shankar, A.; Kamachi Mudali, U.; Raj, Baldev

    2010-01-01

    Spent fuel reprocessing for fast breeder reactor (FBR) requires a dissolver made of a material which has resistance to corrosion as the process involves Nitric Acid as the process medium. Various materials to achieve minimum corrosion rates have been tried for this operation. Particularly the focus was on the use of advanced materials with high performance (corrosion rate and product life) for high concentrations greater than 8 N and temperatures (boiling and vapour) of Nitric Acid employed in the dissolver unit. The different commercially available materials like SS316L , Pure Titanium, Ti - 5% Ta and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. As this is continuous process of evolution of new materials, it was decided to try out zircaloy - 4 as the material of construction for construction due to its excellent corrosion resistance properties in Nitric Acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate in shape. On accepting the challenge of fabrication of dissolver, NFC has made different fixtures for Electron Beam Welding and TIG Welding. Various trials were carried out for optimization of various operating parameter like beam current, Acceleration voltage, welding speed to get adequate weld penetration. Both EB welding and TIG welding process were standardized and qualified by carrying out a number of trials and testing these welds by various weld qualification procedures like radiography, Liquid dye penetrant testing etc. for different intricate weld geometries. All the welds were simulated with samples to optimize the weld parameters. Tests such as include metallographic (for microstructure and HAZ), mechanical (for weld strength) and chemical (material analysis for gases) were conducted and all the weld samples met the acceptable criteria. Finally the dissolver was made meeting stringent specifications. All the welds were checked

  9. Fatigue limit of Zircaloy-2 under variable one-directional tension and temperature 300 deg C

    International Nuclear Information System (INIS)

    Spasic, Z.; Simic, G.

    1968-11-01

    A vacuum chamber wad designed and constructed. It was suitable for study of materials at higher temperatures in vacuum or controlled atmospheres. Zircaloy-2 fatigue at 300 deg C in argon atmosphere was measured. Character of strain is variable one directional (A=1) tension. Obtained results are presented in tables and in the form of Veler's curve. The obtained fatigue limit was σ - 15 kp/mm 2 . The Locati method was allied as well and fatigue limit value obtained was 15,75 kp/mm 2 . Error calculated in reference to the previous value obtained by classical methods was 5% [sr

  10. Spectrophotometric determination of uranium traces in zircaloy-4 and zirconium sponge

    International Nuclear Information System (INIS)

    Correia, R.J.; Weber de D'Alessio, Ana; Zucal, R.H.

    1980-01-01

    The uranium contents of the zircaloy-4 which is used for the fabrication of the fuel cans for the PHWR Atucha and Embalse nuclear power stations must not exceed 3.ppM. A method was developed for performing that control, involving the separation of the uranium from its matrix by partition chromatography and its determination by spectrophotometry with Arsenazo (III). This method is applied within the range of 0.2 to 10 ppM, obtaining a relative standard deviation of 6% for U contents of 3 ppm. (M.E.L.) [es

  11. A new method for residual stress distribution - analysis of corroded zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Godlewski, J.; Cadalbert, R.

    1992-01-01

    An X-ray diffraction method for residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400 deg C under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as a function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. 12 refs., 5 figs., 1 tab

  12. Biaxial creep deformation of Zircaloy-4 in the high alpha phase temperature range

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The ballooning response of Zircaloy-4 fuel tubes during a postulated loss-of-coolant accident may be calculated from a knowledge of the thermal environment of the rods and the creep deformation characteristics of the cladding. In support of such calculations biaxial creep studies have been performed on fuel tubes supplied by Westinghouse, Wolverine and Sandvik of temperatures in the alpha phase range. This paper presents the results of an investigation of their respective creep behaviour which has resulted in the formulation of equations for use in LOCA fuel ballooning codes. (author)

  13. TEM study of microstructure in explosive welded joints between Zircaloy-4 and stainless steel

    International Nuclear Information System (INIS)

    Zhou Hairong; Zhou Bangxin

    1996-10-01

    The microstructure of explosive welded joints between Zircaloy-4 and 18/8 stainless steel has been investigated by transmission electron microscopy (TEM). The metallurgical bonding was achieved by combining effect of diffusion and local melting when the explosive parameters were selected correctly. The molten region which consists of amorphous and crystalline with hexagonal crystal structure is hard and brittle. But the welded joints can be pulled, bent and cold rolled without cracks formed on the bonding layer, so as the molten regions are small and distributed as isolated islands. (6 refs., 6 figs., 1 tab.)

  14. Influence du taux de balles de riz sur la resistance a la compression ...

    African Journals Online (AJOL)

    Le présent article a pour objet la recherche de l'influence du taux des balles de riz sur la résistance à la compression des briques en terre. Ceci permettra de définir le dosage optimum de balles de riz permettant d'avoir une meilleure résistance à la compression des briques en terre. La terre utilisée a une prédominance ...

  15. La mesure du danger

    CERN Document Server

    Manceron, Vanessa; Revet, Sandrine

    2014-01-01

    La mesure du danger permet d’explorer des dangers de nature aussi diverse que la délinquance, la pollution, l’écueil maritime, la maladie ou l’attaque sorcellaire, l’extinction d’espèces animales ou végétales, voire de la Planète tout entière. Au croisement de la sociologie, de l’anthropologie et de l’histoire, les différents articles analysent les pratiques concrètes de mesure pour tenter de comprendre ce qui se produit au cours de l’opération d’évaluation du danger sans préjuger de la nature de celui-ci. L’anthropologie a contribué à la réflexion sur l’infortune en s’intéressant aux temporalités de l’après : maladies, catastrophes, pandémies, etc. et en cherchant à rendre compte de l’expérience des victimes, de leur vie ordinaire bouleversée, de la recomposition du quotidien. Elle s’intéresse aussi aux autres types de mesures, les savoirs incorporés, qui reposent sur l’odorat, la vue ou le toucher et ceux qui ressortent d’une épistémologie « non ...

  16. Determination of the initial oxidation behavior of Zircaloy-4 by in-situ TEM

    International Nuclear Information System (INIS)

    Harlow, Wayne; Ghassemi, Hessam; Taheri, Mitra L.

    2016-01-01

    The corrosion behavior of Zircaloy-4 (Zry-4), specifically by oxidation, is a problem of great importance as this material is critical for current nuclear reactor cladding. The early formation behavior and structure of the oxide layer during oxidation was studied using in-situ TEM techniques that allowed for Zry-4 to be monitored during corrosion. These environmental exposure experiments were coupled with precession electron diffraction to identify and quantify the phases present in the samples before and after the oxidation. Following short-term, high temperature oxidation, the dominant phase was revealed to be monoclinic ZrO 2 in a columnar structure. These samples oxidized in-situ contained structures that correlated well with bulk Zry-4 subjected to autoclave treatment, which were used for comparison and validation of this technique. By using in-situ TEM the effect of microstructure features, such as grain boundaries, on oxidation behavior of an alloy can be studied. The technique presented herein holds the potential to be applied any alloy system to study these effects. - Highlights: • In-situ TEM was used to oxidize samples of Zircaloy-4. • Similar behavior was found in the in-situ oxidized and autoclave-oxidized samples. • Precession diffraction was used to characterize oxide phase and texture.

  17. A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

    1977-01-01

    Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

  18. An investigation of deformed microstructure and mechanical properties of Zircaloy-4 processed through multiaxial forging

    Energy Technology Data Exchange (ETDEWEB)

    Fuloria, Devasri; Nageswararao, P. [Department of Metallurgical and Materials Engineering & Centre of Nanotechnology, IIT Roorkee, Roorkee 247667 (India); Jayaganthan, R., E-mail: rjayafmt@iitr.ernet.in [Department of Metallurgical and Materials Engineering & Centre of Nanotechnology, IIT Roorkee, Roorkee 247667 (India); Department of Engineering Design, Indian Institute of Technology Madras, Chennai 600036 (India); Jha, S. [Nuclear Fuel Complex Limited, Hyderabad 501301 (India); Srivastava, D. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 40085 (India)

    2016-04-15

    In the present work, the mechanical behavior of Zircaloy-4 subjected to various deformation strains by multiaxial forging (MAF) at cryogenic temperature (CT) was investigated. The alloy was strained up to different number of cycles, viz., 6 cycles, 9 cycles, and 12 cycles at cumulative strains of 2.96, 4.44, and 5.91, respectively. The mechanical properties of the alloy were investigated by performing the universal tensile test and the Vickers hardness test. Both the test showed improvement in the ultimate tensile strength and hardness value by 51% and 26%, respectively, at the highest cumulative strain of 5.91. The electron backscattered diffraction (EBSD) measurement and transmission electron microscopy (TEM) were used for analyzing the deformed microstructure. The microstructures of the alloy underwent deformation at various cumulative strains/cycles showed grain refinement with the evolution of shear and twin bands that were highest for the alloy deformed at the highest number of cycles. The effective grain refinement was due to twins formation and their intersection, which led to the improvement in mechanical properties of the MAFed alloy, as observed in the present work. - Highlights: • Zircaloy-4 was subjected to MAF at cryogenic temperature. • Microstructural evolution was studied through EBSD and TEM. • Deformed microstructure was marked with various types of twinning and shear banding. • Twins formations are responsible for effective grain refinement and enhanced mechanical properties.

  19. Plastic deformation and fracture behavior of zircaloy-2 fuel cladding tubes under biaxial stress

    International Nuclear Information System (INIS)

    Maki, Hideo; Ooyama, Masatosi

    1975-01-01

    Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation. The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law. The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10 0 to 15 0 is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material. (auth.)

  20. Measurement of dose rate and estimation of beta activity in zircaloy hull drum

    International Nuclear Information System (INIS)

    Pandey, J.P.N.; Kumar, Pankaj; Shinde, A.M.; Purohit, R.G.; Sarkar, P.K.

    2012-01-01

    Fuel Reprocessing Plant is designed for the processing of spent fuel from reactor for the recovery of plutonium and uranium as PuO 2 and U 3 O 8 respectively. Zircaloy is used as cladding material of natural uranium fuel pins used in the reactors. In reprocessing plants chop and leach method is used to remove the zircaloy clad from the fuel matrix during Head End Treatment. Initially spent fuel bundles are chopped into pieces and collected in perforated baskets kept in dissolvers. All chopped pieces are dissolved in HNO 3 in the dissolvers followed by heating and boiling. Dissolved solutions are transferred to Filtrate Tank (FT) leaving behind un-dissolved zircoloy hull pieces in the dissolver baskets. Un-dissolved and almost dry hull pieces are transferred in hull drum from the dissolver baskets using the Hull Tilting Facility. Hull drums are made of stainless steel having 500 litre capacity and two third of its volume is filled with zircoloy pieces. Hull drums filled with hull pieces are loaded in Hull Removal Cask (HRC) and transported to SWMF (Solid Waste Management Facility) site for interim storage/disposal in tile holes. Hull pieces are high active solid wastes which contain significant amount of fission products. Radiation levels on hull drums are in the range of few hundreds of mGy/h which has high potential of external hazards if not handled properly. Therefore hull drums are handled remotely in specially designed lead shielded cask

  1. Influence of specimen design on the ductility of zircaloy cladding: Experiment and analysis

    International Nuclear Information System (INIS)

    Bates, D. W.; Majumdar, S.; Koss, D. A.; Motta, A. T.

    1999-01-01

    In a reactivity-initiated accident (RIA), a control rod ejection or drop causes a sudden increase in reactor power, which in turn deposits a large amount of energy into the fuel. The resulting thermal expansion and fission gas release loads the cladding into the plastic regime and may cause it to fail. In order to predict cladding survivability, there has been considerable interest and effort in supplementing integral WA tests with separate-effects ring tests of cladding tubes. Such tests can give one insight into failure mechanisms and measure relevant mechanical properties (such as yield strength, uniform elongation, uniaxial stress-strain curve, etc.), for use in computer codes that attempt to predict cladding response during an RIA. The accuracy of such model predictions obviously depends on appropriate and accurate failure data. This study concerns itself with the proper development of ring tensile tests that (i) are similar to the loading conditions present in an RIA, (ii) measure the relevant mechanical properties and (iii) provide insight regarding the influence of the strain paths on the failure mechanisms present if Zircaloy cladding. Based on both experiments and computational modeling, the authors investigate the failure of Zircaloy tubing as a function of specimen geometry, and discuss the limitations of certain ring-test geometries in yielding failure ductility data that are applicable to RIA situations

  2. Experimental investigation of strain, damage and failure of hydrided zircaloy-4 with various hydride orientations

    International Nuclear Information System (INIS)

    Racine, A; Catherine, C.S.; Cappelaere, C.; Bornert, M.; Caldemaison, D.

    2005-01-01

    This experimental investigation is devoted to the influence of the orientation of hydrides on the mechanical response of Zircaloy-4. Ring tensile tests are performed on unirradiated CWSR Zircaloy-4, charged with about 200 or 500wppm hydrogen. Hydrides are oriented either parallel ('tangential'), or perpendicular ('radial') to the circumferential tensile direction. Tangential hydrides are usually observed in cladding tubes, however, hydrides can be reoriented after cooling under stress to become radial and then trigger brittle behavior. In this investigation, we perform, 'macroscopic' or SEM in-situ tensile tests on smooth rings, at room temperature. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. The results lead to the following conclusions: neither the tensile stress-strain response nor the strain modes are affected by hydrogen content or hydride orientation, but the failure modes are. Indeed, only 200wppm radial hydrides embrittle Zy-4: sample fails in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases samples reach at least 750 MPa before failure, with ductile or brittle mode. (authors)

  3. Texture, morphology and deformation mechanisms in β-transformed Zircaloy-4

    International Nuclear Information System (INIS)

    Ciurchea, D.; Furtuna, I.; Todica, M.; Roth, M.

    1996-01-01

    The morphology of the β(bcc) transformed Zircaloy-4 may be treated as a lenticular-twinned martensite. The texture is a consequence of the degeneration of the left angle 0001 right angle α , left angle 1010 right angle α and left angle 1011 right angle α directions into left angle 110 right angle β directions. The crystallographic mechanisms implied in the accommodation of the microscopic Bain strain are (1010) left angle 1120 right angle prism slip, (1012) left angle 101 1 right angle twinning and (1011) left angle 1012 right angle twinning. This degeneration explains the 'parallel plate' and 'basketweave' morphologies observed by microscopy and the texture of the β transformed tube. The macroscopic Bain strain was calculated and agrees with the dimensional measurements. The deformation mechanisms of β transformed Zircaloy-4 are identified from the new texture and from deformation experiments as twinning and interplatelet glide. The interplatelet glide induces a fragile character of fracture in the 'parallel plate' morphology. (orig.)

  4. Determination of Oxygen in Zircaloy Surfaces by Means of Charged Particle Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzen, J; Brune, D

    1973-01-15

    Oxygen in zircaloy surfaces has been determined by means of charged particle activation analysis employing the following two reactions I. 16O (d, n) 17F ->(beta+decay) 17O Q = - 1.63 MeV; II. 16O (d, pgamma) 17O Q = + 1.05 MeV. The detection limits for oxygen in such surfaces has been investigated by measuring the promptly emitted 0.87 MeV gamma rays (reaction II) and also the 511 keV annihilation radiation which arises from beta-decay of 17F (reaction I). The correlation between the detection limit for oxygen in zircaloy, the particle energy and the surface thickness analyzed has been evaluated. At a deuteron energy of 3 MeV a detection limit of 0.7 x 10-7 g/cm2 was obtained from the measurement of the prompt gamma radiation arising from the second of these reactions. The analysis carried out by means of this technique is characterized by a high rapidity

  5. The effects of corrosion conditions and cold work on the nodular corrosion of zircaloy-4

    International Nuclear Information System (INIS)

    You, Gil Sung

    1992-02-01

    The nodular corrosion of Zircaloy-4 was investigated on the effects of corrosion conditions and cold work. Variation of steam pressures, heat-up environments and prefilms were considered and cold work effects were also studied. The corrosion rate of Zircaloy-4 was dependent on pressure between 1 and 100 atm and it followed the cubic law as W=16.85 x P 0.31 for plate specimens and W=12.69 x P 0.27 for tube specimens, where W is weight gain (mg/dm 2 ) and P is the steam pressure (atm). The environment variation in autoclave during heat-up period did not affect the early stage of nodular corrosion. The prefilm, which was formed at 500 .deg. C under 1 atm steam for 4 hours, restrained the formation of the initial small nodules. The oxide film formed under 1 atm steam showed no difference of electrical resistivity from the oxides formed under 100 atm steam pressure. Cold work specimens showed the higher resistivity against nodular corrosion than as-received specimens. The corrosion resistance arising from cold work seems to be due to the texture changes by the cold work. The results showed that cold work can affect the later stage of uniform corrosion and the early stage of nodular corrosion, namely, the nodule initiation stage

  6. Hydrides blister formation and induced embrittlement on zircaloy-4 cladding tubes in reactivity initiated conditions

    International Nuclear Information System (INIS)

    Hellouin-De-Menibus, A.

    2012-01-01

    Our aim is to study the cladding fracture with mechanical tests more representative of RIA conditions, taking into account the hydrides blisters, representative strain rates and stress states. To obtain hydride blisters, we developed a thermodiffusion setup that reproduces blister growth in reactor conditions. By metallography, nano-hardness, XRD and ERDA, we showed that they are constituted by 80% to 100% of δ hydrides in a Zircaloy-4 matrix, and that the zirconium beneath has some radially oriented hydrides. We modeled the blister growth kinetics taking into account the hysteresis of the hydrogen solubility limit and defined the thermal gradient threshold for blister growth. The modeling of the dilatometric behavior of hydrided zirconium indicates the important role of the material crystallographic texture, which could explain differences in the blister shape. Mechanical tests monitored with an infrared camera showed that significant local heating occurred at strain rates higher than 0.1/s. In parallel, the Expansion Due to Compression test was optimized to increase the bi-axiality level from uniaxial stress to plane strain (HB-EDC and VHB-EDC tests). This increase in loading bi-axiality lowers greatly the fracture strain at 25 C and 350 C only in homogeneous material without blister. Eventually, the ductility decrease of unirradiated Zircaloy-4 cladding tube in function of the blister depth was quantified. (author) [fr

  7. NIRVANA, a high-temperature creep model for Zircaloy fuel sheathing

    International Nuclear Information System (INIS)

    Sills, H.E.; Holt, R.A.

    1979-05-01

    We have developed a multi-component model to describe the transient plastic deformation of Zircaloy fuel sheathing during high-temperature transients. From deformation maps we identify three deformation mechanisms which, in principle, occur in all three phase fields of Zircaloy (α, α+β, β): diffusional creep, dislocation creep, and athermal strian. A strain component occurring during the α → β transformation is also identified. Microstructural changes which alter deformation rates -grain structure, recrystallization, phase transformation -are accounted for. The individual components of the model represent known metallurgical phenomena. The combined model gives excellent agreement with transient test data from 700-1800 K, a range of heating rates from 0-100 K.s -1 , and a range of strain rates from 10 -5 to 10 -1 .s -1 . To enable comparison with available data the transient creep model was combined with an axially uniform, thin-walled tube representation having anisotropic material properties. The resulting computer code, NIRVANA provides facilities for simulating uniaxial and biaxial tube tests over specified stress/temperature histories. (author)

  8. Strengthening of Zircaloy-4 using Oxide Particles by Laser Beam Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Kim, Hyun-Gil; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Oxide particles such as Y{sub 2}O{sub 3} and CeO{sub 2} were dispersed homogeneously in a Zircaloy-4 plate surface using an LBS method. From the tensile test at 380 .deg. C, the strength of laser ODS alloying on the Zircaloy-4 sheet was increased more than 50% when compared to the initial state of the sheet, although the ODS alloyed layer was less than 20% of the specimen thickness. This technology showed a good opportunity to increase the strength without major changes in the substrates of zirconium-based alloys. Accident tolerant fuel (ATF) cladding is being developed globally after the Fukushima accident with the demands for the nuclear fuel having higher safety at normal operation conditions as well as even in a severe accident conditions. Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures.

  9. Neutron irradiation effects on intermetallic precipitates in Zircaloy as a function of fluence

    International Nuclear Information System (INIS)

    Etoh, Y.; Shimada, S.

    1993-01-01

    Intermetallic precipitates in Zircaloy-2 and -4, recrystallized at the α-phase temperature, have been examined using analytical electron microscopy. The specimens were irradiated in BWRs up to a fast neutron fluence of 1.4x10 26 n/m 2 (E>1 MeV). Neutron irradiation induces a crystalline-to-amorphous transition, depleting Fe in the amorphous phase of Zr(Fe, Cr) 2 precipitates in the alloys. Amorphization starts from the periphery of the precipitates and all of them are totally amorphized at higher fluences than 1.2x10 26 n/m 2 . The width of the Fe-depleted zone increases in proportion to the 0.45 power of fluence. This result indicates that diffusion of Fe is the rate-controlling process for Fe depletion in Zr(Fe, Cr) 2 precipitates. Dissolution of Zr 2 (Fe, Ni) precipitates in Zircaloy-2 occurs during neutron irradiation. At a high fluence, such as 1.2x10 26 n/m 2 , Zr 2 (Fe, Ni) precipitates are almost completely dissolved into the matrix and the dissolution rate of Fe is faster than that of Ni. (orig.)

  10. Fatigue testing on samples from Zircaloy-4 tubes type SEU-43

    International Nuclear Information System (INIS)

    Olaru, V.; Ionescu, V.; Nitu, A.; Ionescu, D.; Voicu, F.

    2016-01-01

    The paper presents the testing of samples worked from Zicaloy-4 tubes (as-received.. metallurgical state), utilized in the composition of the CANDU SEU-43 fuel bundle. These tests are intended to simulate their behaviour in a power cycling process inside the reactor. The testing process is of low cycle fatigue type, done outside of the reactor, on ''C-ring'' samples, cut along the transversal direction. These samples are tested at 1%, 2% and 3% amplitude deformation, at room temperature. The calibration curves for both types of tube (small and big diameter) are determined by using the finite element analyses with the ANSYS computer code. The cycling test results are in the form of a fatigue life curve (N-e) for zircaloy-4 used in the SEU-43 fuel bundle. The curve is determined by the experimental dependency between the number of cycles to fracture and the deformation amplitude. The low cycle fatigue mechanical tests done at room temperature together with electronic microscopy analyses have reflected the characteristic behaviour of the zircaloy-4 metal in the given environment conditions. (authors)

  11. Chemical inhomogeneity populations in various zircaloy claddings and their association with SCC and corrosion resistance

    International Nuclear Information System (INIS)

    Tasooji, A.; Miller, A.K.; Cheung, T.Y.; Brooks, M.; Santucci, J.

    1987-01-01

    A technique has been developed that permits detection and characterization of sparsely distributed chemical inhomogeneities in Zircaloy. These inhomogeneities have previously been observed at the origins of iodine stress-corrosion cracks but are not detectable by, for example, simple scanning electron microscopy (SEM) examination. The technique uses radioactive iodine to ''label'' the chemical inhomogeneities, autoradiography to detect their locations, and SEM and energy-dispersive X-ray analysis (EDAX) to further characterize them. Large areas of surface have been surveyed and statistically meaningful populations of chemical inhomogeneities measured for five different lots of Zircaloy cladding. Inner surfaces and cladding cross-sectional surfaces have been studied. There are clear differences in chemical inhomogeneity size distribution and composition between the various claddings. For three of the claddings characterized in this work, the previously measured stress-corrosion cracking (SCC) threshold stresses correlate well (inversely) with the new data on their average chemical inhomogeneity sizes. Of special interest is the fact that the most SCC-resistant cladding contains far fewer iron-bearing inhomogeneities than the other claddings

  12. Thermomechanical treatment of β-treated Zircaloy-4 within the upper α-range

    International Nuclear Information System (INIS)

    Chauvy, C.

    2004-09-01

    Zircaloy-4 is a Zr base alloy mainly used for nuclear applications. This study deals with its behaviour during forming, with a special interest for physical mechanisms acting in the upper α-range. The β-treated Zircaloy-4 is first described in terms of microstructure and texture. The α plates are organised in colonies with alternating crystallographic orientations. The Burgers relationships have also been checked. The mechanical behaviour shows two distinct domains (with or without hardening). This could be linked to interactions between the solutes (Sn, O) and mobile dislocations. The evolution of crystallographic texture is characterised by X-ray diffraction. At 550 C, twinning is shown to be the main deformation mechanism under specific experimental conditions. Globularization of the initial lamellar structure is identified as a continuous dynamic recrystallization process, more efficient at higher temperature. High strains are necessary to achieve this phenomenon but meta-dynamic recrystallization can also be used to obtain an equiaxed microstructure after limited strains. (author)

  13. Parametric studies of cutting zircaloy-2 sheets with a laser beam

    International Nuclear Information System (INIS)

    Ghosh, S.; Badgujar, B.P.; Goswami, G.L.

    1996-01-01

    The highly reactive and pyrophoric nature of zirconium alloys limits the use of conventional thermal sources (e.g., plasma arc cutting, oxygen flame cutting, etc.) for the cutting and drilling of these alloys. In this context, a highly coherent laser beam provides a good alternative for the cutting and drilling. In the present paper, laser beam cutting of zircaloy-2 sheets of 1.1 mm and 0.74 mm thickness is performed using a 300 W average power pulsed Nd:YAG laser. Pulse energy, pulse repetition rate, nozzle gap, gas pressure and cutting speed were varied to give different laser cutting conditions. Metallographic study of the cut surfaces showed the presence of transformed beta phase in the heat affected zone (HAZ) near the cut surface. The microhardness value across the cut surface was also measured. It showed a gradual increase in microhardness from the base metal (160 VHN) towards the HAZ having a maximum value of 365 VHN. The results of parametric studies of the cutting indicated that, with proper selection of process parameters, very narrow cuts can be easily made in zircaloy-2 using a pulsed Nd:YAG laser with a saving in material and at a much faster rate than alternative processes such as plasma arc cutting and oxygen flame cutting

  14. Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Azer, N.; Wong, H.

    1994-01-01

    The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

  15. Aerosol material release rates from zircaloy-4 at temperatures from 2000 to 22000C

    International Nuclear Information System (INIS)

    Mulpuru, S.R.; Wren, D.J.; Rondeau, R.K.

    1987-01-01

    During some postulated severe accidents involving loss of coolant and loss of emergency coolant injection, the temperatures in a CANDU reactor fuel channel become high enough to cause failure and melting of the Zircaloy fuel cladding. At such high temperatures, vapors of fission products and structural (fuel and cladding) materials will be released into the coolant steam and hydrogen mixture. These vapors will condense as cooler conditions are encountered downstream. The vapors from structural materials are relatively involatile; therefore, they will condense readily into aerosol particles. These particles, in turn, will provide sites for the condensation of the more volatile fission products. The aerosol transport of fission products in the primary heat transport system (PHTS) will thus be influenced by the structural material release rates. As part of an ongoing program to develop predictive tools for aerosol and associated fission product transport through the PHTS, experiments have been conducted to measure the vapor mass release rates of the alloying elements from Zircaloy-4 at high temperatures. The paper presents the results and analysis of these experiments

  16. The accelerated oxidation of zircaloy-4 at 700∼900 .deg. C in high pressure steam

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, K. H.

    1999-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The specimens used in experiments are commercially available Zircaloy-4 used in Kori nuclear power plants. All the measurements were done at 700∼900 .deg. C in steam. Pressure effects were noticed. The oxide thickness was much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. The enhancement of oxide growth rate at 700∼900 .deg. C in high pressure steam is approximately propotion to the power of 1.0∼1.6 of the ratio of experimental steam pressure to critical steam pressure. There is a critical steam pressure above that the oxidation rate enhances. The critical steam pressure was measured as 30∼40 bar. The enhanced oxidation seems from the oxide cracking due to the tetragonal to monoclinic phase transformation at high pressure steam

  17. Electrochemical corrosion of Zircaloy-2 under PWR water chemistry but at room temperature

    International Nuclear Information System (INIS)

    Waheed, Abdel-Aziz Fahmy; Kandil, Abdel-Hakim Taha; Hamed, Hani M.

    2016-01-01

    Highlights: • There is no simple relation between the corrosion rate and LiOH concentration. • At low concentration, 100 ppm Li, an increase of the rate is due to the pH impact. • LiOH in concentrated solution led to accelerated corrosion by pH effect and porosity. • Boron abates the lithium effect by pH neutralizing and participation in the corrosion. - Abstract: Electrochemical corrosion of Zircaloy-2 was tested at room temperature in lithium hydroxide (LiOH) concentrations that ranged from 2.2 to 7000 ppm and boric acid (H 3 BO 3 ) concentrations that ranged from 50 to 4000 ppm. Following the corrosion experiments, the oxide films of specimens were examined by SEM to examine the oxide existence. LiOH concentrations as high as 1 M (7000-ppm lithium) can lead to significantly increased electrochemical corrosion rate. It is suggested that the accelerated corrosion in concentrated solution is caused by the synergetic effect of LiOH, pH and porosity generation. In solutions containing 100 ppm of lithium, the presence of boron had an ameliorating effect on the corrosion rates of Zircaloy-2. Similar to acceleration of corrosion by lithium, the inhibition by boron is due to a combined effect of pH neutralizing and its participation in the corrosion process.

  18. Zircaloy-4 stress corrosion by iodine: crack kinetics and influence of irradiation on the crack initiation

    International Nuclear Information System (INIS)

    Serres, A.

    2008-01-01

    During the PWR power transients, iodine-induced stress corrosion cracking (I-SCC) is one of the potential failure modes of Zircaloy-4 fuel claddings under Pellet-Cladding Interaction conditions. The primary objective of this study is to distinguish the parameters that contribute to the I-SCC phenomenon in iodized methanol solutions at ambient temperature, on notched tensile specimens, using crack growth rate measurements provided by Direct Current Potential Drop. The results show that for a KI lower than 20 MPa.m 1/2 , the IG and mixed IG/TG velocity of propagation is a linear function of KI, regardless of the propagation mode. Between 20 and 25 MPa.m 1/2 , the TG crack growth rate also depends linearly on KI, but increases at a faster rate with respect to KI than during the IG and mixed IG/TG propagation steps. The crack propagation direction and plane (LT and TL) have an impact on the propagation modes, but no impact on the kinetics. The increase of iodine content induces an increase of the crack growth rate for a given KI, and a decrease of the KI, threshold, allowing the crack propagation. This work enables us to quantify the effect of iodine content and of KI on the crack propagation step, propose a propagation law taking into accounts these parameters, and improve the I-SCC description for models. During operation, a zirconium cladding is neutron-irradiated, modifying its microstructure and deformation modes. The second objective of the study is therefore to investigate the impact of these modifications on I-SCC. For that purpose, smooth specimens in recrystallized Zircaloy-4 are proton-irradiated to 2 dpa at 305 C, the microstructure and deformation modes of unirradiated and irradiated Zircaloy-4 are characterized by TEM and SEM, and the influence of these radiation-induced modifications on the I-SCC susceptibility is studied. The Laves phases precipitates are slightly modified by irradiation. The formation of P -type dislocation loops correlated with

  19. Effect of deformation on crystallite characteristic and yield stress of zircaloy-4

    International Nuclear Information System (INIS)

    Sugondo; Futichah

    2007-01-01

    The effect of deformation (rolling) on micro strain, crystallite size, crystallite density, and yield strength of Zircaloy-4 was characterized by x-ray diffraction. The goal of this investigation is to characterize the cladding materials of PWR and the target is to have data on the crystallography of Zircaloy-4. The as-received material with the composition 1.3% Sn, 0.22% Fe, 0.1% Cr, and Zr balanced was cut 10 mm × 100 mm in size using diamond blade. The samples were cleaned and heated at 1100 °C for 2 hours and then quenched in cold water. Then the sample were cleaned and heated at 750 °C for 2 hours. Afterward the samples were cold rolled with 40%, 75%, and 80% reduction in thickness. After the preparation was completed, the crystals of the samples were characterized using X-ray diffraction. The processes being analysed were quenching followed by annealing, plastic deformation of annealing and reduction from 40% to 80%, and the constancy of the c/a ratio. From the analyses, three conclusions were obtained. Firstly, the annealing process at 750 °C of Zircaloy-4 from the quenched samples resulted in the recrystallization and the grain growth which was proven by the increase of micro strain from 25.05% to 32.83%, the increase of crystallite size from 10.1015 Å to 287.4798 Å, the decrease of dislocation density from 2.94E+16 m/m3 to 3.63E+13 m/m3, and the decrease of yield strength from 1125.52 MPa to 304.44 MPa. Secondly, the process of reduction of Zircaloy-4 from the annealed samples reduced to 80% resulted in the plastic deformation and crystallite which was shown by the decrease of micro strain from 32.83% to 3.15%, the decrease of crystallite size from 287.4798 Å to 10.9764 Å, the increase of dislocation density from 3.63E+13 m/m3 to 2.49E+16 m/m3, and the increase of yield strength from 304.44 MPa to 1057.69 MPa. Thirdly, the process of plastic deformation of Zircaloy-4 was limited by the constancy of the c/a ratio which was verified by the decrease

  20. Rethinking the Zircaloy Embrittlement Criteria and Its Impact on Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Lee, You Ho; Kim, Bo Kyung; No, Hee Cheon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    These fuel rod failure modes include integral thermal shock fracture, and impact tests. It is quite remarkable to see that the proposed Zircaloy embrittlemt criteria attained from ring compression tests, in general, successfully assure structural integrity of fuel rods subject to relevant failure modes in accidents. This fact demonstrates that ductility of Zircaloy is the key metric to structural integrity of fuel rods. However, the Zircaloy embrittlement criteria set in 1970s inevitably pose limitations that have become increasingly important for today's nuclear fuel and reactor operations. In particular, the criteria do not take into account the steady-state hydrogen embrittlement with burnup. This may be understandable considering the markedly lower discharge burnup in 1970s compared to that of today. The revision of the rule has been already conducted by the U.S NRC to account for high burnup effects on ECR while the temperature limit remains unchanged. The newly proposed rule of the U.S NRC stick to the similar ring compression tests conducted in the early 1970s. In the monumental experimental investigation of Hobson and Rittenhouse in 1972 and 1973, the experimental evidence for the current 1204oC was first addressed. The study found a reasonably accurate correlation between zero ductility temperature and the sum of alpha and oxide layer thickness for the specimens oxidized below 2200oF (1204 .deg. C). However, in spite of the similar oxidation degree, specimens oxidized at 2400 .deg. F (1315 deg. C) were markedly more brittle than specimens oxidized at 2200 .deg. F (1204 .deg. C). The study explained this by the increase in solid-solution hardening due to a higher oxygen solubility at a higher temperature. Such a nice experimental correlation attained between the nil ductility temperature and the remaining beta layer thickness fraction below 1204 .deg. C has become a critical basis for the current temperature limit; at 1315 .deg. C- thecorrelation

  1. Effects of stress on the oxide layer thickness and post-oxidation creep strain of zircaloy-4

    International Nuclear Information System (INIS)

    Lim, Sang Ho; Yoon, Young Ku

    1986-01-01

    Effects of compressive stress generated in the oxide layer and its subsequent relief on oxidation rate and post-oxidation creep characteristics of zircaloy-4 were investigated by oxidation studies in steam with and without applied tensile stress and by creep testing at 700 deg C in high purity argon. The thickness of oxide layer increased with the magnitude of tensile stress applied during oxidation at 650 deg C in steam whereas similar phenomenon was not observed during oxidation at 800 deg C. Zircaloy-4 specimens oxidized at 600 deg C in steam without applied stress exhibited higher creep strain than that shown by unoxidized specimens when creep-tested in argon. Zircaloy-4 specimens oxidized at 600 deg C steam under the applied stress of 8.53MPa and oxidized at 800 deg C under the applied stress of 0 and 8.53MPa exhibited lower strain than that shown by unoxidized specimen. The above experimental results were accounted for on the basis of interactions among applied stress during oxidation, compressive stress generated in the oxide layer and elasticity of zircaloy-4 matrix. (Author)

  2. SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

    International Nuclear Information System (INIS)

    Christie, W.H.; Carter, J.A.; Eby, R.E.; Landau, L.; Musick, W.R.

    1980-01-01

    Purpose of this study was to determine the amount of 10 B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 μg/cm 2

  3. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  4. Biaxial creep deformation of Zircaloy-4 PWR fuel cladding in the alpha,(alpha + beta) and beta phase temperature ranges

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Healey, T.; Horwood, R.A.L.

    1985-01-01

    The biaxial creep behaviour of Zircaloy-4 fuel cladding has been determined at temperatures between 973 - 1073 K in the alpha phase range, in the duplex (alpha + beta) region between 1098 - 1223 K and in the beta phase range between 1323 - 1473 K. This paper presents the creep data together with empirical equations which describe the creep deformation response within each phase region. (author)

  5. Effect of impurity elements Al, Mn, and N2 on the corrosion resistance of zircaloy-2 in high temperature water and steam

    International Nuclear Information System (INIS)

    Gadiyar, H.S.

    1978-01-01

    Although the impurity limits are specified in standard zircaloy-2, it is possible that during its manufacture some of the impurities may exceed by a few ppm than the normally set values. It is necessary to understand the corrosion behaviour of such zircaloy-2 which contain a small amount of excessive impurities. This report summarizes some such data of the impurities aluminium, manganese and nitrogen. It is seen that the common impurities which can affect the corrosion of zircaloy-2 significantly are Al and N 2 and to a lesser extent Mn. (author)

  6. 21 CFR 522.1222 - Ketamine hydrochloride injectable dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Ketamine hydrochloride injectable dosage forms. 522.1222 Section 522.1222 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN... ANIMAL DRUGS § 522.1222 Ketamine hydrochloride injectable dosage forms. ...

  7. 21 CFR 520.1696 - Penicillin oral dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Penicillin oral dosage forms. 520.1696 Section 520.1696 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS ORAL DOSAGE FORM NEW ANIMAL DRUGS § 520.1696 Penicillin oral...

  8. 21 CFR 526.1696 - Penicillin intramammary dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Penicillin intramammary dosage forms. 526.1696 Section 526.1696 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS INTRAMAMMARY DOSAGE FORMS § 526.1696 Penicillin...

  9. 21 CFR 522.1660 - Oxytetracycline injectable dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Oxytetracycline injectable dosage forms. 522.1660 Section 522.1660 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES... § 522.1660 Oxytetracycline injectable dosage forms. ...

  10. 21 CFR 520.905 - Fenbendazole oral dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Fenbendazole oral dosage forms. 520.905 Section 520.905 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES... Fenbendazole oral dosage forms. ...

  11. 21 CFR 520.45 - Albendazole oral dosage forms.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Albendazole oral dosage forms. 520.45 Section 520.45 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS, AND RELATED PRODUCTS ORAL DOSAGE FORM NEW ANIMAL DRUGS § 520.45 Albendazole oral...

  12. Archives: les cahiers du cread

    African Journals Online (AJOL)

    Items 1 - 24 of 24 ... Archives: les cahiers du cread. Journal Home > Archives: les cahiers du cread. Log in or Register to get access to full text downloads. Username, Password, Remember me, or Register · Journal Home · ABOUT THIS JOURNAL · Advanced Search · Current Issue · Archives. 1 - 24 of 24 Items. 2016 ...

  13. Le commerce du Nord

    OpenAIRE

    Pourchasse, Pierrick; Bouëdec, Gérard Le

    2015-01-01

    Au XVIIIe siècle, la France s'approvisionne abondamment dans les pays du Nord : bois, chanvre et goudron de la Baltique, tonnellerie de Poméranie, pêche de rogue de Norvège, graines de lin de Courlande, barres de fer suédois… Sa balance commerciale est pourtant positive grâce aux sels, aux vins et surtout des nouvelles marchandises coloniales. Or, la plupart des transactions passent par l’incontournable intermédiaire hollandais. Les explications sur l’absence des Français dans le Nord sont re...

  14. La voie du Centre

    OpenAIRE

    2017-01-01

    Après avoir quelque peu louvoyé dans un discours apparemment anarchique où les souvenirs semblent se bousculer sans autre fil conducteur que la référence obsédante au mal satanique, brusquement le narrateur annonce un événement primordial : Foi um fato que se deu, um dia, se abriu. O primeiro. Depois o senhor verá por quê, me devolvendo minha razão (79). Cette introduction situe la rencontre du Menino comme fondatrice d’un destin dont il reviendrait au narrataire de dégager les enchaînements....

  15. Bulletin du CRDI #126

    International Development Research Centre (IDRC) Digital Library (Canada)

    26 févr. 2018 ... Dans ce numéro, découvrez comment la recherche financée par le CRDI permet d'améliorer la santé des mères et des enfants dans les pays du Sud et comment les innovations techniques et sociales de l'initiative SEARCH permettent de surmonter les défis liés à la cybersanté. N'oubliez pas non plus de ...

  16. Radiation dosage of various CT-methods in lung diagnostics

    International Nuclear Information System (INIS)

    Heinz-Peer, G.; Weninger, F.; Nowotny, R.; Herold, C.J.

    1996-01-01

    Introduction of the computed tomography index CTDI and the multiple scan average dose (MSAD) has led to standardization of the dose description in CT examinations. Despite the use of these dose parameters, many different dosages are reported in the literature for different CT methods. In addition, there is still a wide range of radiation dosimetry results reported for conventional CT, helical CT, and HRCT used in chest examinations. The variations in dosage are mainly due to difference in factors affecting the dose, i.e. beam geometry, beam quality, scanner geometry ('generation'), and operating parameters. In addition, CT dosimetry instrumentation and methodology make a contribution to dosages. Recent studies calculating differences in factors affecting dosage and CT dosimetry and using similar operating parameters, show similar results in CT dosimetry for conventional and helical CT. On the other hand, dosages for HRCT were greatly reduced. This was mainly caused by narrow beam collimation and increasing section spacing. (orig.) [de

  17. Effects of Nitrogen Implantation on the Resistance to Localized Corrosion of Zircaloy-4 in a Chloride Solution

    International Nuclear Information System (INIS)

    Lee, Sung Joon; Kwon, Hyuk Sang; Kim, Wan; Choi, Byung Ho

    1996-01-01

    The influences of ion dose and substrate temperature on the resistance to localized corrosion of nitrogen-implanted Zircaloy-4 are examined in terms of potentiodynamic anodic polarization tests in deaerated 4M NaCl solution at 80 .deg. C. Nitrogen implantations into the Zircaloy-4 were performed under conditions of varying the ion dose from 3 x 10 17 to 1.2 x 10 18 ions/cm 2 and of maintaining the substrate temperatures respectively at 100, 200, and 300 .deg. C by controlling the current density of ion beam. The resistance to localized corrosion of Zircaloy-4 was significantly increased with increasing the ion dose when implanted at substrate temperatures above 200 .deg. C. However, it was not almost improved by implantation at 100 .deg. C. Specifically, the pitting potential increased from 350mV (vs. SCE) for the unimplanted to values of 900 to about 1400mV (vs. SCE) for the implanted alloy depending on the nitrogen dose. This significant improvement in the resistance to localized corrosion of the implanted Zircaloy-4 was found to be associate with the formation of compound layers of ZrO 2 + ZrN during the implantation. The galvanostatic anodization tests on the nitrogen-implanted Zircaloy-4 in 1M H 2 SO 4 at 20 .deg. C demonstrated that an increase in the ion dose and also in the substrate temperature increased the thickness of the compound layer of ZrO 2 + ZrN, and hence increased the pitting potential of the alloy. The low resistance to localized and general corrosion of the alloy implanted at 100 .deg. C was attributed to the increase in surface defect density and also to thinner implanted layer compared with those formed at higher temperatures

  18. Determinations of the temperature of terminal solid solubility in dissolution and precipitation of hydrogen/deuterium in irradiated Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Vizcaino, P [CNEA-CONICET, Centro Atomico Ezeiza (Argentina)

    2012-07-01

    The proposed plan is an approach to the metallurgical consequences of the high neutron fluencies (10''2''2 n/cm''2) on the hydrogen behavior in zirconium based alloys, based on the significance of the microstructural behavior of the high burn up fuel claddings during the dry storage period. The studies are focused on Zircaloy-4, concerning to two processes: Neutron irradiation damage; Hydrogen pick up. The Zircaloy-4 was taken from cooling channels of the PHWR Atucha 1. These components remained more than 10 years in service, reaching neutron fluencies up to 10''2''2 n/cm''2. In the last recent years, measurements of the hydride dissolution temperatures have shown that hydrogen solubility is affected by the neutron irradiation, increasing it respect to the unirradiated Zircaloy solubility. In addition, in this material the amorphization/dissolution of the second phase particles (SPPs) was observed, being proposed an interaction between the hydrogen atoms, the SPPs and the irradiation defects as a possible explanation of the observed behavior. For the present case, attention will be focused on the hydride precipitation process, since it is strongly related with delay hydrogen cracking initiation, a problem of direct concern for the dry storage. The goal of the present proposal is to make an approach to the source of the observed effect, applying several specific techniques as differential scanning calorimetry (DSC), high resolution x-ray diffraction and transmission electron microscopy. The objectives can be divided as follows: Determination of the temperatures of terminal solid solubility in dissolution (TTSSd) and in precipitation (TTSSp) in high fluency irradiated Zircaloy-4, reproducing the temperatures at which the Zircaloy fuel claddings remain during dry storage by an annealing program during the DSC experiments; Observations by optical and transmission electron microscopy of the hydride distribution before (as received material) and after high temperature

  19. Choeur du CERN : Concert

    CERN Multimedia

    CERN Choir

    2017-01-01

    Une œuvre à découvrir! La grande Missa pro defunctis de François-Joseph Gossec (1734-1829) est le chef-d’œuvre tôt venu (à vingt-cinq ans) d’un compositeur qui vivra encore 70 ans après sa création. Elle a connu la gloire, puis s’est fait un peu oublier. Pas du tout le monde cependant : des musicologues ont montré ce que le Requiem de Mozart lui devait ; et il suffit de l’avoir entendue pour comprendre pourquoi Berlioz (qui avait vingt-six ans à la mort de Gossec) en a été impressionné : les nombreux cuivres et bois répartis dans des endroits plus ou moins cachés de la salle de concert pour exprimer les frayeurs du Jugement dernier annoncent son Requiem – et celui de Verdi. Mais « plus encore que par...

  20. Les risques du travail

    CERN Document Server

    Thébaud-Mony, Annie

    2015-01-01

    Depuis les années 1990, les conditions de travail se sont peu à peu imposées dans le débat social. Néanmoins, la situation reste critique. Les risques traditionnels n'ont pas disparu : les manutentions lourdes, l'exposition professionnelle aux cancérogènes, au bruit ou aux vibrations demeurent répandues... De plus, certaines " améliorations " n'ont fait que déplacer et dissimuler les problèmes, telle l'externalisation des risques grâce à la sous-traitance. Dans le même temps, les transformations du travail et des modalités de gestion de la main-d'œuvre ont fragilisé les collectifs et accru l'isolement des salariés, conduisant à une montée visible de la souffrance psychique. Face à ces évolutions, il est plus que jamais nécessaire que tous les acteurs concernés, en particulier les salariés eux-mêmes et leurs représentants, s'approprient les connaissances indispensables pour améliorer la protection de la santé sur les lieux du travail. Tel est le but de ce livre, qui renouvelle int�...

  1. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

  2. The effect of neutron irradiation on the mechanical properties of welded zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Evans, D G

    1962-07-15

    Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260{sup o}C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900{sup o}C for 40 minutes, another group contained welded areas in the centre of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld, Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. The transition temperature of Zircaloy-2 increased appreciably upon welding. This was accompanied by a decrease in absorbed energy values for all temperatures between 0{sup o} and 300{sup o}C. Neutron irradiation had no effect on the impact properties of welded. Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260{sup o}C, and increased the 260{sup o}C PL, YS and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260{sup o}C strength properties of the as-welded material. It was found that the unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. These fractures were similar to those observed in irradiated fuel bundles which had been damaged during transfer operations. A large amount of scatter rendered some

  3. Design of an integrated system to recycle Zircaloy cladding using a hydride–milling–dehydride process

    Energy Technology Data Exchange (ETDEWEB)

    Kelley, Randy, E-mail: rkelley@pitt.edu [Mechanical Engineering Department, 236 Engineering and Science Building, University of Pittsburgh – Johnstown, Johnstown, PA 15904 (United States); McDeavitt, Sean [Texas A and M University, Department of Nuclear Engineering, 327 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

    2013-10-15

    Highlights: • Dehydriding zirconium hydride was studied at relatively low temperatures (<800 °C). • High vacuum pressures decrease dehydriding temperatures. • Specialized equipment was designed, built and demonstrated to process zirconium. • The process hydrided–milled–dehydrided zirconium metal to a fine metal powder. • Two powder samples were analyzed and proved the operation of the machine. -- Abstract: A hydride–dehydride process was evaluated to recover a portion of spent nuclear fuel cladding; a zirconium alloy (Zircaloy), as a metal powder that may be used for advanced nuclear fuel applications. The investigation was part of a broader study that sought to determine the viability of recovering components of used nuclear fuel to for a metal matrix cermet for transuranic burning. The zirconium powder process begins with the conversion of Zircaloy cladding hulls into a brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. In support of this, a specialized piece of equipment was designed to demonstrate the entire zirconium conversion process to transform Zircaloy tubes into metal powder without intermediate handling. This was accomplished by building a milling system that rotates inside of controlled atmosphere chamber with an internal heater. The hydriding process was accomplished using an argon–5% hydrogen atmosphere at 500 °C. The process variables for the dehydriding process were determined using a thermogavimetric analysis (TGA) method. It was determined that a rough vacuum (∼0.001 bar) and 800 °C were sufficient to decompose the zirconium hydride. Zirconium metal powder was created using different milling times: 45 min (coarse powder) and 12 h (fine powder). X-ray diffraction (XRD) analysis indicated that the process produced a zirconium metal. Additionally, visual observations of the samples silvery

  4. Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E

    2000-07-01

    We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of {approx}450 deg C. After irradiation, the samples contained needle-like {beta}-Nb precipitates in the {alpha}-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D{sub 2}O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D{sub 2}0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D{sub 2}O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure

  5. Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

    International Nuclear Information System (INIS)

    Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E.

    2000-01-01

    We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of ∼450 deg C. After irradiation, the samples contained needle-like β-Nb precipitates in the α-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D 2 O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D 2 0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D 2 O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure tube materials. Thus, 10-Me

  6. Determining S-1 dosage at hospitals prioritizing cancer chemotherapy

    International Nuclear Information System (INIS)

    Morimoto, Shigefumi; Kitada, Noriaki; Anami, Setsuko

    2008-01-01

    Although it is recommended that the standard S-1 dosage should be based on how large the body surface area is, an on-site setting of the appropriate dosage is often lower than the standard one, depending on the individual's condition and considering possible side effects and so, on. Here, we investigated usage conditions for S-1 as a part of field training for expert pharmacists at our hospital that performs total clinical treatments. Decreases in dosage per day for elderly patients were although the standard dosage is generally determined according to the amount of a patient's body surface. We conducted a retrospective survey with a total 90 patients by creating a tree-diagram to identify a reduction standard. It was found that the S-1 dosage was decreased when there were side effects, aggravation in performance status, decrease in kidney function, old age, combined injection chemotherapy, and a decrease in radiation therapy performance. The dosage decreases without such medical reasons were seen in only 4 of the 90 patients. At hospitals giving priority to chemotherapy, it became clear that appropriate treatment was promoted by decreasing. The individual target dosage on the basis of daily medical examination. (author)

  7. Investigation of effect of air flow rate on Zircaloy-4 oxidation kinetics and breakaway phenomenon in air at 850 .deg. C

    International Nuclear Information System (INIS)

    Maeng, Yunhwan; Lee, Jaeyoung; Park, Sanggil

    2016-01-01

    This paper analyzed an effect of flow rate on oxidation kinetics of Zircaloy-4 in air at 850 .deg. C. In case of the oxidation of Zircaloy-4 in air at 850 .deg. C, acceleration of oxidation kinetics from parabolic to linear (breakaway phenomenon) occurs. Oxidation and breakaway kinetics of the Zircaloy-4 in air was experimentally studied by changing a flow rate of argon/air mixture. Tests were conducted at 850 .deg. C under constant ratio of argon and air. The effects of flow rate on the oxidation and breakaway kinetics was observed. This paper is based on a revised and considerably extended presentation given at the 21 st International Quench Workshop. The effects of flow conditions on the oxidation kinetics of Zircaloy-4 samples were explained with residence time and percent flow efficiency. In addition, several issues were observed from this study, interdiffusion at breakaway and deformation of oxide structure by breakaway phenomenon

  8. Jouer du piano

    Directory of Open Access Journals (Sweden)

    Fériel Kaddour

    2011-04-01

    Full Text Available La réflexion s’appuie dans un premier temps sur une opposition entre deux attitudes de pianistes  à l’égard du travail à l’instrument : Gould, qui revendique une séparation d’avec le clavier pour ne privilégier que la lecture; Arrau, dont la technique au contraire vise à « faire corps » avec son piano. L’étude de ces deux démarches d’interprètes conduit à une conclusion croisée : l’abstraction gouldienne n’est rien d’autre qu’un déplacement du jeu vers d’autres instruments (ceux qui servent à la prise de son et au montage de ses enregistrements ; le « faire-corps » hérité de la culture pianistique romantique est plus dialectique que fusionnel, et en cela implique une capacité de mise à distance. A partir de cette double conclusion, on tâche enfin de repenser la place du jeu à l’instrument dans la mise en œuvre d’une interprétation, en interrogeant le dialogue qui s’instaure entre la partition telle qu’elle s’écrit et le geste tel qu’il se joue.Our study leans on an opposition between two pianists' attitudes about their work with the instrument. Gould claims a necessary separation from the keyboard in order to prioritize reading. Arrau, on the contrary, relies on a technique which consists in “being one” with his piano. The analysis of these two interprets’ behaviours leads to a crossed conclusion: the gouldian abstraction is nothing else than a displacement of the playing towards another kind of instruments, the ones he uses in sound recording and cut up; Arrau’s “being one” is more dialectic than at first sight, and it therefore implies a real distancing from the piano. This constatation leads to rethink the place of the piano playing in the setting of an interpretation, and to highlight the real dialogue which develops itself between the score as it has been written and the gesture as it is played.

  9. Effect of dynamic strain aging on cyclic stress response and deformation behavior of Zircaloy-2

    International Nuclear Information System (INIS)

    Sudhakar Rao, G.; Verma, Preeti; Mahobia, G.S.; Santhi Srinivasa, N.C.; Singh, Vakil; Chakravartty, J.K.; Nudurupatic, Saibaba

    2016-01-01

    The effect of strain rate and temperature was studied on cyclic stress response and deformation behavior of annealed Zircaloy-2. Dynamic strain aging was exhibited under some test conditions. The cyclic stress response was found to be dependent on temperature and strain rate. At 300 °C, with decrease in strain rate, there was decrease in the rate as well as the degree of cyclic hardening. However, at 400°C, there was opposite trend and with decrease in strain rate both the rate as well as the degree of hardening increased. The deformation substructure showed dislocation bands, dislocation vein structure, PSB wall structure at both the temperatures. Irrespective of the temperature, there was dislocation loop structure, known as corduroy structure, at both the test temperatures. Based on the dislocation structure, the initial linear hardening is attributed to development of veins and PSB wall structure and the secondary hardening to the Corduroy structure. (author)

  10. Irradiation effects on Fe distributions in zircaloy-2 and Zr-2.5Nb

    International Nuclear Information System (INIS)

    Zou, H.; Hood, G.M.; Roy, J.A.

    1995-03-01

    Irradiation of large-grained Zr-2.5Nb (ZN) and Zircaloy-2 (Zy) with 1.5 MeV Ar ions to a fluence of ∼ 10 20 /m 2 (≡ 10 dpa) at 50, 300 and 420 deg C leads to enhanced α-phase Fe levels of 250-1500 ppma, compared to equivalent non-irradiated state values of ∼ 70 ppma. In ZN the β-phase Fe levels fell from about 6000 to 3500 ppma: this result accords, qualitatively, with the loss of Fe from the β-phase following in-service neutron irradiation. Measurements on Zy showed that the Fe concentrations were higher near the specimen surfaces. Limited data for Ni distributions in Zy show similar (to Fe) behaviour. (author). 18 refs., 2 tabs

  11. On-line ultrasonic inside-diameter control system for Zircaloy

    International Nuclear Information System (INIS)

    Tanaka, Y.; Fujii, N.; Komatsu, M.; Kubota, H.

    1984-01-01

    An ultrasonic inside-diameter (ID) control system was used during the final etching process for producing Zircaloy nuclear fuel cladding tubes. This results in establishing automatic inside-diameter control during etching with an automatic etching system. In this system, the inside-diameter at the center point in the length of each tube is continuously measured with the ultrasonic inside-diameter measuring equipment during the etching process and the etching is automatically stopped by a signal from the control equipment when the inside-diameter reaches the target value. This made the final etching process economical and suitable for large-scale production, having an equal or better level at the inside-diameter of tubes etched with this system than those made by a process controlled by an air-micrometer

  12. Stress Corrosion Cracking of Zircaloy-4 in Halide Solutions: Effect of Temperature

    Directory of Open Access Journals (Sweden)

    Farina S.B.

    2002-01-01

    Full Text Available Zircaloy-4 was found to be susceptible to stress corrosion cracking in 1 M NaCl, 1 M KBr and 1 M KI aqueous solutions at potentials above the pitting potential. In all the solutions tested crack propagation was initially intergranular and then changed to transgranular. The effect of strain rate and temperature on the SCC propagation was investigated. An increase in the strain rate was found to lead to an increase in the crack propagation rate. The crack propagation rate increases in the three solutions tested as the temperatures increases between 20 and 90 °C. The Surface-Mobility SCC mechanism accounts for the observation made in the present work, and the activation energy predicted in iodide solutions is similar to that found in the literature.

  13. Hydride reorientation in Zircaloy-4 examined by in situ synchrotron X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Weekes, H.E. [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom); Jones, N.G. [Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Lindley, T.C. [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom); Dye, D., E-mail: david.dye@imperial.ac.uk [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom)

    2016-09-15

    The phenomenon of stress-reorientation has been investigated using in situ X-ray diffraction during the thermomechanical cycling of hydrided Zircaloy-4 tensile specimens. Results have shown that loading along a sample’s transverse direction (TD) leads to a greater degree of hydride reorientation when compared to rolling direction (RD)-aligned samples. The elastic lattice micro-strains associated with radially oriented hydrides have been revealed to be greater than those oriented circumferentially, a consequence of strain accommodation. Evidence of hydride redistribution after cycling, to α-Zr grains oriented in a more favourable orientation when under an applied stress, has also been observed and its behaviour has been found to be highly dependent on the loading axis. Finally, thermomechanical loading across multiple cycles has been shown to reduce the difference in terminal solid solubility of hydrogen during dissolution (TSS{sub D,H}) and precipitation (TSS{sub P,H}).

  14. Corrosion behavior of Zircaloy 4 cladding material. Evaluation of the hydriding effect

    International Nuclear Information System (INIS)

    Blat, M.

    1997-04-01

    In this work, particular attention has been paid to the hydriding effect in PIE and laboratory test to validate a detrimental hydrogen contribution on Zircaloy 4 corrosion behavior at high burnup. Laboratory corrosion tests results confirm that hydrides have a detrimental role on corrosion kinetics. This effect is particularly significant for cathodic charged samples with a massive hydride outer layer before corrosion test. PIE show that at high burnup a hydride layer is formed underneath the metal/oxide interface. The results of the metallurgical examinations are discussed with respect to the possible mechanisms involved in this detrimental effect of hydrogen. Therefore, according to the laboratory tests results and PIE, hydrogen could be a strong contributor to explain the increase in corrosion rate at high burnup. (author)

  15. Observations on the influence of tube manufacturing technique on iodine stress corrosion cracking of unirradiated Zircaloy

    International Nuclear Information System (INIS)

    Syrett, B.C.; Cubicciotti, D.; Jones, R.L.

    1979-01-01

    Closed-end tube pressurization tests at 593 K were used to compare the susceptibilities to iodine stress corrosion cracking (SCC) of two lots of Zircaloy-2 tubing manufactured by different suppliers. Although both tubings were produced to exactly the same specifications in terms of dimensions, chemical composition, burst strength, and certain other properties, as-received specimens from the two lots exhibited markedly different behavior in iodine SCC tests. The tubing from one supplier had a lower SCC threshold stress and failed about 30 times more quickly than the tubing from the other supplier. However, it was found that this difference in SCC susceptibility was eliminated if the internal surfaces of the specimens were polished to a 3 μm finish prior to testing. These observations are discussed in terms of possible effects of surface or near-surface chacteristics of the tubing on SCC susceptibility

  16. Annealing studies of zircaloy-2 cladding at 580-8500C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1978-05-01

    For fuel element cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant accidents (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then much experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 0 C for returning Zircaloy cladding to the annealed condition, so that for any transient, a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  17. NORA-2, a model for creep deformation and rupture of zircaloy at high temperatures

    International Nuclear Information System (INIS)

    Raff, S.; Meyder, R.

    1983-01-01

    A model has been developed to describe Zircaloy cladding behaviour under LOCA and small leak conditions within specified temperature range and strain rates. The deformation model consists of a strain rate equation with two components representing strain rate controlled contributions from different deformation mechanisms. Transition from one mechanism to the other produces the strain rate dependence of the stress exponent of steady state creep. During transient creep the change of creep mechanisms produces a flow softening behaviour which induces unstable creep. Together with a strain hardening model, the strain history can be described for low and high strain values. The influence of oxidation is taken into account by modelling hardening due to solid solution of oxygen, cracking of the brittle oxide and oxygen stabilised α-phase layers, and by an oxidation-induced creep component in steam atmosphere. The rupture criterion is based on a strain fraction rule whose variables are temperature, strain rate or applied stress, and oxygen content. (author)

  18. Development of a Zircaloy creep and failure model for LOCA conditions

    International Nuclear Information System (INIS)

    Raff, S.; Meyder, R.

    1981-01-01

    The present status of NORA model for zircaloy-4 creep and failure in the high temperature region (from 600 deg C up to 1200 deg C) is described. Temperature dependence, strain hardening and oxygen content are found to be the most important features of the strain rate creep equation. The failure criterion is based on a modified strain fraction rule. Variables of this criterion are temperature, strain rate or applied stress respectively and oxygen content. Concerning the application of the deformation model, deduced from uniaxial tests, to tube deformation calculation the axial ballooning shape has to be taken into account. Its influence on the tube stress components and therefore on strain rate is discussed. A further improvement of the deformation model concerning yield drop and irregular creep behaviour aims at the enlargement of the range of applicability and reduction of the error band of the model

  19. Characterization of fatigue-corrosion phenomena for Zircaloy in iodine environment

    International Nuclear Information System (INIS)

    Schuster-Magallon, Isabelle

    1986-01-01

    In this research thesis, the acquisition of data related to crack propagation rates and to smooth specimen lifetime in corrosion-fatigue of zircaloy allowed the quantification of the influence of iodine with respect to material, to loading direction and to test frequency. A systematic fractographic examination of propagation and fatigue strength specimens allowed the fatigue-corrosion fracture scenario to be described. This scenario comprises pitting for a stress higher than a threshold stress, the development of an intergranular corrosion area limited by a threshold stress intensity factor overrun, and the propagation by fatigue-corrosion in steady regime. This propagation is an association of a quasi-cleavage which is typical of stress corrosion cracking, and a plastic deformation under fatigue. This combination leads to the sudden disappearance of cleavage, and to a ductile fracture [fr

  20. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  1. Influence of specimen design on the deformation and failure of zircaloy cladding

    International Nuclear Information System (INIS)

    Bates, D.W.; Koss, D.A.; Motta, A.T.; Majumdar, S.

    2000-01-01

    Experimental as well as computational analyses have been used to examine the deformation and failure behavior of ring-stretch specimens of Zircaloy-4 cladding tubes. The results show that, at least for plastically anisotropic unirradiated cladding, specimens with a small gauge length l to width w ratio (l/w ∼ 1) exhibit pronounced non-uniform deformation along their length. As a result, specimen necking occurs upon yielding when the specimen is fully plastic. Finite element analysis indicates a minimum l/w of 4 before a significant fraction of the gauge length deforms homogeneously. A brief examination of the contrasting deformation and failure behavior between uniaxial and plane-strain ring tension tests further supports the use of the latter geometry for determining cladding failure ductility data that are relevant to certain reactivity-initiated accident conditions

  2. Iodine-induced stress corrosion cracking of fixed deflection stressed slotted rings of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Sejnoha, R.; Wood, J.C.

    1978-01-01

    Stress corrosion cracking of Zircaloy fuel cladding by fission products is thought to be an important mechanism influencing power ramping defects of water-reactor fuels. We have used the fixed-deflection stressed slotted-ring technique to demonstrate cracking. The results show both the sensitivity and limitations of the stressed slotted-ring method in determining the responses of tubing to stress corrosion cracking. They are interpreted in terms of stress relaxation behavior, both on a microscopic scale for hydrogen-induced stress-relief and on a macroscopic scale for stress-time characteristics. Analysis also takes account of nonuniform plastic deformation during loading and residual stress buildup on unloading. 27 refs

  3. Role of internal stresses in the transient of irradiation growth of zircaloy-2

    International Nuclear Information System (INIS)

    Tome, C.N.; Christodoulou, N.; Turner, P.A.; Miller, M.A.; Woo, C.H.; Root, J.; Holden, T.M.

    1995-07-01

    A 'self-consistent' polycrystalline model is used to simulate irradiation growth of Zircaloy-2 samples irradiated at about 330 K. The predictions of the model are compared with experimental measurements obtained from specimens irradiated in the Advanced Test Reactor (ATR) at Idaho Falls. Three types of material are studied here: annealed, cold worked in tension and cold worked by rolling. In general, the growth rate attains a steady-state value after it goes through a transient that depends on the initial state of the material. The transient growth behaviour is explained in terms of the evolution of intergranular residual stresses that are present in the sample, and in terms of the dislocation structure. From this study, information regarding irradiation creep and growth mechanisms occurring at the single crystal level is obtained. (author). 28 refs., 1 tab., 4 figs

  4. Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

    1994-12-31

    The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

  5. An investigation of the microstructures of heat-treated zircaloy-4

    International Nuclear Information System (INIS)

    Bangaru, N.V.

    1985-01-01

    A TEM/STEM investigation of the microstructure and microchemistry of commercial Zircaloy-4 samples subjected to three different final heat treatments in the laboratory has been conducted to understand the processing-microstructure-corrosion relationships in these alloys. Pronounced differences in the volume fraction, morphology, and chemistry of the intermetallic particles as well as in the α phase microstructure have been observed among the beta-quenched, as-received (stress-relieved) and alpha-annealed samples. The beta-quenched sample exhibits the most uniform microstructure consisting of acicular α phase with lath boundary Sn enrichment and fine intermetallic particle formation. The as-received sample has the most inhomogeneous microstructure made up of annealed and deformed α phase. The relevance of the observed microstructural features to the nodular corrosion susceptibility is discussed in the light of some existing models of modular corrosion. (orig.)

  6. Effect of an Excess of Iron and Hydriding on the Metallurgical Properties of Zircaloy-4

    International Nuclear Information System (INIS)

    Ghilarducci, Ada; Ramos, R; Martin, E; Peretti, Hernan; Corso Hugo

    2000-01-01

    Results are presented of mechanical properties and microstructure morphologies obtained in samples of zircaloy-4 of modified composition by an excess in the content of alloying elements as well as hydrides.This work is focused mainly on the effect of 1000 wt. ppm additional Fe as compared to the standard composition alloy.The study is carried out by means of tensile tests at room temperature and at 240 0 C, by hardness tests, by SEM observations and EDS microanalysis.The results indicate that precipitates concentrate along grain boundaries in all cases, and that for higher contents of alloying elements corresponds a higher quantity of precipitates and smaller grain sizes.Except for the hydrided sample, the fracture is ductile with cavities nucleated at precipitates. Finally it is concluded that an increase in Fe content affect the mechanical properties

  7. Analysis of the texture of zircaloy-4 sheet by crystallite orientation distribution function

    International Nuclear Information System (INIS)

    Ryoo, Hwei Soo; Hwang Sun Keum

    1990-01-01

    In order to analyze the texture variation of Zircaloy-4 sheet the Roe's method of calculating the crystallite orientation distribution function(CODF) for hcp system was computer programmed. The coefficients W lmn of CODF were calculated from plane-normal distribution pole figures obtained by X-ray diffraction, and the CODF was computed from a series expansion of spherical harmonics. The Legendre function, which is the basis of the harmonics, was computed up to l=16 to account for the symmetry systems of specimen and hcp crystal. A cross-rolling followed by beta-phase heat treatment and furnace cooling increased the density of basal poles along the sheet normal direction and rotated prism poles around the c axis. (Author)

  8. Behaviour of MZFR-type Zircaloy-4 cans under tensile stress

    International Nuclear Information System (INIS)

    Bordoni, R.A.; Casario, J.A.; Coroli, Graciela; Povolo, Francisco

    1981-01-01

    The paper describes the experimental procedure and results from the tensile tests of Zircaloy-4 fuel cans of the MZFR-type, performed at temperatures ranging from 250 to 450 deg C and for a relative deformation velocity of about 0.5%/min. In the representation of the results by a curve of the type sigma = K epsilon/sup n/, two different stages are observed. By statistically fitting the experimental curves, the values for the K and n parameters were obtained for each stage as a function of temperature. The results are discussed and compared with similar data found in current literature. It is concluded that new tests on tubes of different characteristics are necessary in order to obtain a clearer idea about the mechanical behaviour of these materials. (C.A.K.) [es

  9. Annealing studies of Zircaloy-2 cladding at 580-850 deg C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1983-01-01

    For fuel rod cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then a great deal of experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 deg C for returning Zircaloy cladding to the annealed condition, so that for any transient a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  10. Investigation of the high temperature steam oxidation of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Leistikow, S.; Berg, H. v.; Kraft, R.; Pott, E.; Schanz, G.

    1979-01-01

    Also for the ORNL Zircaloy 4 cladding material, an intermediate decrease of the proportion of the ZrO 2 /α-phase layer was found, followed by an drastic increase when the breakaway of the ZrO 2 -scale occurred. Other reasons for small divergencies were evaluated, for instance temperature and time measurements, metallographic evaluation of layer thicknesses, consequences of one-sided (ORNL) and double-sided (KfK) oxidation. The so-called anomalous effect of steam oxidation during temperature transients was reproduced qualitatively and-in case that a reduced gain of oxygen was observed-explained by the predominant existence of the monoclinic oxide phase. The creep-rupture tests below 800 0 C showed a moderate prolongation of time-to-rupture when the tests were performed in steam (or after preoxidation in steam) instead of argon. Also slightly reduced maximum circumferential strain could be measured. (orig./RW) [de

  11. The influences of deformation velocity and temperature on localized deformation of zircaloy-4 in tensile tests

    International Nuclear Information System (INIS)

    Boratto, F.J.M.

    1973-01-01

    A new parameter to describe the necking stability in zircaloy-4 during tensile tests is introduced. The parameter is defined as: s = ∂Ln (dσ/dε)/∂Ln ((1/L)dL/dt) for constant temperature, deformation and history. Measures of stress strain rate sensitivity n, reduction of the area at fracture, and deformation profiles of tensile fracture, are done. A complete description of the curve of non-uniform deformation variation with the temperature, is presented. The results are compared with existing data for pure commercially titanium. The influence of strain rate and history on s and n parameters, in the temperature range from 100-700 0 C). (author) [pt

  12. Fatigue tests and characterization of resulting microstructure by transmission electron microscope on zircaloy 4

    International Nuclear Information System (INIS)

    Di Toma, S.; Bertolino, G.; Tolley, A.

    2012-01-01

    This work reports the results of load controlled tension-tension fatigue tests on Zircaloy 4 (Zy-4). The resulting microstructure, particularly the kind and density of dislocations was characterized using a Transmission Electron Microscope (TEM). Specimens were cut from a rolled plate, with tensile axis parallel and perpendicular to the rolling direction. The results show a significant anisotropy of the mechanical properties due to the strong texture developed during rolling. Mainly type dislocations were observed, only in a longitudinal tensile axis specimen, dislocations were observed with a much lower density. The Schmid factors corresponding to the different glide systems were determined for specific grains in both tensile directions (author)

  13. Microstructure and Oxidation Behavior of CrAl Laser-Coated Zircaloy-4 Alloy

    Directory of Open Access Journals (Sweden)

    Jeong-Min Kim

    2017-02-01

    Full Text Available Laser coating of a CrAl layer on Zircaloy-4 alloy was carried out for the surface protection of the Zr substrate at high temperatures, and its microstructural and thermal stability were investigated. Significant mixing of CrAl coating metal with the Zr substrate occurred during the laser surface treatment, and a rapidly solidified microstructure was obtained. A considerable degree of diffusion of solute atoms and some intermetallic compounds were observed to occur when the coated specimen was heated at a high temperature. Oxidation appears to proceed more preferentially at Zr-rich region than Cr-rich region, and the incorporation of Zr into the CrAl coating layer deteriorates the oxidation resistance because of the formation of thermally unstable Zr oxides.

  14. A regression model for zircaloy cladding in-reactor creepdown: Database, development, and assessment

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.; Lanning, D.

    1987-01-01

    The paper presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in a PWR and a BWR. This model accounts for variation in the zircaloy cladding heat treatments - cold worked and stress relieved material typically used in a PWR and fully recrystallized material typically used in a BWR. This model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. The paper also presents a comparison between cladding creep calculations by the creepdown model and corresponding test results from the KWU/CE program. ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the creepdown model calculates cladding creep strains reasonably well. (orig./HP)

  15. Solvent effects on stress corrosion cracking of zirconium and Zircaloy-4 in iodine

    International Nuclear Information System (INIS)

    Farina, Silvia B.; Duffo, Gustavo S.; Galvele, Jose R.

    2000-01-01

    Localized corrosion (pitting, intergranular attack and stress corrosion cracking) of Zircaloy-4 and its principal component, zirconium, was investigated in solutions of iodine in different alcohols (methanol, ethanol, 1-propanol, 1-butanol, 1-pentanol and 1-octanol). Intergranular attack was found in all of the solutions tested, and the attack velocity increases when the size of the alcohol molecule decreases. In some cases it was found that intergranular attack is accompanied by pitting. Slow strain-rate experiments showed that the propagation rate of stress corrosion cracks also depends on the size of the solvent molecule. From these results it may be inferred that the cause of the variation in the velocity is the steric hindrance of the alcohol molecules. The surface mobility SCC mechanism may account for these results. (author)

  16. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    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. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  17. Mechanical properties of zircaloy-4 tubes for CAREM 25 fuel rods

    International Nuclear Information System (INIS)

    Juarez, G; Bianchi, D; Flores, A; Vizcaino, P

    2012-01-01

    The aim of the present work was giving support to the development of Zircaloy-4 fuel claddings for the CAREM 25 reactor through microstructural and mechanical properties studies along the manufacturing process. The manufacturing route was defined in 4 cold rolling stages and two thermal treatments, one at the middle and one after the last rolling stage. The first two rolling stages were performed in FAESA and the last two in PPFAE-CNEA using the rolling machine HPTR 8-15. The reference values for the evaluation were those indicated in the technical specification CAREM25 F ET-3-B0610. In this context, four tubes were received from FAESA. To these tubes mechanical properties determinations were performed to characterize the material in each step performed in PPFAE. The mechanical properties of the cladding tubes also achieve the standard values (σ 0.2 = 450 MPa, e = 15%), being σ 0.2 = 530 MPa and 18% the elongation (author)

  18. Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam

    International Nuclear Information System (INIS)

    Pawel, R.E.; Cathcart, J.V.; Campbell, J.J.; Jury, S.H.

    1977-12-01

    A series of scoping tests to determine the influence of steam pressure on the isothermal oxidation kinetics of Zircaloy-4 PWR tubing was undertaken. The oxidation experiments were conducted in flowing steam at 3.45, 6.90, and 10.34 MPa (500, 1000, and 1500 psi) at 905 0 C (1661 0 F), and at 3.45 and 6.90 MPa at 1101 0 C (2014 0 F). A comparison of the results of these experiments with those obtained for oxidation in steam at atmospheric pressure under similar conditions indicated that measurable enhancement of the oxidation rate occurred with increasing pressure at 905 0 C, but not at 1100 0 C

  19. Flow stress and dynamic strain-ageing of β-transformed Zircaloy-4

    International Nuclear Information System (INIS)

    Woo, O.T.; Tseng, D.; Tangri, K.; MacEwen, S.R.

    1979-01-01

    The 0.2% yield stress of β-transformed Zircaloy-4 was found to be independent of prior-β grain size but varied as the inverse of the transformed β plate width. A dislocation loop expansion model originally proposed by Langford and Cohen (1969) for cold-drawn iron wires is used to explain the inverse plate width dependence. Both air-cooled and water-quenched samples exhibited dynamic strain-ageing effects in approximately the same temperature range of 573 to 673 K: (a) a local minimum in strain-rate sensitivity is associated with a peak or an inflection point in the temperature dependence of the 0.2% yield stress for water-quenched or air-cooled samples respectively, and (b) yield drops were observed in strain rate change tests. (Auth.)

  20. Hot hardness studies on zircaloy 2 pressure tube along three orientations

    International Nuclear Information System (INIS)

    Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

    2002-01-01

    Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (author)

  1. Electrochemical behavior of thin anodic oxide films on Zircaloy-4: Role of the mobile defects

    International Nuclear Information System (INIS)

    Salot, R.; Lefebvre-Joud, F.; Baroux, B.

    1996-01-01

    The first stages of the electrochemical oxidation of Zircaloy-4 are investigated using simple electrochemical tests and modeling the passive film modifications occurring as a result of contact with the electrolyte. Variations in electrode potential (open-circuit conditions) or current density (potentiodynamic scans) can be simply explained by a high field (F ∼ 10 6 V/cm) assisted passive film growth. Under open-circuit conditions, this field does not vary with exposure time (in the 2 h to 48 h range). The minimum electric field for the onset of high-field behavior is also evaluated and found smaller than the theoretical value which can be explained by a variation in the concentration of mobile defects throughout the film. Measurements of the electrode potential decay after a potentiodynamic scan confirm this model, allowing interpretation of the film modification as a combination of two separate phenomena: film growth under a high electric field and point defect annihilation

  2. La grammaticalisation du monde.

    Directory of Open Access Journals (Sweden)

    Etienne Pingaud

    2010-01-01

    Full Text Available Ouvrage atypique par le fond comme par la forme, Le devoir et la grâce rend compte du minutieux travail d’élaboration théorique auquel s’attelle Cyril Lemieux depuis plusieurs années. Et le résultat final se veut pour le moins ambitieux : l’auteur propose un système total supposé dépasser d’un même élan le relativisme, le mentalisme, l’universalisme ethnocentrique, l’historicisme, le naturalisme et l’herméneutisme, tout en réconciliant les sciences sociales avec le ...

  3. Le sacre du printemps

    Directory of Open Access Journals (Sweden)

    Denise Pumain

    2002-03-01

    Full Text Available Cybergeo aura six ans en avril : dans la réalité du virtuel, dans l'univers récent et fluctuant de la publication en ligne, cela fait de nous, tout à la fois, des pionniers et des vétérans. De façon plus surprenante, il se trouve que nous sommes aussi uniques : parmi toutes les revues électroniques de sciences sociales, aucune ne combine comme Cybergeo ancienneté, publication exclusivement électronique, liberté d'accès au texte intégral, édition et gestion par des chercheurs, et comité de lec...

  4. Le Brahmane du Komintern

    Directory of Open Access Journals (Sweden)

    Elizabeth Burgos

    2008-01-01

    Full Text Available Le Brahmane du Komintern, largometraje documental del realizador francés Vladimir León, constituye un ejercicio ejemplar de investigación histórica y  de lograda factura de realización. Y, pese a no haber contado con la ayuda de ninguno organismo público, se trata de un ambicioso proyecto que cubre una amplia extensión geográfica que abarca: Estados Unidos, México, Moscú, Berlín, y la India. Gira en torno a una figura que tuvo en su tiempo su hora de gloria. Un bengalí, hijo de braman, la c...

  5. (Sorghum bicolor (L.) Moench) du Nord du Burkina Faso

    African Journals Online (AJOL)

    SARAH

    29 déc. 2014 ... sorghos à grains sucrés ont un cycle court et arrivent donc à maturité avant les autres sorghos et le mil d'où leur exploitation comme aliment de soudure par les paysans. L'organisation de la diversité morphologique des accessions de sorghos à grains sucrés du Nord du. Burkina autour principalement des ...

  6. Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2014-08-15

    Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

  7. Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4

    International Nuclear Information System (INIS)

    Racine, A.

    2005-09-01

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

  8. Stochastic model of texture dependence of iodine SCC susceptibility of a zircaloy-2 alloy

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Nakajima, Shinichi; Node, Shunsaku; Fujisawa, Takashi; Minamino, Yoritoshi

    1991-01-01

    Effects of textures on statistical parameters of tensile elongations in stress corrosion cracking (SCC) of zircaloy-2 using a slow strain rate test (SSRT) method have been investigated by Weibull distribution method based on stochastic process theory. The SCC is analyzed by assuming a probabilistic state transition model. Tensile directions of test pieces were prepared parallel, 45deg and perpendicular to rolling direction of the sheet. The test pieces in evacuated silica tubes were annealed at 1073K for 7.2x10 3 s, and then quenched into ice water. The annealed pieces with tilt angle α between tensile direction and a basal plane {0001} were 0, 18 and 25deg respectively. The tensile elongations of zircaloy-2 in SCC using the SSRT method are found to obey the single Weibull distribution with location parameters, and the SCC phenomena can be described by the Weibull distribution based on the stochastic process. The values of scale parameter η decrease with the tilt angle α, and the SCC susceptibility can be indicated by the values of scale parameter η. The texture dependence of the values of shape parameters m shows the changes of corrosion process in iodine solution and deformation system in air which are observed in the SSRT. The mechanism of decrement in the SCC susceptibility changes with the tilt angle α. The SCC under SSRT method is found to obey the model of probabilistic state transition. The constant load SCC process which obey the model of probabilistic state transition, is found to be effective for estimation of accelerated SCC condition. (author)

  9. Combined effects of radiation damage and hydrides on the ductility of Zircaloy-2

    International Nuclear Information System (INIS)

    Wisner, S.B.; Adamson, R.B.

    1998-01-01

    Interest remains high regarding the effects of zirconium hydride precipitates on the ductility of reactor Zircaloy components, particularly in irradiated material. Previous studies have reported that ductility reductions are much greater at room temperature compared to reactor component temperatures. It is often concluded that the effects of irradiation dominate the ductility reduction observed in test specimens, although there is no consensus as to whether hydriding effects are additive. Many of the tests reported in the literature are difficult to interpret due to variations in test specimen geometry and material history. In this paper, we present the results of an experimental program aimed at clearly describing the combined effects of irradiation and hydriding on ductility parameters under conditions of a realistic test specimen design and well characterized hydride content, distribution and orientation. Experiments were conducted at 295 and 605 K, respectively on Zircaloy-2 tubing segments containing 10-800 ppm hydrogen and neutron fluences between 0.9 x 10 25 nm -2 (E>1 MeV). Tests utilized the well proven localized ductility specimen which applies plane strain tension in the hoop direction of the tubing segment. In all cases, hydrides were also oriented in the hoop or circumferential direction and were uniformly distributed across the tubing wall. Results indicate that at 605 K, the ductility of irradiated material was almost independent of hydride content, retaining above 4% uniform elongation and 25% reduction in an area for the highest fluences and hydrogen contents. Even at 295 K, measurable ductility was retained for irradiated material with up to 600 ppm hydrogen. In the paper, results of fractographic analyses and strain rate are also discussed

  10. Study on the microstructure of recycled zircaloy by X-ray diffraction line profile analysis

    International Nuclear Information System (INIS)

    Ichikawa, Rodrigo U.; Pereira, Luiz A.T.; Imakuma, Kengo; Martinez, Luis G.; Turrillas, Xavier

    2013-01-01

    In the fabrication of nuclear fuel elements parts, Zircaloy machining chips are generated and, as this material is high-valued and controlled, its recycling presents high interest not only in economic aspects but also for environmental reasons and due to its strategic role in nuclear technology. Two processes for the recovery of these Zircaloy chips are being studied at IPEN-CNEN/SP. One of the processes is by conventional remelting of the material in a VAR (Vacuum Arc Remelting) furnace for producing solid ingots. Concurrently it is being studied an alternative process, by powder metallurgy methods, by which the chips are hydrided in order to become brittle and be grinded. The resulting ground powder is then compacted and finally vacuum-dehydrided and sintered in one step to form solid pieces. The VAR-remelted samples were also submitted to heat treatments in order to refine their microstructures, resulting in three different samples named 'as cast', 'annealed' and 'tempered'. The microstructures resulting from both processes and also from heat treatments were studied by metallography and X-ray diffraction (XRD). In this work, results of a XRD study are presented applying X-ray diffraction Line Profile Analysis (XLPA) methods in order to determine the mean crystallite sizes and the RMS microstrains on these samples. Additionally, a study for verify the influence of different standard materials used for the correction of the instrumental breadth in the XLPA was developed. The XLPA results show the influence of the processes and also of heat treatments on mean crystallite sizes and microstrains of the samples and were compared to their metallographic study and hardness. (author)

  11. A regression approach for zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. From data analysis and model development point of views, both the assumption of independence and prior committment to specific model forms are unacceptable. One would desire means which can not only estimate the required parameters directly from data but also provide basis for model selections, viz., one model against others. Basic understanding of the physics of deformation is important in choosing the forms of starting physical model equations, but the justifications must rely on their abilities in correlating the overall data. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) when there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets, (2) regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections

  12. OPERATION DU FOISONNEMENT

    Directory of Open Access Journals (Sweden)

    Gholamreza Djelveh

    2010-04-01

    Full Text Available Mousses alimentaires sont un sous-ensemble des aliments connus sous le nom de produits fouettés ou des produits aérés. Ils sont des produits formulés avec des qualités telles que la légèreté et la souplesse et sont principalement consommés à l'apéritif ou au dessert. Les produits en mousse obtenue par dispersion d'un gaz dans une matrice alimentaire (la phase continue ont connu un développement croissant au cours des années 80 et 90. Le processus d'aération liés à leurs activités de production est appelée l'expansion ou à fouetter. Le document présente les principaux-paramètres du procédé du point permanent de la formulation, la mise en œuvre processus dans les installations pilotes et à l'échelle industrielle, la caractérisation des produits finis, la base énergétique de l'échelle de processus en place, et le lien entre la formulation, émulsion préparation de l'expansion. Cette vue d'ensemble de l'opération d'expansion continue, nous a permis de mettre en évidence le fait qu'il ya des opérations de l'unité encore mal décrite par le génie des procédés et pour lesquels les méthodes et outils pour l'extrapolation et la prédiction sont encore à leurs balbutiements.

  13. Alimentation du nouveau-ne et du nourrisson dans la region ...

    African Journals Online (AJOL)

    Alimentation du nouveau-ne et du nourrisson dans la region centrale du togo : pratiques familiales et communautaires avant la mise en oeuvre de la strategie « prise en charge integree des maladies de l'enfant »

  14. Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

    International Nuclear Information System (INIS)

    Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

    1992-01-01

    Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

  15. Automatic identification and normalization of dosage forms in drug monographs

    Science.gov (United States)

    2012-01-01

    Background Each day, millions of health consumers seek drug-related information on the Web. Despite some efforts in linking related resources, drug information is largely scattered in a wide variety of websites of different quality and credibility. Methods As a step toward providing users with integrated access to multiple trustworthy drug resources, we aim to develop a method capable of identifying drug's dosage form information in addition to drug name recognition. We developed rules and patterns for identifying dosage forms from different sections of full-text drug monographs, and subsequently normalized them to standardized RxNorm dosage forms. Results Our method represents a significant improvement compared with a baseline lookup approach, achieving overall macro-averaged Precision of 80%, Recall of 98%, and F-Measure of 85%. Conclusions We successfully developed an automatic approach for drug dosage form identification, which is critical for building links between different drug-related resources. PMID:22336431

  16. Fourteen days oral administration of therapeutic dosage of some ...

    African Journals Online (AJOL)

    Fourteen days oral administration of therapeutic dosage of some antibiotics reduced serum testosterone in male rats. FO Awobajo, Y Raji, II Olatunji-Bello, FT Kunle-Alabi, AO Adesanya, TO Awobajo ...

  17. Buccal Dosage Forms: General Considerations for Pediatric Patients.

    Science.gov (United States)

    Montero-Padilla, Soledad; Velaga, Sitaram; Morales, Javier O

    2017-02-01

    The development of an appropriate dosage form for pediatric patients needs to take into account several aspects, since adult drug biodistribution differs from that of pediatrics. In recent years, buccal administration has become an attractive route, having different dosage forms under development including tablets, lozenges, films, and solutions among others. Furthermore, the buccal epithelium can allow quick access to systemic circulation, which could be used for a rapid onset of action. For pediatric patients, dosage forms to be placed in the oral cavity have higher requirements for palatability to increase acceptance and therapy compliance. Therefore, an understanding of the excipients required and their functions and properties needs to be particularly addressed. This review is focused on the differences and requirements relevant to buccal administration for pediatric patients (compared to adults) and how novel dosage forms can be less invasive and more acceptable alternatives.

  18. Dosage compensation of serine-4 transfer RNA in Drosophila melanogaster

    International Nuclear Information System (INIS)

    Birchler, J.A.; Owenby, R.K.; Jacobson, K.B.

    1982-01-01

    A dosage series of the X chromosome site for serine-4 transfer RNA consisting of one of three copies in females and one to two in males was constructed to test whether transfer RNA expression is governed by dosage compensation. A dosage effect on the level of the serine-4 isoacceptor was observed in both females and males when the structural locus was varied. However, in males, each dose had a relatively greater expression so the normal one dose was slightly greater than the total female value and the duplicated male had the highest relative expression of all the types examined. Serine-4 levels in males and females from an isogenic Oregon-R stock were similar. Thus the transfer RNA levels conform to the expectations of dosage compensation

  19. Study of free acidity determinations in aqueous solution; Etude des dosages d'acidite libre en solution aqueuse

    Energy Technology Data Exchange (ETDEWEB)

    Kergreis, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1966-04-01

    The object of this work is the study of the principal methods which can be applied to the measurement of 'free' acidity. In the first part, we define the various types of acidity which can exist in aqueous solution; then, after having studied some hydrolysis reactions, we compare the value of the neutralisation pH of the hydrated cation and that of the precipitation of the hydroxide. In the second part we have started to study the determination of the acidity of an aqueous solution. After having rapidly considered the 'total' acidity determination, we deal with the problem of the 'free' acidity titration. We have considered in particular certain methods: extrapolation of the equivalent point, colorimetric titrations with or without a complexing agent, and finally the use of ion-exchange resins with mixed aqueous and solvent solutions. (author) [French] Le but de ce travail est l'etude des principales methodes de determination de l'acidite 'libre'. Dans une premiere partie nous avons defini les differentes sortes d'acidites pouvant exister en solution aqueuse, puis apres avoir etudie quelques reactions d'hydrolyse, nous avons compare la valeur de pH de neutralisation du cation hydrate et celle de precipitation de l'hydroxyde. Dans la seconde partie nous avons aborde l'etuce des dosages de l'acidite d'une solution aqueuse. Apres avoir envisage assez rapidement la determination de l'acidite 'totale', nous traitons du probleme du titrage de l'acidite 'libre'. Nous avons porte notre attention sur certaines methodes: extrapolation du point equivalent, titrimetrie colorimetrique avec ou sans complexant, et enfin utilisation des resines echangeuses d'ions en milieu aqueux et solvant mixte.

  20. Quality Selection of Zircaloy-2 Canning Tubes by Ultrasonic Testing on Small Defects; Controle de la Qualite des Gaines en Zircaloy-2: Detection de Petits Defauts par les Ultrasons; Achestvennyjotb ortrub chatykh obolochek iz tsirkalloya-2 putem vyyavleniya nebol'shikh defektov s pomoshch'yu ul'trazvuka; Control de Calidad de los Revestimientos de Zircaloy-2 por Localizacion Ultrasonica de Pequenos Defectos

    Energy Technology Data Exchange (ETDEWEB)

    Van Der Linde, A. [Reactor Centrum Nederland, Petten (Netherlands); Deraad, J. A. [Roentgen Technische Dienst N.V., Rotterdam (Netherlands)

    1965-09-15

    Zircaloy-2 canning tubes, 10.20 mm I.D. x 0.90 mm wall x 1500 mm length, destined for testing as fuel-rod cladding in a high temperature. 330 Degree-Sign C, in-pile pressurized water loop, were tested ultrasonically for defects to get an impression of the tubes' quality. The tested tubes were delivered by manufacturers in the United States of America, United Kingdom and Scandinavia. Our requirement that all delivered tubes should be free from defects with a length greater than 500- 1000 {mu}m and/or with a depth greater than 50-25 {mu}m was not completely accepted by the manufacturers. They could guarantee that defects longer than 1000 {mu}m and/or with a depth greater than 50 {mu}m should be absent. Because only two of the 93 tested tubes had defects with a depth greater than 50 {mu}m it was decided to apply a more severe test by which defects with a depth in the range 10-50 {mu}m could be detected. To detect and record such small defects, longitudinal as well as transverse, a semi-automatic ultrasonic pulse equipment was used in combination with gating systems and a multiple-channel recorder. The adjustment of the scanning system was such that inner and outer defects of the same size were indicated with equal amplitudes. Calibration of the equipment was made on artificial defects. Longitudinal defects were detected with a separated transmitter-receiver system using a focused beam. Transverse defects were scanned by a single probe acting as transceiver. To obtain the sensitivity required the tests were carried out in immersion at a frequency of 4 MHz whereby the tubes were rotated with 120 rpm. A description is given of the mechanical device, the general set-up and the difficulties encountered. The result was that from the 93 tested tubes 21 had defects in the transverse direction with a depth between 10 and 50 {mu}m. Thus a relative qualification of the tubes was obtained. (author) [French] Des gaines en Zircaloy-2, ayant un diametre interieur de 10, 20 mm, une