Preliminary studies of the kinetics of a reactor by the probability method; Etude preliminaire de la cinetique d'un reacteur par la methode des probabilites

Energy Technology Data Exchange (ETDEWEB)

Bruna, J G; Brunet, J P; Clouet D' Orval, Ch; Caizergues, R; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The {alpha} decay constant of prompt neutrons has been studied in the homogeneous plutonium-fueled, light-water-moderated reactor Alecto, by the probability method. In this method, the probability to count one, two,.... neutrons during a given time is measured. The value of {alpha} can be deduced from this measurement, for various subcritical states of the reactor. The experimental results were then compared with values obtained, for the same reactivities, by the pulsed neutron technique. (authors) [French] On a etudie sur Alecto, reacteur homogene au plutonium, modere a l'eau legere, la constante de decroissance {alpha} des neutrons prompts par la methode des probabilites. Celle-ci consiste a mesurer la probabilite de compter un, deux, etc..., neutrons pendant un intervalle de temps donne. On a pu en deduire la valeur de {alpha}, dans divers etats sous-critiques du reacteur. On a compare les resultats experimentaux a d'autres valeurs obtenues, aux memes reactivites, par la methode des neutrons pulses. (auteurs)

Burst slug detection system in french power reactors (1961); La detection des ruptures de gaines dans les reacteurs de puissance francais (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

Gas samples are taken from the channels of the reactor and the short lived fission products are electrostatically collected to be analysed by a phosphor and photomultiplier system. The electrostatic collection and rotating electrode detector is described and its main uses exposed. Experience has shown the interest of measuring the evolution of fission products activities and not their absolute value only. In this way, data processing equipment have been designed and adapted to the detection apparatus. The system developed and realized for the G-l - G-2 - G-3 - EDF-1 - EDF-2 reactors are compared. (authors) [French] Un prelevement de gaz est effectue dans les canaux du reacteur et les produits de fission a vie courte sont collectes electrostatiquement pour etre analyses par un ensemble scintillateur-photomultiplicateur. Le detecteur a collection electrostatique et electrode tournante est decrit et ses applications principales sont exposees. L'experience a montre l'interet de mesurer l'evolution des activites en produits de fission et non seulement leur valeur absolue. D'ou le developpement d'ensembles de traitement des informations associes aux chaines de detection. Comparaison des realisations sur les reacteurs G-l - G-2 - G-3 - EDF-1 et EDF-2. (auteurs)

Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

Science.gov (United States)

Mahjoub, Mehdi

La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type

[Present conceptions of the C.E.A. concerning] the development of fast neutron reactors in France; [Les conceptions actuelles du C.E.A. concernant] la filiere des reacteurs a neutrons rapides en France

Energy Technology Data Exchange (ETDEWEB)

Vendryes, G; Gaussens, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Pasquer, R [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

. (authors) [French] 1 - Situation des reacteurs a neutrons rapides dans le programme d'energie nucleaire francais. En developpant un programme base sur l'uranium naturel, la France se trouvera dotee d'un stock important de plutonium riche on isotopes superieurs. L'existence de ce plutonium et de l'uranium appauvri provenant des memes reacteurs a pour consequence logique leur emploi dans des reacteurs a neutrons rapides. Justifiee par cet interet a court terme, la mise au point de reacteurs a neutrons rapides repond par ailleurs a une necessite pour l'avenir. 2 - Enonce des caracteristiques d'une centrale a neutrons rapides de 1000 MW el. Nous indiquons les caracteristiques d'une future centrale a neutrons rapides chargee au plutonium et refroidie au sodium. Si incertaines qu'elles soient, elles constituent un guide necessaire a l'orientation de nos travaux. 3 - Etudes effectuees a ce jour: Nous donnons un apercu des etudes souvent tres preliminaires qui ont permis de retenir les caracteristiques citees plus haut. Les principaux domaines techniques abordes sont les suivants: - Neutronique (masses critiques, taux de regeneration, enrichissements, aplatissement du flux de neutrons, coefficients de reactivite, evolution de la reactivite en fonction de l'irradiation), - Dynamique, controle et surete, - Combustible, - Technologie (conception du bloc-pile, des circuits de sodium, des dispositifs pour la manutention des assemblages). Ces etudes techniques se completent de considerations economiques. Le choix de caracteristiques optimales est lie a l'existence de programmes de production d'electricite et, dans ces programmes, a celle des reacteurs a neutrons thermiques producteurs de plutonium. On montre comment il y a lieu de tenir compte de l'existence du plutonium dans ce contexte, et quels sont les mecanismes qui rattachent l'economie de ce plutonium au choix des parametres essentiels des reacteurs surgenerateurs. 4 - Reacteur prototype: On justifie l'interet d'une etape

G2 - G3 inventive properties, the first french nuclear plants; Caracteristiques generales et aspects originaux des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Pascal,; Horowitz,; Bussac,; Joatton,; de Meux, De Lagge; Martin, [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

This paper points out the inventive properties of the frenchctors G2 and G3. These are dual purpose reactors, i.e. designed for the production of both plutonium and energy (30 electrical MW); in this respect, they can be considered as the start point of the french electrical energy produced from nuclear fuel. The following points are specially discussed in this paper: the choice of the prestressed concrete pressure vessel, the horizontal arrangement of the channels, the interest of neutron flux flattening, the advantages of the charging and discharging device working during pile operation. (author)Fren. [French] Les caracteres originaux des reacteurs fran is G2 et G3 sont decrits dans ce rapport. Ce sont des reacteurs a double fin, plutonigenes et aussi producteurs d'energie (30 MW electriques); ils constituent a ce titre le point de depart de la production fran ise d'electricite d'origine nucleaire. Sont discutes, en particulier, dans ce rapport: le choix du caisson en beton precontraint pour tenir la pression, la disposition horizontale des canaux, l'interet de l'aplatissement du flux neutronique, les avantages de l'appareil permettant le chargement et le dechargement du combustible sans arreter la pile. (auteur)

The functioning of the reactors G2-G3 at Marcoule and E.D.F. 1; Experience de fonctionnement des reacteurs G2-G3 de Marcoule et enseignements des essais de demarrage du reacteur E.D.F. 1 de Chinon

Energy Technology Data Exchange (ETDEWEB)

Boussard, R; Conte, F [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Stolz, J M [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

After resuming briefly the characteristics of the installations G2-G3 at Marcoule and EDF 1 at Chinon, the authors review the main aspects of the tests, the starting and the exploitation of these reactors. Among the various points examined, particular emphasis is given to the devices of original nature such as tubular fuel elements, flattening of the neutron flux by stuffing, behaviour of the reactor tanks and the cooling circuits, the blowers, unloading devices, regulation and functioning of the informations. This analysis deals equally with the performances obtained and the difficulties and the various incidents experienced during the initial starting period. Among the more interesting results, the progressive increase in the power of the Marcoule reactors is mentioned, obtained through a better knowledge of the parameters covering the functioning of the reactors such as the distribution of the flux and the temperatures etc... acquired during the course of the exploitation of the reactor. The conclusion reached by the authors is that the experience gained on these installations has shown: - that during an initial period, adjustments became necessary, all of which turned out to be possible, - that an analysis of their functioning has permitted the progressive movement towards a truly industrial exploitation. (authors) [French] Les auteurs, apres un bref rappel des caracteristiques des installations G2 - G3 de MARCOULE et E.D.F. 1 de CHINON, passent en revue les principaux aspects des essais, de la mise en service et de l'exploitation de ces centrales. Parmi les divers points examines, une attention speciale est accordee aux dispositifs presentant un caractere original tels que elements combustibles tubulaires, aplatissement du flux neutronique par gavage, comportement des caissons des reacteurs et des circuits de refroidissement, soufflantes, appareils de dechargement, regulation et fonctionnement des informations. L'analyse presentee porte tant sur les

Dynamic problems of power reactors and analogic devices; Les problemes dynamiques du reacteur de puissance et les machines analogiques

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

The raise of the nuclear physics came with heavy mathematical developments. The analogical installations became especially useful for precise calculations of parameters which depend the running of a reactor. They permit between other to study of kinetic problems and especially ''cybernetics'' of nuclear reactors. It doesn't make a doubt that their use will become widespread, not only in the calculations laboratories, in services for servo-mechanisms study, but also in the control panels of the reactors themselves. (M.B.) [French] L'essor de la physique nucleaire s'est accompagne de lourds developpements mathematiques. Les montages analogiques sont devenus particulierement utiles pour les calculs precis des parametres dont depend le fonctionnement d'un reacteur. Elles permettent entre autre l'etude des problemes cinetiques et surtout ''cybernetiques'' des reacteurs nucleaires. Il ne fait pas de doute que leur usage se generalisera, non seulement dans les laboratoires de calculs, les services d'etudes de servomecanismes, mais aussi pres des tableaux de commande des reacteurs eux-memes. (M.B.)

Experimental methods of reactor physics; Methodes experimentales de physique des reacteurs a neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Breton, D; Lafore, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

This paper is a synthesis of various experimental methods in use with the reactors of the Commissariat a l'Energie Atomique. The main techniques used are mentioned and the difficulties encountered and the accuracy obtained are particularly dwelt upon. The application of these various methods to reactors in order to obtain specific results is also indicated. This paper consists of five parts. I - General methods. Macroscopic and microscopic flux distribution (anisotropy effect), power distribution, etc... II - Kinetic measurements a) pulsed neutron technique: apparatus and accuracy; application to {lambda}t and to anti reactivity measurements; application to graphite, light water and beryllium oxide. b) oscillation techniques: equipment and accuracy; application to the measurements of effective cross sections and resonance integrals. c) fluctuations: apparatus and technique of measurement. III - Poison methods. Description of methods for introducing and extracting the poison, difficulties encountered with light and heavy water, measurement of temperature coefficients and anti-reactivity. IV - Spectra measurements. Choice and development of foils, problems of measurement, application to spectral measurements for thermalization studies, application to dosimetry. V - Experimental shielding measurements. The technique and apparatus recently developed in this field are presented. (authors) [French] Cette communication fait une synthese des differentes methodes experimentales mises en oeuvre sur les reacteurs du CEA. Elle presente les principales techniques utilisees et insiste plus particulierement sur les difficultes rencontrees et la precision obtenue; elle indique egalement l'application de ces differentes methodes sur les reacteurs, en vue de l'obtention des resultats determines. Elle comporte cinq parties: I - METHODES GENERALES: Distribution de flux macroscopique et microscopique (effet d'anisotropie), distribution de puissance, etc... II - MESURES CINETIQUES: a

Presence of Tritium in the Cooling Circuits of the Reactors G2 and G3; Presence de tritium dans les circuits de refroidissement des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Estournel, R [Commissariat a l' Energie Atomique. Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

1962-07-01

In a reactor of the G 2-G 3 type, tritium can be formed by the neutronic bombardment of many elements present in the core. Tritium was found to be present in the cooling circuits of the reactors G 2 and G 3 in the water coming from the regeneration of the CO{sub 2} dehydrating columns. (author) [French] Dans un reacteur du type G 2 - G 3, le tritium peut etre forme par le bombardement. neutronique de nombreux elements existant dans le c r. La presence de tritium dans les circuits de refroidissement des reacteurs G 2 - G 3 a ete mis en evidence dans l'eau provenant de la regeneration des colonnes de deshydratation du CO{sub 2}. (auteur)

Contribution to the study of the stability of water-cooled reactors; Contribution a l'etude de la stabilite des reacteurs refroidis par de l'eau

Energy Technology Data Exchange (ETDEWEB)

Coudert, C [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1969-06-01

This work is devoted to the study of the stability of reactors cooled by water subjected only to natural convection. It is made up of two parts, a theoretical study and experimental work, each of these parts being devoted to a consideration of linear and non-linear conditions: - calculation of the transfer function of the reactor using neutronic and hydrodynamic linear equations with the determination of the instability threshold; - demonstration of the existence of the limiting oscillation cycle in the case of a linear feedback using MALKIN'S method; - measurement and interpretation of the reactor's transfer functions and of the hydrodynamic transfer functions; and - analysis of the noise due to boiling. (author) [French] Dans ce travail on etudie la stabilite des piles refroidies par de l'eau circulant en convection naturelle. Cette etude se divise en deux parties: un travail theorique et un travail experimental, chacune de ces parties comportant une etude lineaire et une etude non-lineaire: - calcul de la fonction de transfert du reacteur a partir des equations lineaires de la neutronique et de l'hydrodynamique avec determination du seuil d'instabilite; - demonstration de l'existence du cycle limite des oscillations dans le cas d'une retroaction lineaire en utilisant la methode de MALKIN; - mesure et interpretation de la fonction de transfert du reacteur et des fonctions de transfert hydrodynamiques; et - analyse du bruit d'ebullition. (auteur)

Detection of burst cans in the reactors cooled by gaseous phase; Detection des ruptures de gaine dans les reacteurs refroidis par phase gazeuse

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

In a nuclear reactor including the bars or plates cooled by a gaseous fluid, burst risks to occur in the sheath assuring the tightness separation between the cooling gas and the fissile materials. It is necessary to be able to detect the formation of these cracks as possible in order to avoid all risk of fission products release or any reaction of uranium to the contact of the refrigerating gas. It is however the increase of the radioactivity in the cooling gas due to the scattering of the fission products that permits to signal the apparition of a crack or to follow its evolution. It is possible to detect cracks of the order of the square millimeter. In this report, we will detail the principle and the realization of a device used for the surveillance of a natural uranium reactor cooled by air circulation. (M.B.) [French] Dans un reacteur nucleaire comportant des barres ou des plaques refroidies par un fluide gazeux des fissures risquent de se produire dans les gaines assurant la separation etanche entre le gaz de refroidissement et les materiaux fissiles. II est necessaire de pouvoir detecter la formation de ces fissures des que possible afin d'eviter tout risque de liberation de produits de fission ou de reaction de l'uranium au contact du gaz refrigerant. C'est cependant l'augmentation de la radioactivite du gaz de refroidissement due a la dispersion des produits de fission qui permet de signaler l'apparition d'une fissure ou de suivre son evolution. On peut ainsi detecter des fissures de l'ordre du millimetre carre. Dans ce rapport, nous detaillerons le principe et la realisation d'un appareil utilise pour la surveillance d'un reacteur a uranium naturel refroidi par circulation d'air. (M.B.)

Neutron detection in an atomic reactor core using semi-conductors; Detection des neutrons par semi-conducteur dans un coeur de reacteur atomique

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

In this paper, the first part describes the principle of nuclear particle detection by means of semiconductor diodes and the general application of these. The second part describes fabrication of the device used to estimate thermic neutron fluxes in core of a swimming pool type reactor. The useful volume (2.9 mm thickness) is in the light water moderator, between combustible elements plates. The results, principally obtained in the core of Siloette reactor at the 'Centre d'Etudes Nucleaires de Grenoble' at low power, are mentioned in the third part. Flux maps have been set and comparison between converter's products: Bore 10, Lithium 6, Uranium 235 is made. (author) [French] Dans ce rapport, une premiere partie porte sur la description du principe de detection des particules nucleaires par diodes a semi-conducteur et sur l'application generale de celles-ci. Une deuxieme partie s'attache a decrire la fabrication du materiel utilise pour evaluer les flux de neutrons thermiques dans un coeur de reacteur type pile piscine. L'espace de mesure (2,9 mm d'epaisseur) se situe entre les plaques des elements combustibles, dans le moderateur eau legere. Les resultats, obtenus principalement dans le coeur du reacteur Siloette du Centre d'Etudes Nucleaires de Grenoble aux basses puissances de fonctionnement, sont rapportes dans la troisieme partie. Des cartes de flux ont ete dressees et une comparaison est faite entre les produits 'convertisseurs' suivants: Bore 10, Lithium 6, Uranium 235. (auteur)

Study relating to the physico-chemical behaviour of heavy water in nuclear reactors; Etudes relatives au comportement physico-chimique de l'eau lourde dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Chenouard, J; Dirian, G; Roth, E; Vignet, P; Platzer, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1959-07-01

Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [French] Parmi les degradations subies par l'eau lourde dans les reacteurs nucleaires, les deux plus importantes sont la pollution chimique et isotopique et la decomposition radiolytique. La pollution chimique a conduit a mettre au point pour le cas particulier des reacteurs, des circuits d'epuration par echange d'ions. On decrit ici en detail la mise en oeuvre de cette methode dans les reacteurs de recherche du CEA; les controles qu'elle necessite, les resultats obtenus et leur interpretation. En ce qui concerne la dissociation radiolytique de l'eau, les renseignements obtenus sur ces memes reacteurs sont communiques, ainsi que les details des dispositifs de recombinaison et des moyens de controle. Enfin, on fait le point des etudes poursuivies au CEA sur ces memes problemes de recombinaison dans le cas des reacteurs de puissance. (auteur)

Economic Effect on the Plutonium Cycle of Employing {sup 235}U in Fast Reactor Start-Up; Incidence Economique du Demarrage des Reacteurs Rapides a l'Aide d'Uranium-235 sur le Cycle du Plutonium

Energy Technology Data Exchange (ETDEWEB)

Van Dievoet, J.; Egleme, M.; Hermans, L. [BELGONUCLEAIRE, Bruxelles (Belgium)

1967-09-15

factors, inventory factors) from one cycle to another, with a comparative study of the use of {sup 235}U in thermal and fast reactors, variations in the discounted fuel cycle costs from one cycle to another, and weight and characteristics of the recycled fuel, of the additional fuel required and of excess fuel. (author) [French] Le memoire presente les premiers resultats d'une etude entreprise dans le cadre d'un contrat d'association Euratom-Belgique et destinee a evaluer l'interet de l'alimentation de reacteurs rapides en uranium-235. Plusieurs possibilites se presentent pour le demarrage d'un reacteur rapide a l'aide d'uranium-235. 1. Le reacteur peut etre alimente en permanence avec de l'uranium enrichi, le plutonium produit servant a demarrer et a alimenter d'autres reacteurs; dans ce cas, l'uranium est recycle dans le reacteur en y ajoutant de l'uranium enrichi. 2. Le plutonium produit dans le reacteur peut etre partiellement recycle dans celui-ci, ainsi que l'uranium; dans ce cas, le reacteur se transforme progressivement en un reacteur au plutonium. Ces deux cas peuvent etre combines pour un reacteur a plusieurs zones d'enrichissement, ou l'on peut appliquer simultanement les deux politiques a des zones differentes, c'est-a-dire: alimenter, par exemple, la zone interne en uranium enrichi et recycler le plutonium dans la zone externe. Le mode de traitement du combustible irradie rend egalement le probleme complexe, selon que l'on traite ensemble ou separement le coeur et les couvertures axiales; de meme, pour un reacteur a plusieurs zones d'enrichissement, celles-ci peuvent etre traitees ensemble ou separement. Les calculs sont effectues a l'aide d'un code de calcul utilisant, pour lavpartie relative aux caracteristiques des reacteurs successifs, les coefficients d'equivalence definis par Baker and Ross et, pour la partie economique, la methode du cout actualise du cycle du combustible. Dans la premiere phase des travaux, une analyse approcheedu phenomene a ete

Study relating to the physico-chemical behaviour of heavy water in nuclear reactors; Etudes relatives au comportement physico-chimique de l'eau lourde dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Chenouard, J.; Dirian, G.; Roth, E.; Vignet, P.; Platzer, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1959-07-01

Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [French] Parmi les degradations subies par l'eau lourde dans les reacteurs nucleaires, les deux plus importantes sont la pollution chimique et isotopique et la decomposition radiolytique. La pollution chimique a conduit a mettre au point pour le cas particulier des reacteurs, des circuits d'epuration par echange d'ions. On decrit ici en detail la mise en oeuvre de cette methode dans les reacteurs de recherche du CEA; les controles qu'elle necessite, les resultats obtenus et leur interpretation. En ce qui concerne la dissociation radiolytique de l'eau, les renseignements obtenus sur ces memes reacteurs sont communiques, ainsi que les details des dispositifs de recombinaison et des moyens de controle. Enfin, on fait le point des etudes poursuivies au CEA sur ces memes problemes de recombinaison dans le cas des reacteurs de puissance. (auteur)

Concept of transfer functions for a nuclear reactor; Notion de fonction de transfert pour un reacteur nucleaire

Energy Technology Data Exchange (ETDEWEB)

Dalfes, Abdi [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires. Departement d' Electronique Generale, Service d' Electronique des Reacteurs

1966-07-01

The solution to the correlation equations are expressed in terms of the eigenvalues and Eigen-matrices of the transport operator, for a subcritical zero power reactor. This allows to define, for each point of the reactor and for detectors detecting neutrons of given velocities, correlation and transfer functions driven by the same white-noise source. A precise meaning is also given to the importance operator, which is the adjoin of the transport operator. (author) [French] La solution des equations regissant les matrices de correlation est exprimee en fonction des valeurs et matrices propres de l'operateur de transport pour un reacteur sous-critique et de puissance nulle. Ceci permet de definir, en chaque point du reacteur et pour des detecteurs repondant a des neutrons de vitesse definie, des fonctions de correlation et de transfert dont les entrees sont attaquees par une meme source de bruit blanc. Le role joue par l'operateur importance, adjoint de l'operateur de transport, est aussi precise. (auteur)

Automation of nonlinear calculations in the theory of fusion reactor; Automatisation des calculs non lineaires dans la theorie des reacteurs a fusion

Energy Technology Data Exchange (ETDEWEB)

Braffort, P; Chaigne, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1) Introduction: The difficulties of the formulation of the equations of phenomena occurring during the operation of a fusion reactor are underlined. 2) The possibilities presented by analog computation of the solution of nonlinear differential equations are enumerated. The accuracy and limitations of this method are discussed. 3) The analog solution in the stationary problem of the measurement of the discharge confinement is given and comparison with experimental results. 4) The analog solution of the dynamic problem of the evolution of the discharge current in a simple case is given and it is compared with experimental data. 5) The analog solution of the motion of an isolated ion in the electromagnetic field is given. A spatial field simulator used for this problem (bidimensional problem) is described. 6) The analog solution of the preceding problem for a tridimensional case for particular geometrical configurations using simultaneously 2 field simulators is given. 7) A method of computation derived from Monte Carlo method for the study of dynamic of plasma is described. 8) Conclusion: the essential differences between the analog computation of fission reactors and fusion reactors are analysed. In particular the theory of control of a fusion reactor as described by SCHULTZ is discussed and the results of linearized formulations are compared with those of nonlinear simulation. (author)Fren. [French] 1) Introduction. On souligne les difficultes que presente la mise en equation des phenomenes mis en jeu lors du fonctionnement d'un reacteur a fusion. On selectionne un certain nombre d'equations generalement utilisees et on montre les impossibilites analytiques auxquelles on se heurte alors. 2) On rappelle les possibilites du calcul analogique pour la resolution des systemes differentiels non lineaires et on indique la precision de la methode ainsi que ses limitations. 3) On decrit esolution analogique du probleme statique de la mesure du confinement de la decharge

Transient regimes in a heavy water reactor; Regimes transitoires dans un reacteur a eau lourde

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

We studied the variations of power and reactivity of a reactor when we raise in a continuous way the starting plates. During the subcritical regime (negative reactivity), the power is determined by reactivity and by the intensity of the sources of photo neutrons, produced during the previous work of the reactor. When, during the rise of the plates, the reactor, pass by the critical regime (zero reactivity), one notes that the reached power is independent of the initial reactivity. During the sur-critical regime (positive reactivity), the elevation of temperature of the uranium bars slows down the growth of reactivity due to the movements of the plates. The power stretches then toward a value that depends only on the regime of cooling of the reactor and the excess of the available reactivity. This survey permits to choose such a rise speed, that reactivity remains constantly lower to a value beyond which the piloting of the reactor becomes difficult. This result is not more valid, if the intensity of the sources is insufficient, what takes place during the first divergences and after a stop of long length. (author) [French] On etudie les variations de puissance et de reactivite d'un reacteur quand on leve d'une facon continue les plaques de demarrage. Pendant le regime subcritique (reactivite negative), la puissance est determinee par la reactivite et par l'intensite des sources de photoneutrons, produites pendant la marche anterieure du reacteur. Quand, au cours de la montee des plaques, le reacteur passe par le regime critique (reactivite nulle), on constate que la puissance atteinte est independante de la reactivite initiale. Pendant le regime surcritique (reactivite positive), l'elevation de temperature des barres d'uranium ralentit l'accroissement de reactivite due aux mouvements des plaques. La puissance tend alors vers une valeur qui ne depend plus que du regime de refroidissement du reacteur et de l'exces de la reactivite disponible. Cette etude permet de

A study of switch circuits for use as safety devices in nuclear reactors; Etude de circuits de commutation destines a la securite des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

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

1960-12-15

The author reviews briefly a few basic assemblies using electromagnetic relays for safety circuits in nuclear reactors; he then studies the use of static relays with a shorter time of response, based on impedance changes in a self-inductance consisting of a coil with a magnetic core having a rectangular hysteresis cycle. The author examines in particular the way in which it functions and the method of determining the parameters. (author) [French] L'auteur apres avoir examine sommairement en revue quelques montages de base des circuits de securite des reacteurs nucleaires utilisant des relais electromecaniques, etudie l'emploi des relais statiques a plus grande vitesse de reponse bases sur la variation d'impedance que presente une self-inductance realisee a l'aide d'une bobine enroulee autour d'un noyau magnetique a cycle d'hysteresis rectangulaire. En particulier, il en examine le mode de fonctionnement et la determination des parametres. (auteur)

The Pegase reactor loops; Les boucles du reacteur Pegase

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-07-01

After 4 years operation, experimentation and maintenance of the gas loops built especially for the nuclear fuel testing reactor Pegase, it appears desirable not only to gather together in a single document the essential characteristics and particularities of these devices and of their associated equipment, but also to give the reasons for the technical modifications and the way in which they were carried out; this is done here by the persons themselves who were responsible, day after day, for operating these loops. This essentially practically experience thus complements the careful research and preliminary testing carried out on these loops or on their prototypes. It should be of interest to those who deal with problems concerned with the design or operation of irradiation loops in experimental reactors or of similar equipment. (authors) [French] Apres 4 annees de fonctionnement, d'experimentation et d'entretien sur les boucles a gaz, construites specialement pour le reacteur d'essai des combustibles nucleaires Pegase, il a paru souhaitable non seulement de rassembler dans un meme document les caracteristiques et les particularites essentielles de ces dispositifs et des appareillages qui leur sont associes, mais aussi d'y preciser les raisons et les modalites des mises au point techniques, apportees par ceux qui, jour apres jour pendant cette periode, ont eu la charge de mettre en oeuvre ces boucles. Cette experience essentiellement pratique complete donc les etudes minutieuses et les essais preliminaires de ces boucles ou de leurs prototypes. Elle doit etre de quelque interet pour ceux qui sont confrontes aux problemes de conception ou d'exploitation de boucles d'irradiation dans des reacteurs experimentaux ou des dispositifs analogues. (auteurs)

Economic aspects of electricity and industrial heat generating reactors; Aspect economique des reacteurs produisant de l'electricite et de la chaleur industrielle

Energy Technology Data Exchange (ETDEWEB)

Gaussens, J; Moulle, N; Dutheil, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Aldebert, J [Institut National des Sciences et Techniques Nucleaires (INSTN), CEA Saclay, 91 - Gif sur Yvette (France)

1964-07-01

The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [French] L'interet economique de centrales nucleaires productrices d'electricite decroit lorsque la puissance decroit. Cependant, lorsqu'on associe une turbine a contrepression a un reacteur et qu'il est possible d'utiliser dans de bonnes conditions la chaleur residuelle, on peut montrer que dans certaines conditions assez realistes, des equipements nucleaires d'une puissance unitaire peu elevee peuvent etre competitifs avec des equipements conventionnels. Cette communication a donc pour but de mettre en evidence quelles sont ces conditions particulieres de rentabilite de l'energie nucleaire. Elles sont liees a la localisation de la centrale et a son contexte economique general, a la structure de la demande d'energie electrique et thermique a laquelle elle doit satisfaire, au niveau des couts des combustibles et des investissements

The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

Energy Technology Data Exchange (ETDEWEB)

Vautrey, L; Zaleski, C P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1962-07-01

In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

The CO{sub 2} cooling gas for the reactors G2/G3 (leaking, analysis, activity); Le CO{sub 2} de refroidissement des reacteurs G2/G3 (fuites, analyse, activite)

Energy Technology Data Exchange (ETDEWEB)

Meiffren, J; Dupay, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1965-07-01

The main objective of this study is to publicise the data obtained during five years operation of the reactor G2 and G3 at Marcoule as far as the cooling gas is concerned, from storage of reserves up to its slow escape into the atmosphere, and including all the stages of its practical use, its chemical examination, its nuclear behaviour and its possible physicochemical transformation. This work can not only yield information about the operations carried out at Marcoule but can also provide useful suggestions for improving the sealing and for decreasing the activity of the pressurized gas circuits in reactors similar to G2/G3. (authors) [French] Le but principal de cette etude est de diffuser les connaissances acquises au cours de cinq annees d'exploitation des reacteurs G2 et G3 de Marcoule en ce qui concerne le gaz de refroidissement, depuis son stockage d'appoint jusqu'a son echappement lent dans l'atmosphere, en passant par tous les stades de son utilisation pratique, de son etude chimique, de son comportement nucleaire, eventuellement de ses transformations physico-chimiques. Cette etude peut, non seulement renseigner sur les operations effectuees couramment a Marcoule, mais egalement donner des suggestions interessantes pour l'amelioration de l'etancheite et la diminution de l'activite des circuits de gaz en pression dans des reacteurs analogues a G2/G3. (auteurs)

General problems arising from the analogical resolution of the kinetic equations of nuclear reactors (1961); Problemes generaux poses par la resolution analogique des equations cinetiques des reacteurs nucleaires (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

The author reviews precisely the analogical techniques used for the resolution of the kinetic equations of nuclear reactors. Prior to this, he recalls the reasons which oblige physicians and engineers, even today, to use electronic machines in this domain. The author then considers the technological problems posed by the range of values which the various nuclear parameters adopt. In each case, he shows that a compromise is possible allowing an optimum precision. He compares the results to those obtained by arithmetic calculation and uses the examples chosen in a critical analysis of the present possibilities of the two methods of calculation. (author) [French] L'auteur cherche a faire un point aussi exact que possible des techniques analogiques utilisees pour resoudre les equations cinetiques des reacteurs nucleaires. Il rappelle auparavant les raisons pour lesquelles physiciens et ingenieurs sont obliges, encore aujourd'hui, de faire appel aux machines electroniques dans ce domaine. Puis il etudie les problemes technologiques que souleve le champ des valeurs prises par les differents parametres nucleaires. Dans chacun des cas, il montre l'existence d'un compromis qui permet d'atteindre une precision optimum. Il compare les resultats obtenus a ceux provenant de calculateurs arithmetiques et profite des exemples choisis pour faire une analyse critique des possibilites actuelles offertes par les deux modes de calcul. (auteur)

Economic aspects of electricity and industrial heat generating reactors; Aspect economique des reacteurs produisant de l'electricite et de la chaleur industrielle

Energy Technology Data Exchange (ETDEWEB)

Gaussens, J.; Moulle, N.; Dutheil, F. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Aldebert, J. [Institut National des Sciences et Techniques Nucleaires (INSTN), CEA Saclay, 91 - Gif sur Yvette (France)

1964-07-01

The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [French] L'interet economique de centrales nucleaires productrices d'electricite decroit lorsque la puissance decroit. Cependant, lorsqu'on associe une turbine a contrepression a un reacteur et qu'il est possible d'utiliser dans de bonnes conditions la chaleur residuelle, on peut montrer que dans certaines conditions assez realistes, des equipements nucleaires d'une puissance unitaire peu elevee peuvent etre competitifs avec des equipements conventionnels. Cette communication a donc pour but de mettre en evidence quelles sont ces conditions particulieres de rentabilite de l'energie nucleaire. Elles sont liees a la localisation de la centrale et a son contexte economique general, a la structure de la demande d'energie electrique et thermique a laquelle elle doit satisfaire, au niveau des couts des

Chemical elimination of alumina in suspension in nuclear reactors heavy water; Elimination de l'alumine en suspension dans l'eau lourde des reacteurs nucleaires par voie chimique

Energy Technology Data Exchange (ETDEWEB)

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

1967-02-01

Corrosion of aluminium in contact with moderating water in nuclear reactor leads to the formation of an alumina hydrosol which can have an adverse effect on the operation of the reactor. Several physical methods have been used in an attempt to counteract this effect. The method proposed here consists in the elimination of the aluminium by dissolution and subsequent fixation in the ionic form on mixed-bed ion-exchange resin. In order to do this, the parameters and the values of these parameters most favorable to the dissolution process have been determined. If the moderator is heavy water, the deuterated acid can be prepared by converting a solution in heavy water to a salt of the acid using a deuterated cationic resin. (author) [French] La corrosion de l'aluminium au contact de l'eau moderatrice des reacteurs nucleaires, donne lieu a la formation d'un hydrosol d'alumine nuisible au bon fonctionnement des reacteurs. Plusieurs methodes physiques ont ete mises en oeuvre pour pallier ces inconvenients. On propose ici d'eliminer l'alumine par solubilisation pour la fixer ensuite sous forme ionique par des resines echangeuses d'ions, en lit melange. A cette fin on determine les parametres et leurs grandeurs favorables a cette solubilisation. Si le moderateur est de l'eau lourde la preparation d'acide deutere peut etre effectuee par passage d'une solution en eau lourde a un sel de l'acide sur resine cationique deuteree.

Calculation of the working capital invested in fuel cycles and its interest charges (1963); Calcul des immobilisations financieres des cycles de combustible (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

All the processes undergone by the nuclear material, including the various steps of fuel element manufacturing and of irradiated fuel reprocessing lead to working capital investments varying with the type of reactor, that must be taken into account in the kWh cost calculation. The author deals with a calculation method called: 'present worth method' and gives some examples concerning reactors the main fuel of which being either natural uranium or enriched uranium or plutonium. He especially points out the importance these investments may take in the case of fast breeder reactors. (author) [French] L'ensemble des etapes parcourues par la matiere fissile comprenant les divers stades d'elaboration des elements combustibles et de leur traitement apres irradiation, implique des immobilisations financieres tres differentes d'un type de reacteur a l'autre, dont il convient de tenir compte dans le calcul du cout du kWh. L'auteur expose une methode de calcul dite 'd'actualisation des couts' et donne quelques exemples relatifs aux reacteurs utilisant l'uranium naturel, l'uranium enrichi et le plutonium comme combustible principal. Il montre en particulier l'importance que peuvent avoir ces immobilisations dans le cas des reacteurs surregenerateurs. (auteur)

The dangers of irradiate uranium in nuclear reactors; Les dangers de l'uranium irradie dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

The danger of the uranium cans sur-activated by the use in the nuclear reactors is triple: - Irradiation from afar, during manipulations of the cans. - Contamination of air when decladding. - Contamination of air by fire of uranium in a reactor in function The first two dangers are usual and can be treated thanks to the rules of security in use in the atomic industry. The third has an accidental character and claimed for the use of special and exceptional rules, overflowing the industrial setting, to reach the surrounding populations. (authors) [French] Le danger des cartouches d'uranium suractive par utilisation dans les reacteurs nucleaires est triple: - Irradiation a distance, lors des manipulations des cartouches. - Contamination de l'air au moment de leur degainage. - Contamination de l'air par incendie d'uranium dans un reacteur en fonctionnement. Les deux premiers dangers sont habituels et peuvent etre traites grace aux regles de securite en usage dans l'industrie atomique. Le troisieme revet un caractere accidentel et reclame l'emploi de regles speciales et exceptionnelles, debordant le cadre industriel, pour atteindre celui des populations environnantes. (auteurs)

Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-07-01

This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

A review of calculation methods for fast and intermediate reactors; Expose des methodes pour le calcul de reacteurs a neutrons rapides et intermediaires; Obzor metodov rascheta reaktorov na promezhutochnykh i bystrykh nejtronakh; Estudio panoramico de los metodos de calculo de los reactores rapidos e intermedios

Energy Technology Data Exchange (ETDEWEB)

Marchuk, G I [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

1962-03-15

This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author) [French] L'auteur examine la mise au point de methodes pour le calcul de reacteurs a neutrons rapides et intermediaires . Il decrit diverses manieres d'aborder les problemes des calculs sur la physique des reacteurs, notamment le calcul des effets de resonance. Il s'attache particulierement aux points suivants: systemes d'equations fondamentales et conjuguees a plusieurs groupes; diverses applications de la theorie des perturbations aux problemes de calculs sur la physique des reacteurs; methodes numeriques pour resoudre les equations fondamentales et conjuguees, voisines de la methode des harmoniques spheriques. L'auteur decrit ensuite une maniere d'appliquer la methode de la reponse aux problemes de la masse critique ainsi que des methodes pour le calcul de reacteurs ralentis a l'hydrogene. Il decrit les caracteristique s fondamentale s d'un modele de reacteur a un groupe effectif. (author) [Spanish] El autor analiza el desarrollo de los metodos de calculo de los reactores nucleares que trabajan con neutrones rapidos y con neutrones intermedios. Examina diversos planteos de los problemas del calculo fisico. Indica la forma de tomar en cuenta los efectos de resonancia y menciona los sistemas

The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

Energy Technology Data Exchange (ETDEWEB)

Heineman, R. E. [General Electric Company, Richland, WA (United States)

1964-02-15

use is described in the light of the trends which are observed. (author) [French] Des mesures exponentielles sont faites aux laboratoires de Hanford sur des reseaux uranium-graphite depuis pres de quinze ans. Les resultats de ces experiences ont ete utilises pour determiner les laplaciens de reacteurs de production que l'on se proposait de construire, mais ils ont servi egalement a ameliorer les connaissances dans le domaine de la physique de ces systemes. On s'est rendu compte tres rapidement qu'en raison des dimensions des assemblages et de leur manque de sensibilite aux petites perturbations localisees du systeme, l'experience exponentielle n'a qu'une utilite limitee. On a donc envisage de mettre au point des experiences integrales avec un reacteur de maniere a reduire au minimum la quantite de matieres necessaires pour se procurer des donnees valables. A cet effet, on a construit une installation critique perfectionnee a plusieurs regions, qu'on a appelee 'reacteur d'etude des constantes physiques' (RECP), dont on s'est servi pour determiner les constantes physiques de plusieurs reacteurs de puissance. On s'en est servi aussi couramment pour mesurer des sections efficaces et determiner des parametres differentiels et integraux de la physique des reacteurs pour divers types de milieux multiplicateurs. Apres la construction de RECP, on a encore employe les experiences exponentielles, bien que RECP ait largement comble les espoirs qui avaient ete places en lui. L'auteur indique quelques donnees caracteristiques obtenues a l'aide de ces deux genres d'installations et compare leurs roles respectifs pour l'etude de nouveaux reacteurs de puissance, pour la modification de reacteurs en fonctionnement, comme moyens de recherche sur la physique des reacteurs et comme moyen de formation. Il compare egalement les montants des capitaux investis dans ces installations et des frais de fonctionnement. Il indique comment ont ete mises au point de nouvelles methodes experimentales

Burnup determination of power reactor fuel elements by gamma spectrometry; Determination par spectrometrie {gamma} du taux d'irradiation des elements combustibles des reacteurs de puissance

Energy Technology Data Exchange (ETDEWEB)

Robin, M; Jastrzeb, M; Boisliveau, S; Boyer, R; Vidal, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1969-07-01

This report describes a method for determining by {gamma} spectrometry the burn up and the specific power of fuel elements irradiated in power reactors. The energy spectrum of {gamma} rays emitted by fission products is measured by means of a simple equipment using a sodium iodide detector and a multichannel analyzer. In order to extract from the spectrum a quantity proportional to the burn up, it is necessary to: - isolate an activity specific of one emitter,- give the same importance to fissions in uranium and plutonium - take into account the radioactive decay during and after irradiation. One hundred fuel elements were studied and burn up values obtained by {gamma} spectrometry are compared to results given by chemical analyses. Preliminary measurements show that the accuracy of the results is greatly increased by the use of a germanium detector, due to its good resolution. (authors) [French] Ce rapport expose une methode de determination par spectrometrie {gamma} du taux d'irradiation et de la puissance specifique des elements combustibles irradies dans les reacteurs de puissance. Une installation simple utilisant un detecteur d'iodure de sodium et un selecteur multicanaux mesure le spectre en energie du rayonnement {gamma} emis par les produits de fission. Afin d'extraire du spectre une quantite proportionnelle au taux de combustion, il faut: - isoler une activite specifique a un emetteur, - donner la meme importance aux fissions survenues dans l'uranium et le plutonium, - prendre en compte la decroissance radioactive pendant et apres l'irradiation. Les mesures ont porte sur une centaine d'elements combustibles et les taux de combustion obtenus par spectrometrie {gamma} sont compares aux resultats des analyses chimiques. Des mesures preliminaires montrent que l'utilisation d'un detecteur de germanium augmente considerablement la precision des resultats, en raison de son excellente resolution. (auteurs)

Molten salts in nuclear reactors; Les sels fondus dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Dirian, J; Saint-James, [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

Collection of references dealing with the physicochemical studies of fused salts, in particular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thorium are examined, and the physical properties, density, viscosity, vapour pressure etc... going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recuperation after irradiation in a nuclear reactor is discussed. (author) [French] Bibliographie regroupant l'etude physico-chimique des sels fondus, en particulier des halogenures alcalins et alcalino-terreux. On etudie de nombreux systemes binaires, ternaires et quaternaires de ces halogenures avec des halogenures d'uranium, et de thorium. On etudie egalement les proprietes physiques des halogenures ou des melanges d'halogenures (densite, viscosite, tension de vapeur, etc...). On donne egalement des references quant a la corrosion des materiaux par ces sels, et le traitement de ceux-ci en vue de recuperation, apres irradiation dans un reacteur nucleaire. (auteur)

Physical measurements in Marcoule reactors (1962); Mesures physiques sur les reacteurs de Marcoule (1962)

Energy Technology Data Exchange (ETDEWEB)

Teste du Bailler, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

A brief description of the physical measurements in Marcoule reactors is given here. During commissioning and subsequent years of operation, various experiments ha been carried out to check design data, and improve the operating conditions and also test theoretical models for kinetic studies. (author) [French] On presente une rapide description des mesures physiques effectuees sur les reacteurs de Marcoule. Au cours du demarrage et pendant les premieres annees de fonctionnement de G-2 - G-3, de nombreuses experiences ont ete effectuees pour verifier les donnees du projet, ameliorer les conditions de fonctionnement et eprouver des modeles theoriques de calculs de cinetique. (auteur)

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.

Contribution to the study of can deformations in the fuel elements of gas-graphite reactors during thermal cycling; Contribution a l'etude des deformations des gaines des elements combustibles de reacteur graphite-gaz au cours du cyclage thermique

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M; Boudouresques, B; Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cans of fuel cartridges used in reactors of the gas-graphite type have either longitudinal fins of variable thickness, short herring-bone fins, or else a mixture of the two. An important test of the strength of these cartridges is their behaviour during thermal cycling carried out in cells reproducing in-pile conditions. It has been observed during with rapid cooling that there occurs a shortening at the base of the fins which can be accompanied in particular by a compression effect at the fin type, which has a tendency to curl, and by a tractive force acting on the body of the can at the ends of the longitudinal fins; this last phenomenon can result in a fracturing of the welds at the extremities or of the ends of the cartridge. This report presents first of all the way in which the stress diagram can be drawn for a can touching the fuel, and then the effect of the ratchet along a fin fixed to a bar with or without grooves. Finally the importance is shown of the test cycling variables (temperature, heating and cooling rates). (authors) [French] Les gaines des cartouches combustibles des reacteurs de la filiere graphite-gaz comportent soit des ailettes longitudinales plus ou moins epaisses, soit de courtes ailettes a chevrons, soit un ensemble des deux. Un test important de la tenue des cartouches, est la tenue au cyclage thermique en cellule pour reproduire le comportement en pile. On a observe au cours des cyclages a refroidissement rapide, un raccourcissement a la base des ailettes qui peut s'accompagner notamment d'une mise en compression du sommet de l'ailette qui a tendance a friser, et d'une traction exercee sur le corps des gaines au bout des ailettes longitudinales; ce dernier phenomene peut se traduire par des ruptures de soudures d'extremites ou des parties terminales de la cartouche. Ce rapport presente d'abord la maniere dont peut etre trace le diagramme des contraintes dans une gaine liee au combustible, puis l'effet du rochet le long d

Experience gained in two years operation of G1; Experience acquise au cours de deux ans de fonctionnement du reacteur G1

Energy Technology Data Exchange (ETDEWEB)

de, Rouville; Pascal, [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Scalliet, [Electricite de France (EDF), 75 - Paris (France)

1958-07-01

Technical specifications in respect of the first plutonium generating graphite reactor, the G1 at Marcoule, were stated in a paper read at the first Geneva Conference in 1955. We shall not therefore deal further with the technical characteristics of G1 in the present note, but rather propose to define - in the characteristic fields we think will be of major interest to foreign specialists - the results obtained in two and a half years operation since G1 first became critical on january 7, 1956. (author)Fren. [French] Les caracteristiques techniques du premier reacteur plutonigene, au graphite, de Marcoule, G1, ont ete donnees dans une communication presentee a la premiere conference de Geneve, en 1955. Nous n'y reviendrons donc pas dans la presente note qui a pour objet de faire le point, dans quelques domaines caracteristiques, qui nous ont paru les plus susceptibles d'interesser les specialistes etrangers, des resultats obtenus et des experiences faites au cours des deux annees et demi de fonctionnement du reacteur qui ont suivi sa divergence, le 7 janvier 1956. (auteur)

The use and evolution of the CEA research reactors; Utilisation et evolution des reacteurs de recherche du C.E.A

Energy Technology Data Exchange (ETDEWEB)

Rossillon, F; Chauvez, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

reacteurs en montrant ce qu'a ete jusqu'a present leur utilisation, et comment certaines modifications ont permis de les adapter a l'evolution des programmes. Ils precisent egalement les raisons qui ont conduit a l'elaboration du projet de la nouvelle pile OSIRIS, La pile ZOE, la plus ancienne du CEA, est en service au Centre de Fontenay-aux-Roses depuis 1948. Elle est principalement utilisee pour les mesures de section efficace d'absorption du graphite, et pour diverses irradiations de courte duree ne necessitant que des flux peu eleves. La Pile EL2, en service depuis 1952, a permis les premieres etudes liees au refroidissement par gaz. Elle a ete tres utilisee pour la production des radioisotopes et pour de nombreuses experiences de physique, de metallurgie et de physico-chimie - le vieillissement de certaines parties du reacteur a conduit a decider l'arret prochain de cette installation. La Pile EL. 3 a ete tres utilisee pour les experiences de physique et pour l'etude des combustibles. L'adoption d'une nouvelle structure pour le coeur (solution 'Cristal de neige') va permettre d'accroitre considerablement les possibilites de la pile pour les irradiations en neutrons rapides. La pile TRITON-I, piscine de 2 MW, est surtout utilisee pour les irradiations en neutrons rapides et en gamma. Certaines modifications, actuellement en cours, permettront d'accroitre la puissance du reacteur jusqu'a 4 ou 5 MW. Dans un compartiment voisin de TRITON-I est implantee la Pile TRITON-II, de meme structure generale, mais dont la puissance maximum est de 100 kW. TRITON-II est utilisee exclusivement pour les etudes de protections. MELUSINE, pile piscine de 2 MW est en fonctionnement au Centre d'Etudes Nucleaires de Grenoble depuis 1959. Elle a permis l'execution d'un programme important concernant surtout la physique du solide, l'etude fondamentale de combustibles refractaires et de graphites speciaux, et l'etude du comportement des liquides organiques sous radiations. Les installations de

Prospects for the Use of Plutonium in Reactors; Prospective d'Utilisation du Plutonium dans les Reacteurs

Energy Technology Data Exchange (ETDEWEB)

Fossoul, E.; Haubert, P. [BELGONUCLEAIRE (Belgium); Hirschberg, D.; Morlet, E. [International Business Machines of Belgium, Bruxelles (Belgium)

1967-09-15

son acquisition a l'exterieur, ce qui permet de faire une bonne approximation et d'eliminer la grande inconnue du prix de marche du plutonium dans les decennies a venir. Etant donne pour ce systeme une politique d'implantation de centrales nucleaires, c'est-a-dire un ensemble de decisions d'installer des centrales de type et de puissance donnes a des dates donnees, les techniques de programmation lineaire permettent d'optimiser l'utilisation du plutonium produit de facon a minimiser le cout total actualise de la production cumulee d'energie electrique pendant une periode determinee. Une etape ulterieure est l'optimisation, par des techniques differentes, non seulement de l'utilisation mais aussi de la production de plutonium, et cela en choisissant les types de reacteurs a installer dans les differentes centrales. (author)

Contribution to the study on the flow rate adjustment for gas cooled power reactors (1964); Contributiom a l'etude de reglage du debit pour les reacteurs industriels refroidis par gaz (1964)

Energy Technology Data Exchange (ETDEWEB)

Milliot, B [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1961-06-15

1. This original study firstly defines the problem of the adjustment of the coolant flow rate in a reactor channel as a function of the corresponding heat transfer equations and of the local and temporal neutron flux. The necessity of such an adjustment is pointed out and the modifying parameters are studied. An adjustment study using the envelope of the possible flux curves is developed. A short study on the technology and the economical advantage of this adjustment is presented. Some measurements, made on G-1 and G-2, show the precision one can obtain from adjustment apparatus itself as well as from the complete reactor adjustment system. 2. Evolution of nuclear properties of fuel in an heterogeneous thermal reactor. In the first port of this paper, the phenomena of fuel evolution have been mainly pointed out. Now a bibliographical study more qualitatively than quantitatively has been done. This survey specifies the present theories and relates to a real effective cross section and also yields to the bases of such a nuclear calculation. (author) [French] 1. Cette etude originale definit d'abord le probleme du reglage du debit de refrigerant dans un canal de reacteur en fonction de la formulation du calcul des performances thermodynamiques de ce canal et des variations du flux neutronique dans l'espace et le temps. La necessite du reglage est ensuite mise en evidence et les parametres le modifiant sont etudies. Une methode de reglage, basee sur l'emploi d 'une courbe enveloppe des courbes de flux possibles, est donnee. Une breve etude de la technologie et des incidences economiques du reglage est presentee. Des mesures effectuees sur les reacteurs G-1 et G-2 montrent la precision que l'on peut attendre des dispositifs de reglage comme du reglage d'ensemble du reacteur lui-meme. 2. Evolution des proprietes nucleaires du combustible dans un reacteur heterogene a neutrons thermiques. Les phenomenes d'evolution du combustible tiennent une place importante dans l

The Role of Non-Destructive Testing in Test-Reactor Operation at the National Reactor Testing Station; Role des Essais Non Destructifs dans l'Exploitation des Reacteurs d'Essai au Centre National d'Essais de Reacteurs; Rol' nedestruktivnykh ispytanij pri ehkspluatatsii ispytatel'nykh reaktorov na natsional'noj stantsii po ispytaniyam reaktorov; Papel de los Metodos No Destructivos en la Explotacion de los Reactores de la National Reactor Testing Station

Energy Technology Data Exchange (ETDEWEB)

Francis, W. C.; Brown, E. S.; Burdick, E. E.; Gibson, G. W.; Tingey, F. H. [Phillips Petroleum Company, Atomic Energy Division, Idaho Falls, Idaho (United States)

1965-10-15

surface cracks, thermal anneal tests for blistering, and gamma-scanning of irradiated plates. Hydraulic testing of statistical sampling of fuel elements is used to confirm structural integrity, particularly the fuel plate-side plate-joint strength. A continuous effort is made to improve existing techniques and to develop new non-destructive inspection procedures. (author) [French] Les investissements tres importants (plus de 100 millions de dollars) consacres aux reacteurs d'essai du Centre national d'essais de reacteurs et la necessite d'exploiter ces reacteurs en toute securite exigent un controle extremement strict de la qualite des reacteurs et de leurs parties constitutives, notamment des elements combustibles et du dispositif de commande. Les essais non destructifs ont donc joue un role essentiel dans le controle de la qualite de ces pieces avant leur utilisation dans les. reacteurs d'essai. Bien qu'un grand nombre de ces essais non destructifs soient executes selon des procedures bien etablies, on a mis au point de nombreuses methodes inedites et introduit de nouvelles utilisations du materiel classique. On applique depuis longtemps au Centre d'essais les methodes ultrasonores pour la detection des cavites, des defauts de liaison et des craquelures internes. Recemment, on a etendu ces methodes a l'exploration automatique des plaques courbes et a l'inspection des elements combustibles irradies dans les canaux de stockage. Des travaux tres interessants ont permis d'appliquer la methode des ultrasons a la detection des fractures qui peuvent se produire dans l'ame lors du faconnement. Une methode d'exploration par rayons gamma, pour determiner la teneur d'elements combustibles en {sup 23}5{sup U}, s'est revelee tellement fiable qu'elle a ete adoptee pour calculer les penalisations financieres pour les articles non conformes aux specifications. Les radiographies de plaques de combustible donnent les dimensions de l'ame et, associees aux explorations'a l'aide d

Production of artificial radioelements; Production des radioelements artificiels

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

The techniques used in the production of artificial radioelements are described, with special emphasis on the following points: - nuclear reactions and use of reactors; - chemical separation methods and methods for enriching the activity of preparations; - protection of personnel and handling methods. (author) [French] On decrit l'ensemble des techniques utilisees dans la fabrication des radioelements artificiels en insistant notamment sur les points suivants: - reactions nucleaires et utilisation des reacteurs; - methodes de separations chimiques et methodes d'enrichissement d'activite des preparations; - protection du personnel et methodes de manipulation. (auteur)

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...

Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

Energy Technology Data Exchange (ETDEWEB)

Vendryes, G.; Zaleski, C.P. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les

Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

Energy Technology Data Exchange (ETDEWEB)

Vendryes, G; Zaleski, C P [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les projets de reacteurs futurs

Problems related with the power regulation of reactors by physico-chemical methods, and the behaviour of water and heavy water in nuclear reactors; Comportement de l'eau et de l'eau lourde dans les reacteurs nucleaires et problemes de la regulation de puissance par voie physico-chimique

Energy Technology Data Exchange (ETDEWEB)

Dolle, L; Conan, D; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

. Comportement de l'eau lourde dans les reacteurs en exploitation. Pollution isotopique de l'eau lourde: Sa vitesse est liee au type de reacteur et a certains incidents caracteristiques. L'utilisation d'une colonne de reconcentration est un moyen efficace pour maintenir le titre de l'eau lourde dans un reacteur dont la pollution isotopique lente ne peut etre exclue. Detection des fuites d'eau lourde: Elle permet de mesurer les taux instantanes de fuites faibles, de localiser la fuite, et de controler la contamination atmospherique dans l'enceinte du reacteur. On procede par analyse isotopique du deuterium ou par dosage du tritium sur des echantillons d'eau de condensation. Pollution chimique et epuration de l'eau lourde: La pollution chimique de l'eau lourde constitue un des problemes les plus complexes de la chimie des reacteurs. La corrosion des materiaux constituant le coeur et le circuit d'eau lourde varie dans de larges limites avec l'etat de purete de l'eau lourde, les performances des circuits d'epuration et des mesures directes permettent d'en apprecier l'importance. Les connaissances acquises permettent de degager des normes de purete dont l'observation est susceptible de garantir un fonctionnement satisfaisant du reacteur. 4) Decomposition radiolytique de l'eau lourde: Une meilleure connaissance de son allure quantitative dans les reacteurs est necessaire pour prevoir les degagements de gaz tonnant dans les reacteurs de puissance. Le taux de radiolyse evolue avec la purete chimique de l'eau et la puissance instantanee du reacteur. L'experience des reacteurs a eau lourde du CE.A. et l'etude systematique de la decomposition radiolytique de l'eau dans le coeur des piles piscines sont exposees Mise en oeuvre du controle de la reactivite par voie physico-chimique. Controle de la reactivite par empoisonnement homogene du moderateur: Une comparaison de l'evolution de l'empoisonnement Xenon avec l'antireactivite residuelle du poison en solution pendant sa consommation

A fly-wheel drive with controlled-torque clutch for a reactors cooling circuit pumps; Entrainement des pompes du circuit de refrigeration d'un reacteur par volant a embrayage sous couple controle

Energy Technology Data Exchange (ETDEWEB)

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

1964-10-15

After a theoretical study on the slowing down of a centrifugal pump, the motion equations have been checked by means of experimental tests. In order to have important slowing down times (which is the case of the cooling pumps of a research reactor) it is necessary to add a fly-wheel. To prevent troubles when starting, a block pump-fly-wheel with clutch under controlled torque was developed. It is so possible to start the fly-wheel progressively without increasing too much power of the driving motor. (author) [French] Apres une etude theorique sur le mouvement de ralentissement d'une pompe centrifuge, les equations du mouvement ont ete verifiees par des essais pratiques. Pour obtenir des temps de ralentissement importants (cas des pompes de refrigeration d'un reacteur de recherche) il est necessaire d'y adjoindre un volant d'inertie. Pour eviter les inconvenients au demarrage, on a etudie un ensemble pompe-volant avec embrayage sous couple controle. Cette solution permet de lancer progressivement le volant sans augmentation appreciable de la puissance du moteur d'entrainement. (auteur)

Contribution to the study on the flow rate adjustment for gas cooled power reactors (1964); Contributiom a l'etude de reglage du debit pour les reacteurs industriels refroidis par gaz (1964)

Energy Technology Data Exchange (ETDEWEB)

Milliot, B. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1961-06-15

1. This original study firstly defines the problem of the adjustment of the coolant flow rate in a reactor channel as a function of the corresponding heat transfer equations and of the local and temporal neutron flux. The necessity of such an adjustment is pointed out and the modifying parameters are studied. An adjustment study using the envelope of the possible flux curves is developed. A short study on the technology and the economical advantage of this adjustment is presented. Some measurements, made on G-1 and G-2, show the precision one can obtain from adjustment apparatus itself as well as from the complete reactor adjustment system. 2. Evolution of nuclear properties of fuel in an heterogeneous thermal reactor. In the first port of this paper, the phenomena of fuel evolution have been mainly pointed out. Now a bibliographical study more qualitatively than quantitatively has been done. This survey specifies the present theories and relates to a real effective cross section and also yields to the bases of such a nuclear calculation. (author) [French] 1. Cette etude originale definit d'abord le probleme du reglage du debit de refrigerant dans un canal de reacteur en fonction de la formulation du calcul des performances thermodynamiques de ce canal et des variations du flux neutronique dans l'espace et le temps. La necessite du reglage est ensuite mise en evidence et les parametres le modifiant sont etudies. Une methode de reglage, basee sur l'emploi d 'une courbe enveloppe des courbes de flux possibles, est donnee. Une breve etude de la technologie et des incidences economiques du reglage est presentee. Des mesures effectuees sur les reacteurs G-1 et G-2 montrent la precision que l'on peut attendre des dispositifs de reglage comme du reglage d'ensemble du reacteur lui-meme. 2. Evolution des proprietes nucleaires du combustible dans un reacteur heterogene a neutrons thermiques. Les phenomenes d'evolution du combustible

Natural uranium-graphite system. Critial experiments on the G1 reactor; Systeme uranium naturel-graphite. Experiences critiques sur le reacteur G1

Energy Technology Data Exchange (ETDEWEB)

Schmitt, A P; Tanguy, P; Teste du Bailler, A; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

A number of experiments have been performed during the start up period of the G1 (1956) and G2 (1958) reactors in Marcoule, both on their lattices and on different lattices (hollow rods, clusters, under moderated lattices). The first chapter gives a thorough description of the two reactors. The second chapter deals with buckling measurements, both absolute (flux plots) and relative by the method of progressive substitution. The experimental results are summarised in Table VI. The third chapter contains a number of other measurements performed on G1. (author)Fren. [French] Le demarrage des reacteurs G1 (1956) et G2 (1958) de Marcoule nous a permis d'effectuer une serie d'experiences tant sur les reseaux de ces piles que sur des reseaux differents (elements tubulaires ou divises, reseaux sous-moderes, etc...). Dans une premiere partie, nous donnons une description detaillee des deux reacteurs. Dans la deuxieme partie, relative aux mesures de laplaciens, nous decrivons d'abord les mesures absolues de laplaciens (cartes de flux), puis les mesures relatives effectuees par la methode originale de remplacement progressif. Les resultats experimentaux sont rassembles dans le tableau VI. Dans la troisieme partie, nous rappelons un certain nombre d'autres mesures effectuees sur G1. (auteur)

Assessment of End-Plug Welding of Fuel Elements; Evaluation des Soudures Terminales des Elements Combustibles; Otsenka kachestva privarki kontsevoj probki toplivnykh ehlementov; Inspeccion de la Soldadura del Tapon Terminal de los Elementos Combustibles

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Y.; Aoki, T. [Tokai Refinery, Atomic Fuel Corporation (Japan)

1965-10-15

It is very important to correlate the testing results with the performance in reactor service, as well as to develop non-destructive testing techniques themselves. However, it is rather difficult to obtain these correlations because of high expense and radioactivity. Several kinds of assessments in out-of-pile state were carried out simulating the in-reactor conditions. Some details of these assessments on JRR-3 fuel elements are described. The reactor is a heavy-water moderated and cooled research reactor of 10-MW capacity, with aluminium-clad metallic uranium fuel elements. As the elements have only mechanical bonding between cladding and core, there might be a tensile stress at the end plug as a result of irradiation growth of the uranium core. Thermal cycling will cause a similar stress in the welds. Preferential corrosion by hot water might occur in the vicinity of the welds because of the difference of micro-structure. It is essential to keep leak-tightness during and after the reactor service. Specially designed specimens were used for tensile testing, high-temperature creep testing, thermal cycling and corrosion testing. Many sorts of weld characters were examined non-destructively before the tests and leak-checked at intervals of the tests. Evaluations of these results may be used for the establishment of inspection standards such as X-ray radiography and visual inspection of the end-plug welding. Some other results on Magnox-clad and Zircaloy clad fuel elements will also be described. (author) [French] Il est tres important de mettre en correlation les resultats d'essais et les performances d'un reacteur en service, et d'ameliorer les methodes d'essais non destructifs. Toutefois, cette correlation est souvent difficile S obtenir du fait des depenses elevees necessaires et de difficultes tenant S la radioactivite. Plusieurs sortes d'evaluations ont ete faites hors pile en simulant les conditions en pile. Le memoire donne certains details des evaluations

Preliminary handling studies in large size fast piles; Etudes preliminaires de manutention dans les reacteurs a neutrons rapides de grande taille

Energy Technology Data Exchange (ETDEWEB)

Leduc, J.; Marmonier, P. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

This report examines the various fuel handling systems which presently seem feasible for a fast power reactor. It tries to point out the advantages and / or the the disadvantages and the fabrication problems for each solution involved and makes, a tentative to evaluate the time required for a fuel loading and / or unloading operation. One has investigated the influence of the maximum allowable irradiation, the number of of shut-downs, the power distribution shape within the core on the storage capacity needed, the load factor expected and the average irradiation obtained. (authors) [French] On a examine dans ce rapport les differents systemes de manutention, qui semblent actuellement realisables pour un reacteur a neutrons rapides de puissance, en essayant de faire ressortir les avantages, les inconvenients et les difficultes de realisation de chaque systeme, et de chiffer les temps de manutention auxquels ils conduisent. On a aussi regarde l'influence des variations du taux d'irradiation maximal,de la cadence des arrets ou de la forme du flux dans le coeur du reacteur, sur la capacite du stockage, le taux de disponibilite et le taux d'irradiation moyen. (auteurs)

Preliminary handling studies in large size fast piles; Etudes preliminaires de manutention dans les reacteurs a neutrons rapides de grande taille

Energy Technology Data Exchange (ETDEWEB)

Leduc, J; Marmonier, P [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

This report examines the various fuel handling systems which presently seem feasible for a fast power reactor. It tries to point out the advantages and / or the the disadvantages and the fabrication problems for each solution involved and makes, a tentative to evaluate the time required for a fuel loading and / or unloading operation. One has investigated the influence of the maximum allowable irradiation, the number of of shut-downs, the power distribution shape within the core on the storage capacity needed, the load factor expected and the average irradiation obtained. (authors) [French] On a examine dans ce rapport les differents systemes de manutention, qui semblent actuellement realisables pour un reacteur a neutrons rapides de puissance, en essayant de faire ressortir les avantages, les inconvenients et les difficultes de realisation de chaque systeme, et de chiffer les temps de manutention auxquels ils conduisent. On a aussi regarde l'influence des variations du taux d'irradiation maximal,de la cadence des arrets ou de la forme du flux dans le coeur du reacteur, sur la capacite du stockage, le taux de disponibilite et le taux d'irradiation moyen. (auteurs)

Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

1964-02-15

reliance placed in the past on exponential and critical systems for fulfilling Argonne's responsibilities in reactor development. An indication of their future role is provided by a brief summary of the current and planned programmes for the existing members of, and anticipated additions to, Argonne's family of operating zero-power reactors. (author) [French] Avec le reacteur de puissance zero du Laboratoire national d'Argonne, on a procede a des etudes de reacteurs tres divers; reacteurs de recherche, generatrices nucleaires, reacteurs pour la propulsion, pour la production de radioisotopes et reacteurs experimentaux; les ensembles associes - exponentiels et critiques non empoisonnes - ont fourni les donnees debase. Afin de rendre compte d'experiences recentes et de montrer quelle masse de renseignements sur la physique des reacteurs on peut obtenir avec des systemes a bas flux, les auteurs exposent les programmes experimentaux ci-apres: 1. Etude des proprietes des elements combustibles en oxydes d'uranium et de thorium, immerges dans l'eau lourde, en s'attachant particulierement aux donnees necessaires pour l'etude d'un deuxieme coeur pour le reacteur experimental a eau bouillante du Laboratoire d'Argonne; 2. Maquette d'un reacteur de recherche a haut flux, qui permettra de verifier les calculs faits au cours de l'etude, de determiner la geometrie optimale et d'estimer l'effet du taux de combustion; 3. Determination des repartitions energetiques et de l'effet de l'immersion des cartouches sur la reactivite pour un reacteur experimental a ebullition et a surchauffe combinees; 4. Etude d'un coeur de reacteur surgenerateur plutonigene a neutrons rapides, alimente en U{sup 235} et refroidi au sodium qui constituerait la charge initiale du Deuxieme reacteur surgenerateur experimental d'Argonne; 5. Etude des caracteristiques d'un reacteur a deux regions, l'une thermique et l'autre rapide, en interaction. Dans l'expose de ces programmes, les auteurs expliquent pourquoi on a

Reactor Radiation Loops as Large Gamma Sources; Boucles d'irradiation des reacteurs nucleaires utilisees comme sources gamma intenses; Radiatsionnye kontury yadernykh reaktorov kak moshchnye gamma-istochniki; Empleo de circuitos de irradiacion de los reactores como fuentes gamma de gran intensidad

Energy Technology Data Exchange (ETDEWEB)

Ryabukhina, Yu. S.

1963-11-15

. On a etudie le comportement de deux alliages eutectiques de l'indium en presence de certains materiaux de construction; la premiere installation a ndium-gallium est entree en service au debut de 1960. Des travaux ulterieurs ont permis d'equiper le reacteur IRT de l'Academie des sciences de Georgie d'une boucle modele permettant d'obtenir dans le.canal d'irradiation une activite maximum equivalent a environ 100 g de radium, et d'installer une boucle d'essai a indium-gallium-etain dans le canal du reacteur IRT appartenant a l'Institut de l'energie atomique de l'Academie des sciences de l'URSS. Enfin, en 1962, une boucle a indium - gallium - etain a ete mise en service dans le reacteur IRT de l'Academie des sciences de Lituanie, en vue d'executer des irradiations a une echelle semi-industrielle. Son activite maximum atteignait, dans le dispositif d'irradiation, un niveau equivalent a 30 000 g de radium. Le memoire se compose des quatre parties suivantes: 1. ''Calcul des boucles d'irradiation''; les auteurs generalisent les resultats des travaux sur les methodes de calcul des boucles d'irradiation. 2. ''Modele d'une boucle d'irradiation a indium-gallium pour le reacteur IRT-2000 de Tbilisi''; les auteurs decrivent le fonctionnement de la boucle. 3. ''Boucle d'irradiation a indium-gallium-etain du reacteur nucleaire IRT de l'Academie des sciences de Lituanie''; les auteurs decrivent le fonctionnement de la boucle. 4. des boucles d'irradiation''; les auteurs decrivent des experiences, donnent des schemas et indiquent les calculs sur la base desquels il devient possible-de construire des boucles a manganese solide et des boucles utilisant des alliages liquides d'indium. (author) [Spanish] Desde 1957 se vienen realizando en la Union Sovietica estudios sobre la construccion de circuitos de irradiacion. Se idearon metodos de calculo de tales sistemas y se averiguaron las posibilidades que ofrecen los distintos portadores gamma. En

A new detector for the measurement of neutron flux in nuclear reactors; Nouvelle methode de mesure des flux de neutrons dans les reacteurs atomiques

Energy Technology Data Exchange (ETDEWEB)

Koch, L; Labeyrie, J; Tarassenko, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high {gamma} ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their {beta} and {gamma} disintegration. Thermal neutron fluxes as low as few neutrons per cm{sup 2} per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm{sup 2}) for fluxes higher than 10{sup 8} n/cm{sup 2}/s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author)Fren. [French] On decrit un appareil permettant la mesure instantanee des flux de neutrons thermiques accompagnes de flux intenses de rayons {gamma} et situes dans des enceintes pouvant etre portees a des temperatures superieures a 500 deg. C. On utilise la radioactivite des atomes resultant de la fission des noyaux lourds; ces atomes sont entraines par un courant gazeux vers un detecteur de radioactivite qui enregistre leurs desintegrations {beta} et {gamma}. On peut mesurer des flux partir de quelques neutrons thermiques par cm{sup 2} et par seconde. L'appareil permet de suivre la puissance d'un reacteur atomique, de tracer des cartes de densite de neutrons avec une tres bonne definition (1 mm{sup 2}) dans le cas de flux superieurs a 10{sup 8} cm{sup 2}/s. Le temps de reponse du systeme a une variation du flux de neutrons est limite, poes flux importants, par le temps de transit du gaz entre l'emetteur de produits de fission et le detecteur: soit quelques dizaines de

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)

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

Notes on a homogeneous reactor project; Idees sur un projet de reacteur homogene

Energy Technology Data Exchange (ETDEWEB)

Benveniste, J; Bernot, J; Eidelman, D; Grenon, M; Portes, L; Raspaud, G; Tachon, J; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod, L; Cohen de Lara, G; Delachanal, M; Fontanet, P; Halbronn, G [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France)

1958-07-01

An attempt has been made to develop certain ideas concerning homogeneous reactors. The project under consideration is based on the simultaneous use of a suspension of uranium dispersed in heavy or light water and of boiling in the reactor for heat extraction. However, the studies of suspensions and of boiling are relatively independent and can also be developed for reactors of different types using one or the other. Our aim is a minimum investment in fissile material; for this we propose to extract the steam directly from the core and to make use of a cyclone to accelerate this extraction; a cyclone-type circulation creating a field of increasing tangential velocities of the fluid towards the axis causes the droplets of vapour to accelerate towards the axial vortex in which they are collected; the steam output is then evacuated to the external heat utilisation system, for example an exchanger of the condenser-boiler type. The input speed of water into the reactor being one of the important parameters in the running of the pile, a spiral supply input chamber is used, allowing this speed to be regulated in amount and direction. (author)Fren. [French] Nous nous sommes attaches a developper certaines idees relatives aux piles homogenes. Le projet que nous etudions est base sur l'emploi simultane d'une suspension contenant de l'uranium disperse dans l'eau legere ou lourde et de l'ebullition dans le reacteur pour l'extraction de chaleur. Neanmoins, les etudes de suspensions et d'ebullition sont relativement independantes et peuvent egalement etre developpees pour des reacteurs de type different utilisant l'une ou l'autre. Le but que nous cherchons a atteindre est un investissement minimum en matiere fissile; pour cela, nous proposons d'extraire directement la vapeur dans le coeur et de recourir a un dispositif cyclone pour accelerer cette extraction; une circulation type cyclone creant un champ de vitesses tangentielles du fluide croissantes veraxe a pour effet d

Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

Energy Technology Data Exchange (ETDEWEB)

Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

1964-04-15

effects in APEX, HERO and AGR and for determining fine structure data and power distribution in the complex fuel assemblies are of particular interest. Current and future theoretical work is concentrated primarily on development of an alternative method to hetrecontrol and FTD2 for dealing with reactor cores after considerable burn-up of the fuel. The experimental programme on HERO is designed to test these methods with complex cores including plutonium bearing fuel. Additional information on the effect of plutonium will be derived from operation of AGR and physics measurements on fuel after irradiation. (author) [French] Le memoire relate les recherches experimentales et theoriques auxquelles on a procede lois de l'etude, de la realisation et de la mise en service du reacteur perfectionne refroidi par un gaz (AGR) de Windscale et, d'une facon generale, pour la mise au point d'un filiere de ce type en vue de la production d'energie electrique industrielle. Il decrit l'important volume de travail qui a ete necessaire en vue d'elaborer les methodes theoriques voulues pour calculer: a) la repartition du flux et l'equilibre de la reactivite dans un coeur complexe; b) la repartition de la puissance dans des geometries de combustible complexes-, c) les effets de l'irradiation sur le cycle du combustible et la repartition de la puissance. A titre d'introduction, le memoire resume la documentation experimentale et les methodes theoriques qui sont le resultat des recherches sur la filiere a uranium gaine de magnox et decrit la documentation experimentale obtenue par le programme commun des industries britanniques (BICEP); toutes ces donnees ont servi de point de depart pour l'elaboration de methodes theoriques applicables a l'AGR. On s'est servi de l'ensemble critique APEX et du reacteur HERO de puissance zero avec des configurations de reseau regulieres et diverses combinaisons de perturbateurs (notamment des barres de commande) pour calculer les parametres de reseau de l'AGR et

Integral physics data for fast-reactor design; Donnees de physique integrale intervenant dans les etudes de reacteur a neutrons rapides; Integral'nye fizicheskie dannye dlya raschetov reaktorov na bystrykh nejtronakh; Datos fisicos integrales para el diseno de reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W B; Meneghetti, D [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

systems. (author) [French] La compilation recente du chapitre sur la physique des reacteurs a neutrons rapides dans la preparation de la deuxieme edition de 'Reactor Physics Constants' a entraine une recapitulation des resultats disponibles des mesures experimentales globales. Le choix des donnees integrales connues relatives a la physique des reacteurs a neutrons rapides a faire figurer dans cette compilation a ete fait en fonction de deux criteres : a) informations recueillies a partir de reacteurs relativement simples et qui se pretent a des analyses theoriques simples, et b) informations recueillies a partir de reacteurs complexes, representant des prototypes ou des maquettes, et qui offrent un interet general pour les reacteurs de puissance a neutrons rapides. Le premier critere a pour objet de donner une enumeration des informations concernant les systemes les plus couramment utilises pour verifier les parametres des sections efficaces et les methodes de calcul. Le deuxieme critere est fonde sur la representation des informations courantes concernant les reacteurs a surgeneration, a neutrons rapides, existant. Ces informations sont trop compliquees pour qu'il soit possible de proceder a leur egard a des analyses theoriques simples. Elles prouvent la complexite du reacteur reel, par rapport a l'experience critique plus schematique et plus facile a analyser. Les donnees integrales intervenant dans les calculs de reacteurs sont les resultats des mesures faites, sur des types de reacteurs critiques ou non, des diverses grandeurs de la physique des reacteurs qui presentent un interet pratique et/ou theorique. Elles caracterisent le type de reacteur et aident a sa comprehension. Les mesures portent sur la masse critique, le facteur forme du coeur, les pourcentages de detection, les spectres des neutrons, les experiences de substitution de materiaux, le gain reflecteur, le temps de vie des neutrons, l'{alpha} de Rossi et sur d'autres grandeurs similaires. Les auteurs

Nuclear reactor (1960); Reacteurs nucleaires (1960)

Energy Technology Data Exchange (ETDEWEB)

Maillard, M L [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Leo, M B [Electricite de France (EDF), 75 - Paris (France)

1960-07-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [French] Les premiers reacteurs industriels plutonigenes francais G1 - G2 - G3 du Centre de Marcoule comportent une installation de recuperation d'energie. La production d'electricite de G1 ne compense pas l'energie depensee par ailleurs pour le fonctionnement de l'ensemble, par contre, G2 et G3 doivent fournir chacun une puissance de 25 a 30 MW au reseau national d'Electricite de France. Cette puissance est modeste, mais l'experience acquise grace a ces reacteurs est tres grande et c'est grace a elle qu'il nous sera possible de mettre en exploitation les reacteurs energetiques EDF1 - EDF2 - EDF3. Le memoire decrit comment, avant tout demarrage du reacteur, les essais effectues, en particulier ceux concernant l'installation de recuperation d'energie et le caisson, ont permis d'abreger la phase de montee en puissance. (auteur)

[Project for] a high-flux extracted neutron beam reactor [for physicists]; Un [projet de] reacteur a haut flux et faisceaux sortis [pour physiciens

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

tubes and the experimental equipment which can support doses much higher than the ones which are biologically permissible. The final part of the communication describes the studies carried out on the realization of a liquid hydrogen cold sink, one of the most important experimental devices envisaged. (authors) [French] Les besoins francais en canaux pour sortie de neutrons de differentes energies sont brievement indiques. L'interet bien connu des neutrons froids (plus de 4 Angstroem) est souligne. Les grandes lignes d'un reacteur permettant de satisfaire les physiciens sont esquissees. Ce sont les suivantes: 1 - Flux dans l'eau lourde du reflecteur de l'ordre de 7. 10{sup 14} thermiques. 2 - Souplesse d'emploi maximum obtenue par: - separation physique du coeur et du reflecteur, - independance des experiences entre elles, - possibilite de modification, sans interruption notable du fonctionnement de la pile, des experiences physiques jusqu'a - et y compris - la nature du reflecteur utilise, - reduction au minimum des protections fixes; emploi largement generalise des protections liquides (eau) et fluidisees (sables). 3 - Continuite technologique aussi grande que possible avec les reacteurs de recherche francais existant ou en construction (SILOE, PEGASE, OSIRIS). 4 - Surete de fonctionnement recherche par la simplicite de conception. 5 - Minimisation des frais de construction. La reduction des frais d'exploitation est recherchee plutot indirectement par la simplicite des solutions et la reduction du personnel d'exploitation, que directement par la minimisation des consommations d'elements combustibles et d'energie. La solution preconisee peut etre decrite comme un reacteur de type piscine a coeur clos, non pressurise, tres sous modere par l'eau legere de refroidissement. Entourant le reacteur, se trouvent un certain nombre de 'canaux boucles' comprenant chacun: - une portion du reflecteur (eau lourde dans l'exemple decrit), - une portion de canal d'extraction de neutrons

The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

) [French] L'auteur presente les calculs du comportement d'EBR-II statique, dynamique et sous evolution a long terme de la reactivite ainsi que les resultats et l'analyse des experiences critiques seches faites sur EBR-II et en simulation sur ZPR-III. Il insiste particulieremen t sur les problemes de physique des reacteurs qui, dans l'elaboration du projet, suivent le choix du modele theorique et precedent la construction ou la mise en exploitation. L'auteur presente des analyses de la securite des reacteurs ainsi que diverses considerations sur l'evaluation des risques sous l'angle de leur influence sur le projet de reacteur. Il decrit la simulation d'EBR-II, a partir des renseignements fournis par le ZPR-III ainsi que les mesures critiques seches sur EBR-II. Ces experiences, leur analyse et les previsions des calculs servent de bases pour predire le comportement physique du reacteur. L'auteur approfondit quelque peu la validite intrinseque de l'application des donnees experimentales au fonctionnement du reacteur de puissance. Ceci comprend les donnees precises des dimensions du coeur et/ou de l'enrichissement de l'alliagne combustible, le choix convenable des valeurs de la reactivite prevues en exploitation et pendant l'arret, la determination des coefficients de reactivite a la temperature et a la puissance de fonctionnement, et la distribution precise de la puissance et du flux en fonction de la position dans l'ensemble du reacteur. L'auteur decrit le probleme de l'application des renseignements obtenus a partir d'une geometrie simple, ideale, analytique ou experimentale, a la geometrie reelle hexagonale du reacteur. Il compare le rendement nucleaire, y compris la surgeneration, du reacteur reel par rapport a celui du modele theorique. Il decrit la reactivite a long terme et le comportement energetique de la couche fertile du reacteur dans le cadre de l'etude du cyclage propose du combustible et de l'alliage fertile. L'auteur etudie les questions de securite considerant

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

A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

Energy Technology Data Exchange (ETDEWEB)

Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

1962-03-15

fonctionne a la puissance nominale depuis le mois de decembre 1960. Ce reacteur est utilise comme source puisee de neutrons pour les experiences de physique fondees sur la methode du temps de vol. On l'emploie pour etablir la section efficace totale et la section efficace de capture des neutrons intermediaires, pour etudier l'interaction des neutrons lents et des corps solides ou liquides et pour mesurer les spectres neutroniques dans differents milieux. Le memoire decrit les caracteristique s essentielles de la construction du reacteur et les resultats d'experiences faites a l'aide de ce reacteur. Le regime de fonctionnemen t normal est celui des impulsions periodiques. Les impulsions de puissance sont produites par un deplacement rapide de la partie mobile du coeur a travers sa partie immobile. La partie mobile se trouve fixee sur un disque tournant et se deplace a une vitesse d'environ 230 m/s. Une zone mobile auxiliaire permet de modifier la frequence des impulsions de puissance entre 2,3 et 88 ips. Le reacteur a une puissance moyenne de 1 kW. La demi-largeur d'une impulsion de puissance est de 36 {mu}/s. Le reacteur est dote d'un systeme de commande et de securite qui assure le maintien automatique de la puissance moyenne et un arret rapide en cas de fonctionnement irregulier. Il est equipe d'un systeme de canalisations sous vide pour le passage des neutrons, qui permettent de mesurer le temps de vol. Le canal principal a 1000 m de long. Lors du demarrage du reacteur et durant les experiences de physique dont il a fait l'objet, on a etudie l'effet que produit sur la reactivite le deplacement des organes de commande et des parties mobiles du coeur; on a mesure la longueur des impulsions a des regimes de fonctionnement differents et etudie les fluctuations d'amplitude des impulsions de puissance. En outre, les auteurs ont procede a des mesures en vue de determiner la duree de vie des neutrons instantanes, la fraction effective de neutrons retardes et les coefficients de

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.

Report by the AERES on the unit: Reactor Study Department (DER) under the supervision of the establishments and bodies: Atomic Energy and Alternative Energies Commission (CEA); Rapport de l'AERES sur l'unite: Departement d'Etudes des Reacteurs (DER) sous tutelle des etablissements et organismes: CEA

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-02-15

This report is a kind of audit report on a research laboratory, the DER (Departement d'Etudes des Reacteurs, Reactor Study Department) whose activity if focused on four main themes: neutron transport simulation in reactor cores, thermal-hydraulic simulation of reactors, design and safety of innovative reactors, nuclear instrumentation for reactors. The authors discuss an assessment of the whole unit activities in terms of strengths and opportunities, aspects to be improved, risks and recommendations, productions and publications, scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project. These same aspects are then discussed and commented for each theme

The cryogenic installations for irradiation in the reactors Melusine and Siloe; Les installations cryogeniques pour irradiations des reacteurs Melusine et Siloe

Energy Technology Data Exchange (ETDEWEB)

Bochirol, L; Le Calvez, J; Doulat, J; Verdier, J; Lacaze, A; Weil, L [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

vaporized in the atmosphere and without any pollution of the refrigerating circuit. Lastly, a few words are said about the liquid helium loop, a prototype of which has worked, and which is being rebuilt with an increased power. (authors) [French] L'etude des defauts crees par l'irradiation dans les solides est d'un interet theorique et pratique, considerable. L'irradiation a basse temperature permet d'obtenir les defauts dans leur etat le plus simple, leur etat 'primaire' sans que l'agitation thermique permette leur annihilation ou leur rearrangement. L'irradiation en pile a basse temperature pose un certain nombre de problemes techniques provenant de la puissance de refrigeration necessaire, qui est quelquefois considerable, des reactions chimiques possibles sous rayonnement et du manque d'espace dans un reacteur. Enfin, la necessite de faire toute l'irradiation et les mesures ulterieures sans rechauffer les; echantillons impose que le dispositif fonctionne en continu sans defaillance et qu'il soit equipe de facon a permettre la recuperation des echantillons froids, ou bien leur mesure et leur rechauffage controle 'in situ'. On decrit la facon dont ces problemes ont ete resolus a Grenoble, pour des dispositifs d'irradiation a 78 deg. K, 28 deg. K et 4 deg. K dans les deux piles piscines Melusine et Siloe. Quelques resultats d'exploitation sont donnes sur la boucle a azote liquide, dite type A, qui fonctionne depuis plusieurs annees dans Melusine. En particulier certaines observations sont faites sur les reactions chimiques qui peuvent se produire sous irradiation dans l'azote liquide impur. On decrit assez en detail la boucle a azote liquide, dite type A, qui vient d'etre installee dans le reacteur Siloe. Les traits essentiels de cet appareil sont: qu'il permet l'irradiation dans des flux plus eleves que le precedent et que son exploitation est grandement facilitee grace a un mode de realisation qui permet l'acces aux echantillons sans demontage ni deconnexion de l

Fluctuations in a system depending on several random parameters. Application to reactors (1962); Fluctuations d'un systeme dependant de plusieurs parametres aleatoires. Application aux reacteurs nucleaires (1962)

Energy Technology Data Exchange (ETDEWEB)

Blaquiere, A [Faculte des Sciences de Paris, 75 (France); Pachowska, R [Universite Technique de Varsovie (Poland)

1962-07-01

We have previously developed a method for studying neutronic fluctuations in nuclear reactors using the analogy between the behaviour of a reactor and that of certain common radioelectric circuits. The fluctuations may then be calculated by introducing into the circuit a suitable noise source. By this method we have been able to consider the overall fluctuations in a particularly simple form and we have provided a physical significance for certain results obtained more laboriously by other methods. The object of the present report is to generalise this method and in particular to extend it to the case of a reactor having a cellular structure and to apply it to fluctuations within a cell. It is thus shown that the fluctuations in a cell are the resultant of two terms: - a rapidly evolving Poissonian noise, not related to the overall fluctuations; - a slowly evolving noise, when the reactor is not too far from criticality, which is related to the overall fluctuations. The first term arises from a rapid 'ordering' of the system, during which time the cells come mutually into equilibrium. The second term is due to the coordinated evolution of all the cells, after the end of the first transitory phase. The conclusions reached show that it would be useful to complete the study with an analysis of non-linear phenomena which can considerably influence the transitory behaviour of the cells during the initial pre-equilibrium phase. This report also Stresses the relationship of the new method to the old methods. It tends also to place pile fluctuation theory in a more general framework, that of the fluctuations of a system depending on several random parameters; from this point of view, the method could easily be transposed and adapted to the study of other physical problems of this type. (authors) [French] Nous avons precedemment developpe une methode d'etude des fluctuations neutroniques des reacteurs nucleaires mettant a profit l'analogie entre le comportement d

Critical mass, rod values and reactivity coefficients for Rapsodie; Masse critique, valeur des barres et coefficients de reactivite de rapsodie

Energy Technology Data Exchange (ETDEWEB)

Stevens, L; Gourdon, J [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1967-07-01

Besides a brief general description, the report contains a description and discussion of the aims, the methods used and the results of critical mass, rod worth and static reactivity coefficient measurements on the Rapsodie reactor. (authors) [French] Apres une breve description generale, le rapport decrit et discute le but, les methodes employees et les resultats des mesures de masse critique, de reactivite des barres et des coefficients de reactivite statiques du reacteur RAPSODIE. (auteurs)

Study of new structures adapted to gas-graphite and gas-heavy water reactors; Etude de structures nouvelles adaptees aux reacteurs graphite-gaz et eau lourde-gaz

Energy Technology Data Exchange (ETDEWEB)

Martin, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [French] L'experience acquise par l'exploitation des reacteurs de MARCOULE, la construction et le demarrage des reacteurs d

Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

Energy Technology Data Exchange (ETDEWEB)

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

1964-10-01

The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible en

Problems presented by the filtration and sampling of aerosols in the atomic energy programme; Problemes poses par la filtration et le prelevement des aerosols dans le cadre de l'energie atomique

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

The maximum permissible limits for radioactive aerosols are much lower than those for aerosols encountered in the non-nuclear industries. These limits depend on numerous factors such as: nature of the radiation, half-life, etc. The radioactive aerosols can be prepared by various methods. The filtering of the air in high activity laboratories or in plutonium treatment factories necessitates an installation consisting of: - aspiration filters, - extraction filters of very high efficiency (those used for {alpha} emitter cells: designed to be replaced without incurring contamination risks; those used for {gamma} emitter cells: designed to be replaced by remote control). The filtering in nuclear reactors is also effectuated by filter papers: - the G1 reactor with open circuit: the air coolant is entirely filtered at the entry and on leaving; - the G2, G3 and EDF1 reactors with closed circuits: filtering under pressure of a small portion of the coolant gas. (author) [French] Les limites maxima permises des aerosols radioactifs sont beaucoup plus faibles que celles des aerosols rencontres dans l'industrie classique. Elles dependent de nombreux facteurs tel que: nature du rayonnement, periode radioactive, etc... La formation des aerosols radioactifs est de nature diverse. La filtration des laboratoires de haute activite, ou d'usines d'elaboration de plutonium conduit a des types d'installations comportant: - des filtres d'aspiration; - des filtres d'extraction a rendement extremement eleve (type pour cellules emettrices {alpha} concu pour etre change sans risque de contamination, type pour cellules emettrices {gamma}: concu pour etre change a distance) La filtration des reacteurs nucleaires sont egalement effectuee par des papiers filtres: - reacteur G1 a circuit ouvert: air de refroidissement totalement filtre a l'aspiration et a l'extraction; - reacteurs G2, G3, EDF1: a circuit ferme: filtration sous pression d'une faible partie du gaz de refroidissement. (auteur)

Determination of local boiling in light water reactors by correlation of the neutron noise; Determination de l'ebullition locale dans les reacteurs a eau legere par correlation du bruit neutronique

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author) [French] La limitation de la puissance des reacteurs nucleaires de type piscine est due au phenomene d'apparition de 'burn out'. Pour determiner cette limitation, nous nous sommes proposes dans ce rapport de detecter l'ebullition locale qui apparait generalement avant le 'burn out'. L'ebullition locale a ete simulee par une plaque chauffee electriquement et placee dans le coeur du reacteur SILOETTE. L'etude de l'ebullition locale, qui est basee sur les proprietes des fonctions de correlation du bruit neutronique de detecteurs places clans le coeur, fait apparaitre une frequence privilegiee dans le spectre de puissance du bruit. On envisage dans l'avenir, de determiner l'influence des divers parametres sur cette frequence caracteristique. (auteur)

Some physics aspects of cermet and ceramic fast systems; Quelques aspects de la physique des reacteurs a neutrons rapides utilisant des cermets et des ceramiques comme combustibles; Nekotorye fizicheskie aspekty kermetnykh i keramicheskikh sistem na bystrykh nejtronakh; Algunos aspectos fisicos de los sistemas rapidos a base de combustibles cermet y ceramicos

Energy Technology Data Exchange (ETDEWEB)

Codd, J; James, M F; Mann, J E [United Kingdom Atomic Energy Authority, Reactor Group (United Kingdom)

1962-03-15

The characteristics of a system using an iron-based oxide cermet as fuel material are discussed. A transport theory investigation to develop methods of predicting the effect of core heterogeneity on reactivity and flux distribution is described. Some preliminary calculations are also given of resonance self-shielding and Doppler temperature effects in a cermet system. (author) [French] Les auteurs etudient les caracteristique s d'un reacteur utilisant comme combustible un cermet d'oxydes a armature de fer. Ils exposent une application de la theorie du transport a la mise au point des methodes permettant de prevoir l'effet de l'heterogeneite du coeur sur la reactivite et sur la distribution du flux. Ils donnent egalement quelques calculs preliminaires d'effets d'autoprotection due a la resonance et d'effet Doppler du a la chaleur dans un reacteur utilisant un cermet. (author) [Spanish] La memoria discute las caracteristicas de un sistema que emplea como combustible un oxido tipo cermet a base de hierro. Describe una investigacion de la teoria de transporte con miras a desarrollar metodos para evaluar el efecto de la heterogeneidad del cuerpo sobre la reactividad y la distribucion de flujo. Tambien da algunos calculos preliminares de los efectos del autoblindaje por resonancia y de la temperatura de Doppler en un sistema de tipo cermet. (author) [Russian] Obsuzhdayutsya kharakteristiki sistemy, ispol'zuyushchej v kachestve toplivnogo materiala oksidnye kermety, razrabotannye na osnove zheleza. Opisyvaetsya issledovanie teorii perenosa, chtoby razvit' metody predskazaniya vliyaniya geterogennosti aktivnoj zony na reaktivnost' i raspredelenie potoka. Dayutsya takzhe nekotorye predvaritel'nye raschety ehffektov rezonansnoj samozashchity i temperaturnogo ehffekta Dopplera v kermetnoj sisteme. (author)

The Use of Prestressed Concrete Vessels in the French Power Reactor Programme; Les caissons en beton precontraint dans le programme francais des reacteurs de puissance; Korpusy iz predvaritel'no napryazhennogo betona vo frantsuzskoj programme ehnergeticheskikh reaktorov; Empleo de recipientes de presion de hormigon pretensado en el programa frances de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Conte, F. [Centre d' Etudes Nucleaires de Marcoule (France); Dambrine, C. [Centre d' Etudes Nucleaires de Fontenay-aux-Roses (France); Gaussot, D. [Electricite de France, Clamart (France)

1963-10-15

This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [French] La communication traite de l'application du beton precontraint aux reacteurs G2 et G3 de Marcoule et au reacteur EDF 3, en construction a Chinon. Les reacteurs sont en puissance depuis respectivement 1959 et I960; le CEA indique les problemes qui se sont poses pendant la construction du caisson du reacteur, et la lecon tiree des observations faites en service, qui tend a demontrer la tres grande securite de ces appareils. La construction du caisson de EDF3 a commence a Chinon dans la deuxieme partie de 1961; elle est en cours actuellement et sera terminee vers la fin de 1963. L'EDF presente les raisons du choix de ce caisson, les resultats des calculs et des essais sur maquette ainsi que les problemes poses par la construction. Diverses etudes ont ete faites sur les perspectives futures des ouvrages en beton precontraint pour reacteurs. Il semble que l 'on puisse realiser, si on le desire, une elevation

The Control of Fast Reactors: Current Methods and Future Prospects; Controle des Reacteurs a Neutrons Rapides. Methodes Actuelles et Perspectives d'Avenir; Upravlenie reaktorami na bystrykh nejtronakh. sushchestvuyushchie metody i dal'nejshie perspektivy; Control de Reactores Rapidos: Metodos Actuales y Perspectivas

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W. B. [Argonne National Laboratory, IL (United States)

1964-06-15

regarding the specification of this parameter. These considerations are discussed in terms of control reactivity in existing fast reactors as opposed to the amount that is really required for fast power-breeder reactor operation. Typical power- and temperature-dependent feedback parameters are cited for determination of their influence upon the control reactivity requirements. The methods used to predict the reactivity worth of control mechanisms have evolved from crude estimates to quite reliable calculations which can be confirmed by experimental data from critical assemblies. Experimental results and currently reliable analytical techniques are described. Critical experiments for the current generation of fast reactors included many investigations pertaining to the reactivity worth of their control mechanisms as well as peripheral experiments for larger-core-volume advanced systems. Exploratory analytical studies, which indicate that detailed experimental mockup investigations may not be required in the future, are cited. (author) [French] L'auteur examine dans ce memoire les aspects pratiques du probleme qui consiste a fournir une reactivite suffisante pour le controle des reacteurs a neutrons rapides; ce probleme differe dans une grande mesure de celui du controle des reacteurs a neutrons thenniques. Ces differences sont dues en premier lieu au fait que les sections efficaces d'absorption des neutrons rapides sont assez faibles. Il n'existe pas de poisons forts dans un reacteur a neutrons rapides. En consequence, les poisons forts que sont certains produits de fission dans un reacteur thermique (par exemple Xe et Sm) exigent un exces de reactivite beaucoup moins important que n'en exige la perte de reactivite due a la destruction de produit fissile par fission et capture. Comme les sections efficaces pour les neutrons rapides sont relativement petites comparees aux valeurs correspondantes pour les neutrons thermiques, la densite atomique du materiau joue un role

Operating Experience in Nuclear Power Plants with Boiling-Water Reactors; Experience acquise dans l'exploitation des reacteurs a eau bouillante; Opyt ehkspluatatsii kipyashchago reaktora; Experiencia adquirida con la explotacion de reactores de agua hirviente

Energy Technology Data Exchange (ETDEWEB)

Ascherl, R. J. [General Electric Company, San Jose, CA (United States)

1963-10-15

radioactivity exposure considerations. Recent full-scale inspection and overhaul of the Dresden turbine provided no maintenance problems, after over 12 000 h of operation on direct-cycle steam and after operation with known failed fuel elements in the reactor. (author) [French] On a maintenant acquis une experience appreciable dans l'exploitation des centrales equipees de reacteurs a eau bouillante. Vers la fin de 1962, on avait produit plus de 2,2.10{sup 9} kWh dans trois centrales nucleaires rattachees a des reseaux de distribution: la centrale de Dresden (Commonwealth Edison Company, Morris, Illinois), la centrale de Vallecitos (Pacific Gas and Electric Company and General Electric Company, Pleasanton, Californie) et la centrale de Kahl (Rheinish-Westfaiisches Elektrizitatswerk et Bayemwerk, a Kahl-sur-le-Main, Republique federale d'Allemagne). Le rendement de ces reacteurs a eau bouillante, exploites dans les conditions normales de production d'electricite, est excellent. On peut donc s'attendre que les centrales a eau bouillante continueront d'etre sures, etant donne le facteur de disponibilite et le facteur de puissance des reacteurs et des installations de ce type. Au cours de 1963, quatre nouvelles centrales equipees de reacteurs a eau bouillante entreront en service: la centrale de Big Rock Point (Consumers Power Company, Charlevoix, Michigan), la centrale de Humboldt Bay (Pacific Gas and Electric Company, Eureka, Californie), la centrale de Garigliano (Societa Elettronucleare Nazionale, Scauri, Italie) et la centrale de demonstration japonaise (Institut de recherches nucleaires du Japon, Tokai Mura, Japon). Les resultats obtenus lors du demarrage et pendant le fonctionnement initial de ces installations confirment les espoirs suscites par les centrales de Dresden, Kahl et Vallecitos. Les journaux de marche des centrales de Dresden, Kahl et Vallecitos mettent en evidence la stabilite et la securite des reacteurs a eau bouillante. De plus, les niveaux de rayonnements

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.

Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

Energy Technology Data Exchange (ETDEWEB)

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

1964-10-01

The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies; Determination de l'Efficacite des Barres de Reglage dans les Ensembles Combustibles du reacteur VVER; Opredelenie ehffektivnosti reguliruyushchikh sterzhnej v sborkakh reaktora VVEHR; Determinacion de la Eficacia de las Barras de Control en los Conjuntos de Elementos Combustibles del Reactor VVER

Energy Technology Data Exchange (ETDEWEB)

Semenov, V. N.; Lunin, G. L.; Komissarov, L. V.; Kamyshan, A. N.; Halizev, V. I.; Andrianov, G. Ja.; Voznesenskij, V. A.; Kuz' micheva, V. A.; Lebedev, V. I. [Ordena Lenina Institut Atomnoj Energii Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

1964-06-15

The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author) [French] Le memoire decrit les experiences qui ont ete faites pour determiner l 'efficacite des absorbants contenus dans les barres de compensation, l'effet cavitaire et l 'efficacite des absorbants gaines de materiaux divers, au moyen d'assemblages homogenes de cartouches de combustible du reacteur VVER (reacteur de puissance ralenti et refroidi a l 'eau ayant le meme taux d'enrichissement. On y trouve en outre des donnees sur certaines experiences executees a l 'aide d'assemblages de cartouches de combustible taux d'enrichissement differents. Ces travaux permettent de verifier la methode de calcul et d'evaluer ses possibilites d'application aux reacteurs non homogenes. (author) [Spanish] Se describen en la memoria experimentos para determinar la eficacia de los materiales absorbentes contenidos en las barras de compensacion, el efecto de cavitacion y la eficacia de los materiales absorbentes revestidos de diversos materiales, realizados con ayuda de los conjuntos homogeneos de elementos combustibles del reactor VVER (reactor de potencia moderado y refrigerado por agua) con un solo grado de enriquecimiento. Ademas, se exponen datos sobre los experimentos efectuados con ayuda de conjuntos de grados de enriquecimientos; variados. Tales experimentos permiten verificar el metodo de calculo teorico, utilizad o y evaluar la posibilidad de aplicarlo a los reactores no homogeneos. (author

Study of a Slightly Enriched R Reactor Fuel by Means of a Pulsed Neutron Source; Etude d'un reacteur a combustible legerement enrichi (rubeole) a l'aide de sources pulsees de neutrons

Energy Technology Data Exchange (ETDEWEB)

Sagot, M; Tellier, H [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1962-04-01

A Be O moderated reactor using slightly enriched uranium oxide as fuel was studied by the pulsed neutron source technique. The neutron lifetime was measured in two different cores without reflector, then attempts were made at the measurement of great negative reactivities introduced into the reactor under the following forms: decrease of the volume of the un reflected core, introduction of absorbing cadmium rods, removal of fuel at the periphery of the critical core while maintaining a constant height, and substitution of fuel elements by less reactive elements. In all cases, the results are compared with the data obtained by another type of experiment or by computation. (author) [French] Nous avons applique la methode des sources pulsees de neutrons a un reacteur utilisant de l'oxyde d'uranium legerement enrichi, modere a l'oxyde de beryllium et, apres avoir mesure le temps de vie des neutrons dans deux coeurs differents non reflechis, nous avons porte notre effort, sur la mesure de reactivites negatives importantes introduites dans le reacteur sous differentes formes: - diminution du volume du coeur non reflechi, - introduction de barres absorbantes en cadmium, - enlevement de combustible a la peripherie du coeur critique, tout en conservant une hauteur constante, - substitution d'elements de combustible par des elements moins reactifs. Dans tous les cas, les resultats sont compares aux valeurs obtenues par un autre type d'experience ou par le calcul. (auteur)

The Role of Non-Destructive Testing in the Los Alamos Reactor Programme; Role des Essais Non Destructifs dans le Programme de Reacteurs de los Alamos; Rol' nedestruktivnykh ispytanij materialov v Los-Alamosskoj reaktornoj programme; Papel de los Metodos de Ensayo No Destructivo en el Programa de Reactores de Los Alamos

Energy Technology Data Exchange (ETDEWEB)

Tenney, G. H. [University of California, Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

1965-10-15

the work on this subject has not been previously published. (author) [French] Le Laboratoire scientifique de Los Alamos, exploite par l'Universite de Californie pour la Commission de l'energie atomique des Etats-Unis, s'occupe depuis plus de vingt ans de l'etude, de la mise au point et de la construction de quatre types de reacteurs nucleaires: reacteurs de recherche, reacteurs de puissance, reacteurs pour la propulsion des fusees et assemblages critiques. Le Groupe des essais non destructifs collabore a presque tous les projets et travaux du Laboratoire. Le memoire decrit quelques-unes des methodes inedites d'essais non destructifs qui y ont ete mises au point et sont appliquees dans le cadre du programme de reacteurs. Le reacteur de puissance experimental LAMPRE est fonde sur l'utilisation d'une solution de phosphate d'uranium a haute temperature. Cette solution est tres corrosive et toutes les parties en contact avec elle ont un revetement en or. On a eu recours a des techniques radiographiques speciales pour controler l'or pendant le processus de laminage d'un lingot coule. On a procede de la meme maniere a l'inspection des soudures. Une methode d'inspection fondee sur les variations de potentiel aux electrodes a ete mise au point, pour la detection d'impuretes sur les surfaces d'or. Le reacteur experimental au plutonium fondu LAPRE est fonde sur l'utilisation de plutonium metallique, sous forme liquide plutot que sous forme solide, comme combustible. Le combustible est contenu dans des capsules en tantale. On a eu recours a des methodes non destructives pour verifier le bon etat du metal de base et des soudures pendant la fabrication des capsules, ainsi que pour controler les capsules remplies de plutonium avant, pendant et apres les essais de fusion et solidification. L'essai d'une pompe a plutonium fondu a ete suivi par des methodes radiographiques, en utilisant notamment un circuit ferme de television a rayons gamma. Pour le reacteur experimental a tres haute

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)

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

Purification by molecular sieve of helium used as inert cover gas in nuclear reactors; Epuration de l'helium de couverture des reacteurs nucleaires par adsorption sur tamis moleculaire

Energy Technology Data Exchange (ETDEWEB)

Rozenberg, J; Kahan, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

A method carried out at fairly low temperatures (between -50 and -80 deg. C) has been studied for the purification of the helium used as cover gas for heavy water in reactors. The use of the 5A molecular sieve has been adopted because of its superiority over other adsorbents in this temperature range. The particular problems connected with adsorption under dynamic conditions have been dealt with separately. The nitrogen adsorption isotherms have been plotted and the heat of adsorption calculated. (authors) [French] Une methode d'epuration, a temperature moderement basse (comprise entre -50 et -80 deg. C) de l'helium servant de couverture inerte a l'eau lourde des reacteurs a ete etudiee. L'emploi au tamis moleculaire 5A a ete retenu pour la superiorite de celui-ci sur d'autres adsorbants dans ce domaine de temperatures. Les problemes particuliers a l'adsorption en regime dynamique ont ete separement traites. Les isothermes d'adsorption d'azote ont ete tracees et la chaleur d'adsorp. tion calculee. (auteurs)

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)

The control equipment of the Melusine II reactor; L'equipement de controle du reacteur Melusine II

Energy Technology Data Exchange (ETDEWEB)

Cordelle, M; Delcroix, V; Denis, P; Gariod, R

1963-07-01

Melusine II, low-power reactor, used for the study of Siloe core has diverged at the CEA Grenoble, the 23. May 1962; its monitoring board studied and carried out in this center is the first in France to be entirely transistorized. The first months of running have justified the hope put in the new electronics to improve the stability and the safety of running. The article describes the design of the control and gives the main characteristics of the measurement chains and of the actions on reactivity. (O.M.) [French] Melusine II, reacteur de faible puissance destine a l'etude du coeur de Siloe a diverge au Centre d'Etudes Nucleaires de Grenoble, le 23 mai 1962, son tableau de controle etudie et realise dans ce Centre est le premier en France a etre entierement transistorise. Les premiers mois de fonctionnement ont justifie l'espoir mis dans la nouvelle electronique pour ameliorer la stabilite et la surete de fonctionnement. L'article decrit la conception du controle et donne les principales caracteristiques des chaines de mesure et des actions sur la reactivite. (auteurs)

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

New Instruments and Principles for the Dimensional Measurement and Measurement of Spacing of Reactor Components; Nouveaux Instruments et Procedes de Mesure des Dimensions et de l'Espacement des Elements d'un Reacteur; Novye pribory i printsipy izmereniya razmerov i raspolozheniya komponentov reaktora; Nuevos Instrumentos y Principios para Medir las Dimensiones y la Separacion Entre Componentes de Reactor

Energy Technology Data Exchange (ETDEWEB)

Mueller, P. [Institut Dr. Foerster, Reutlingen, Federal Republic of Germany (Germany)

1965-09-15

instrument for reactor components are discussed. Special attention is given to the possibility of using a small and versatile pick-up by means of manipulators in the ''hot'' zones and on ''hot'' materials. The increase of surface roughness with increasing irradiation dose is discussed. (author) [French] Full text: L'auteur presente les problemes de mesure de l'epaisseur de feuilles et des parois de tubes et recipients en aciers austenitiques ou en metaux non ferreux. Deux methodes de mesure des epaisseurs sans contact sont discutees: la mesure, par courants de Foucault, de l'epaisseur de feuilles et des parois de recipients en metaux non ferreux ou en aciers austenitiques, au moyen de bobines se deplacant le long des pieces a examiner: la mesure, par courants de Foucault, de l'epaisseur des parois de tubes, au moyen de bobines dans lesquelles se deplacent les pieces a examiner. L'auteur decrit des instruments appropries et le mode d'utilisation. Il discute egalement la mesure de l'epaisseur des parois de parties constitutives de reacteurs, en metaux non ferreux, par la 'methode de la bille magnetique' et explique le principe de ce nouveau type de mesure et son domaine d'utilisation - notamment pour les mesures par points; il decrit un instrument approprie. L'auteur examine la mesure des revetements non magnetiques de materiaux magnetiques; il explique les principes de mesure (methodes fondees sur les champs magnetiques des courants continus et des courants alternatifs) et decrit des instruments de mesure de revetements non magnetiques dont l'epaisseur varie entre 3 {mu}m et 20 mm. Il expose le probleme special de la mesure des depots de stellite sur les parois en aciers ferritiques des cuves de reacteurs. La mesure des revetements non conducteurs de metaux non ferreux est etudiee. Le memoire explique le principe de mesure (courants de Foucault). Il decrit un instrument approprie et donne des exemples de mesures typiques. L'auteur examine egalement la mesure sans contact, en

Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

The industrial production of fuel elements; La fabrication en france des elements combustibles

Energy Technology Data Exchange (ETDEWEB)

Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)

1964-07-01

-pool type reactors. The authors show how the problem of the industrial production of rolled fuel elements has been solved in France, and give the three steps involved: 1 - Assembly of the plates made in the U.S.A., 2 - Rolling of the cores made in the U.S.A. to obtain the plates, 3 - Fabrication of the U-Al alloy and production of the cores. They then recall briefly the characteristics of the different fuel elements now in production. A description is given of the various stages of the production including information about the equipment; stress is laid on the extent of the controls carried out at each stage. In conclusion the authors consider the future development of this type of production taking into account the improvements planned and those which are possible. (authors) [French] Les auteurs traitent successivement de la fabrication industrielle des elements combustibles pour reacteurs de puissance de la filiere U naturel graphite-gaz et plus particulierement pour les centrales energetiques d'E.D.F. et de celle des elements combustibles a base d'U enrichi destines aux reacteurs experimentaux du type 'piscine'. 1ere Partie - LES ELEMENTS COMBUSTIBLES AVANCES POUR LES REACTEURS E.D.F.: Apres un bref rappel des caracteristiques des elements combustibles actuellement fabriques industriellement pour les reacteurs de MARCOULE et de CHINON, les auteurs indiquent les differentes etapes suivies pour aboutir au stade de la fabrication industrielle d'un element combustible nouveau, tant en ce qui concerne la gaine et eventuellement la chemise de graphite que le combustible lui-meme. Pour ce qui est de l'elaboration du combustible, ils decrivent les differentes operations en insistant sur les points originaux de la fabrication et de l'appareillage tels que: - coulees en moules chauds, - traitement thermique des alliages U.Mo 1 p. 100, - soudure des pastilles de fermeture des tubes, - gainage - controle aux differents stades. En ce qui concerne la fabrication des gaines, ils

Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor; Etude des consequences de la rupture d'un tube de force dans la cuve d'un reacteur modere a l'eau lourde et refroidi au gaz

Energy Technology Data Exchange (ETDEWEB)

Hareux, F; Roche, R; Vrillon, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

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

Study of a Slightly Enriched R Reactor Fuel by Means of a Pulsed Neutron Source; Etude d'un reacteur a combustible legerement enrichi (rubeole) a l'aide de sources pulsees de neutrons

Energy Technology Data Exchange (ETDEWEB)

Sagot, M.; Tellier, H. [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1962-04-01

A Be O moderated reactor using slightly enriched uranium oxide as fuel was studied by the pulsed neutron source technique. The neutron lifetime was measured in two different cores without reflector, then attempts were made at the measurement of great negative reactivities introduced into the reactor under the following forms: decrease of the volume of the un reflected core, introduction of absorbing cadmium rods, removal of fuel at the periphery of the critical core while maintaining a constant height, and substitution of fuel elements by less reactive elements. In all cases, the results are compared with the data obtained by another type of experiment or by computation. (author) [French] Nous avons applique la methode des sources pulsees de neutrons a un reacteur utilisant de l'oxyde d'uranium legerement enrichi, modere a l'oxyde de beryllium et, apres avoir mesure le temps de vie des neutrons dans deux coeurs differents non reflechis, nous avons porte notre effort, sur la mesure de reactivites negatives importantes introduites dans le reacteur sous differentes formes: - diminution du volume du coeur non reflechi, - introduction de barres absorbantes en cadmium, - enlevement de combustible a la peripherie du coeur critique, tout en conservant une hauteur constante, - substitution d'elements de combustible par des elements moins reactifs. Dans tous les cas, les resultats sont compares aux valeurs obtenues par un autre type d'experience ou par le calcul. (auteur)

Radiation hazards in the neighbourhood of uranium reactors; Dangers des rayonnements aupres des piles a uranium

Energy Technology Data Exchange (ETDEWEB)

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

1956-07-01

Radiation hazards near uranium reactors may be divided in two groups. Hazards when the reactor is normally operating: {gamma} radiation from hot uranium or air contamination by fission gases, {gamma} radiation or contamination by the coolant (air, nitrogen, heavy-water), {gamma} radiation from radioisotopes. Hazards in the case of an accident: presence of hot uranium in the atmosphere, soil contamination. (author) [French] Les dangers d'irradiation aupres des piles a uranium sont a classer essentiellement en deux groupes. Les dangers existant aupres d'une pile exploitee normalement: irradiation {gamma} par l'uranium irradie ou contamination de l'air par des gaz de fission, irradiation {gamma} ou contamination par les fluides de refroidissement (air, azote, eau lourde), irradiation {gamma} par les radioelements fabriques. Les dangers en cas d'accident survenant a un reacteur en fonctionnement, ayant pour consequence : la presence dans l'air d'uranium irradie, la contamination du sol. (auteur)

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.)

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

Operating Experience with the BR-5 Reactor; Experience acquise aupres du reacteur BR-5; Opyt ehkspluatatsii reaktora BR-5; Experiencia practica con el reactor BR-5

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A. I.; Kazachkovskij, O. D.; Pinkhasik, M. S.; Aristarkhov, N. N.; Karpov, A. V.; Larin, E. P.; Efimov, I. A.

1963-10-15

The paper discusses the carrying-out of repair and maintenance work on the radioactive liquid-metal circuit of the BR-5 fast neutron reactor. Attention is also given to problems of reactor operation after achievement of the planned 2% fuel burn-up with some disturbance of leak-tightness in individual fuel elements. An account is given of experience in discharging the active section, examining the condition and leak-tightness of the fuel elements, and decontaminating the equipment and piping of the first radioactive circuit after reaching 5% fuel burn-up. (author) [French] Dans ce memoire les auteurs decrivent l'execution des reparations et des travaux d'entretien dans le circuit radioactif liquide-metal du reacteur a neutrons rapides BR-5. Ils etudient egalement les problemes lies au fonctionnement du reacteur au taux de combustion de 2% prevu avec quelques defauts d'etancheite dans des elements combustibles particuliers. Ils decrivent le dechargementen zone active et examinent les conditions d'etancheite des elements combustibles. Ainsi que la decontamination de l'appareillage et des tuyauteries du premier circuit radioactif apres avoir atteint un taux de combustion de 5%. (author) [Spanish] En la memoria se examinan los problemas planteados por el mantenimiento del circuito radiactivo de metal liquido del reactor de neutrones rapidos BR-5. Se tratan cuestiones relacionadas con la explotacion del reactor una vez alcanzado el grado de combustion de 2%, previsto en el proyecto y luego de producirse ciertas alteraciones de la densidad de determinados elementos combustibles. Se describen la experiencia adquirida durante la descarga del cuerpo del reactor, las investigaciones del estado general y de la hermeticidad de los elementos combustibles y las operaciones de descontaminacion de la instalacion y de las tuberias del circuito radiactivo primario despues de alcanzado un grado de combustion de 5%. (author) [Russian] V doklade rassmatrivayutsya voprosy proizvodstva

Improvements of the sensitivity of burst cartridge detection; Amelioration du seuil de detection des ruptures de gaine

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

I - Special tests for improving the sensitivity of burst cartridge detection equipment in power reactors II - Scintillator purge-flow tests using aged gas in the B.C.D. /E.D.F. 2 Summary. - The first part of this report describes the tests carried out on fission product detectors by a process in which gas is continuously injected in front of the scintillator. Using this system, the background is reduced and perturbations caused by pneumatic switches on the prospecting circuits are eliminated. The quality of the signals thus obtained permits better processing of the data and thus leads to a possible improvement in the sensitivity of burst cartridge detection. The second part gives results of tests carried out with both fresh and aged gases, the economic advantage of the latter being that it permits recycling through the reactor. Reduction of the background is less pronounced but the advantage of the stable signals is conserved. (author) [French] I - Essais speciaux pour ameliorer le seuil de detection des installations de D.R.G. des reacteurs de puissance II- Essais de balayage sous scintillateur avec gaz vieilli a la D.R.G. /E.D.F. 2 Sommaire. - La premiere partie de ce rapport decrit les essais effectues sur les detecteurs de produits de fission par un procede d'injection continue de gaz sous le scintillateur. Grace a ce systeme on obtient une reduction du bruit de fond et l'elimination des perturbations causees par les commutations pneumatiques des circuits de prospection. La qualite des signaux obtenus ainsi permet un meilleur traitement des informations d'ou une amelioration possible du seuil de detection des ruptures de gaines. La seconde partie donne les resultats d'essais effectues avec du gaz propre et vieilli, l'utilisation de ce dernier presentant l'avantage economique d'etre recycle du reacteur. La reduction du bruit de fond est moins importante mais on conserve l'avantage de la stabilisation des signaux. (auteur)

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)

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)

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

Energy Technology Data Exchange (ETDEWEB)

Roche, R; Gaudez, J C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration, l'equilibrage des pression entre l'eau lourde et le gaz, le montage

Improvements in gas supply systems for heavy-water moderated reactors; Etudes de perfectionnements aux systemes d'alimentation en gaz d'un reacteur modere a l'eau lourde

Energy Technology Data Exchange (ETDEWEB)

Aubert, G; Hassig, J M; Laurent, N; Thomas, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [French] Dans un reacteur modere a l'eau lourde et refroidi au gaz sous pression, un probleme important du point de vue du trace du bloc pile et de son economie est le choix du systeme d'alimentation en gaz. Pour une solution a tubes de force, l'ensemble des structures du bloc reacteur est a temperature relativement faible, alors que les organes d'alimentation en gaz sont a celle, notablement plus elevee, du gaz. Ces organes, traverses par le debit du caloporteur, doivent lui opposer le minimum de resistance afin de ne pas necessiter un supplement onereux de puissance de

Instrumentation for Sodium Circuits; Instrumentation des Circuits de Sodium

Energy Technology Data Exchange (ETDEWEB)

Cambillard, E. [CEA, Centre d' Etudes Nucleaires de Fontenay-aux-Roses (France); Lions, N. [CEA, Centre d' Etudes Nucleaires de Cadarache (France)

1967-06-15

RAPSODIE. A description is given of the modifications carried out in connection with the mechanical zero adjustment and the measurement chain. (author) [French] Les instruments de mesure qui ont ete principalement etudies et experimentes au CEA pour les reacteurs ''a sodium comportent des debitmietres electromagnetiques, des indicateurs de niveau et des manometres differentiels. Les auteurs donnent les caracteristiques principales des debitmietres du reacteur RAPSODIE, qui sont a aimant permanent ou a electro -aimant (sur les circuits primaires). Ils decrivent les methodes d'etalonnage utilisees qui font appel a des diaphragmes ou des Venturis comme debitmietres etalons et indiquent les resultats de mesure obtenus pour des debits de sodium maximaux de 400 m{sup 3}/h. Trois types d'indicateurs continus de niveau ont ete etudies: Indicateur a resistance. Les auteurs decrivent deux variantes equipant les circuits d'essai de RAPSODIE de 1 et 10 MW. L'une comporte une resistance de compensation disposee sur toute la hauteur de l'element de mesure (les indicateurs continus du reacteur RAPSODIE sont actuellement de ce type). L'autre possede un dispositif permettant le chauffage d e l ''element de mesure en vue d {sup e}mpecher la formation- eventuelle de depots conducteurs (les essais en sodium de prototypes sont termines). Indicateur a induction Il comprend deux bobines couplees et un dispositif permettant une compensation des effets de temperature. Les auteurs decrivent le prototype qui a ete construit et indiquent les resultats obtenus au cours des essais en sodium. Indicateur ultra-sons. Il est caracterise par l'utilisation d'un transmetteur place en haut et a l'exterieur de la cuve de sodium, et d'un guide d'ondes vertical dont l'extremite inferieure plongeant dans le metal liquide possede un systeme reflechissant qui renvoie le faisceau ultra-sonore vera la surface. Des reperes fixes permettent un etalonnage permanent; l'ensemble de l'appareil est entierement soude. Cet

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.)

Operating Experience with the VERA Zero-Energy Fast Reactor; Fonctionnement du Reacteur VERA a Neutrons Rapides, de Puissance Zero; Opyt ehkspluatatsii reaktora VERA na bystrykh nejtronakh nulevoj moshchnosti; Experiencia Adquirida con el Reactor Rapido VERA de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Weale, J. W.; McTaggart, M. H.; Goodfellow, H.; Paterson, W. J. [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

1964-02-15

The design of a two-halves zero-energy fast reactor is briefly described, particular emphasis being placed on those features which determine the practicability and precision of reactor physics measurements. The advantages and disadvantages of the design are discussed with reference to the two years' operating experience of the reactor. The following topics are dealt with: the experimental convenience of the lay-out and of the two halves design; the size and precision of the fuel pieces and the accuracy of location of the fuel elements; the effects of edge irregularities and heterogeneity of structure on the accuracy with which the critical mass of an 'ideal' equivalent assembly is determined; reproducibility of the critical condition after dismantling the assembly, or separating the two halves; variation of reactivity with separation of the halves, including effects of asymmetric loading; sensitivity of various counters, neutron source strength, use of an accelerator neutron source; speed of response of safety circuits and consequent restrictions on rate of assembly of the two halves; additional precautions necessary in using plutonium fuel; and notes on the accuracy of measurement of reactivity and on the practical limitations affecting various other reactor physics measurements. (author) [French] Les auteurs decrivent brievement ce modele de reacteur a neutrons rapides et de puissance zero construit en deux moities, en insistant particulierment sur les caracteristiques qui determinent la possibilites de faire des mesures relatives a la physique des reacteurs et la precision de ces mesures. Ils exposent les avantages et les inconvenients de ce modele compte tenu de l'experience acquise au cours des deux annees de fonctionnement du reacteur. Ils traitent les sujets suivants: interet pratique, au point de vue experimental, du plan de ce reacteur et de sa constitution en deux moities; dimension et precision des pieces de combustible et exactitude de l'emplacement des

Methods of Containment Adopted for the EL4 Reactor and Projected Heavy-Water, Gas-Cooled Plants; Mode de Confinement Adopte pour le Reacteur EL4 et les Projets de Centrales Eau Lourde-Gaz

Energy Technology Data Exchange (ETDEWEB)

Schulhof, P.; Justin, F. [Commissariat a l' Energie Atomique, Paris (France)

1967-09-15

After a brief description of the plant, the paper explains the principles adopted for preventing the release of waste gas, from the EL4 reactor and refers to some of the difficulties associated with this type of containment. From the economic standpoint, the authors present the results of a comparative civil engineering study of pre-stressed concrete and steel shells for a projected 60 MW(e) power station, giving various values for accidental pressures. They demonstrate the influence of the stress values adopted. (author) [French] Les auteurs rappellent les principes adoptes dans le reacteur EL4 pour le confinement des rejets gazeux, apres une description sommaire des installations. Suivent quelques aspects des difficultes introduites par ce type de confinement. Dans le domaine economique, ils presentent le resultat d'une etude comparative de genie civil d'enceintes en beton precontraint et en acier pour un projet de centrale de 600 MW(e), avec diverses valeurs de pression accidentelle. Dans cette etude, ils font ressortir l'influence des valeurs admises pour le taux de travail des materiaux. (author)

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

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

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.

General views about specimen irradiations in reactors; Considerations generales sur'les irradiations d'echantillons dans les reacteurs

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

Specimen irradiation of fissile or non-fissile materials, carried out under circumstances becoming more and more severe and in reactor of increasing flux bas led to an evolution of irradiation rigs. A survey of the problems arising from irradiating under these various circumstances leads to conclude that it is possible to devise one capsule type suitable to every particular case, and that in a wide temperature range. Consequently, once the various irradiation-parameters known, a general method of calculation can be followed so as to determine the various sizes of the parts constituting the capsule. These theoretical calculations might sometimes be corrected through benefits gained from previous irradiations. Similarly, practical experimentation might allow to foresee more handy assembling of the capsule, specimen loading-and unloading being easier at the same time. (author) [French] L'irradiation d'echantillons, fissiles ou non fissiles, dans des conditions imposees de plus en plus strictes et dans des reacteurs a flux de plus en plus eleve, a eu pour consequence une evolution dans la conception des dispositifs d'irradiation. Lorsqu'on examine les problemes souleves par ces differentes irradiations, on en conclut qu'il est possible de concevoir un type de capsule capable de donner satisfaction dans chaque cas particulier, et ce, dans une tres large gamme de temperature. Par consequent, les differents parametres de l'irradiation etant connus, une methode generale de calcul peut etre suivie pour determiner les differentes cotes des pieces constitutives de la capsule. Ces calculs theoriques devront quelquefois etre corriges grace aux enseignements tires d'irradiations precedentes. De meme, l'experience acquise permettra d'envisager un montage plus aise de la capsule, tout en facilitant l'enfournement et le defournement des echantillons.

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

The experimental nuclear reactor: AQUILON; Le reacteur nucleaire experimental: AQUILON

Energy Technology Data Exchange (ETDEWEB)

Girard, Y; Koechlin, J C; Moreau, J M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

Information derived from French studies and achievements in the field of uranium isotope separation; Enseignements tires des etudes et realisations francaises relatives a la separation des isotopes de l'uranium

Energy Technology Data Exchange (ETDEWEB)

Frejacques, C; Galley, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The work carried out in the field of uranium isotope separation, by gaseous diffusion and by ultracentrifugation, is reviewed. An economic estimate of the various parameters involved in the cost is given, and it is shown that only very large gaseous diffusion plants, corresponding to a programme of enriched uranium reactors of at least 4000 MWe to be installed yearly, can give an economically acceptable enriched uranium production. (authors) [French] La communication passe en revue les realisations effectuees dans le domaine de la separation des isotopes de l'uranium, par diffusion gazeuse et par ultracentrifugation. Elle donne une estimation economique des differents parametres intervenant dans les couts et met en evidence que seules les tres grandes usines de diffusion gazeuse, correspondant a un programme d'installation de reacteurs a uranium enrichi d'au moins 4000 MWe nouveaux par an, peuvent conduire a des productions d'uranium enrichi economiquement acceptables. (auteurs)

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

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)

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.

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)

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.

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

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)

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)

Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

The Economical Application of Non-Destructive Testing to Reactor Components, Especially Jacket Tubing; Avantages Economiques du Controle Non Destructif des Pieces de Reacteurs, Notamment des Tubes de Gainage; Ehkonomicheskoe primenenie nedestruktivnykh ispytanij dlya reaktornykh komponentov, v chastnosti obolochechnykh trub; Aplicacion en Condiciones Economicas de Ensayos No Destructivos a las Piezas de los Reactores, en Especial a los Tubos de Revestimiento

Energy Technology Data Exchange (ETDEWEB)

Renken, C. J. [Metallurgy Division Argonne National Laboratory Argonne, IL (United States)

1965-10-15

electro-magnetic method for technical as well as economic reasons. The optimum area of application of these two methods is explained as well as the large area of overlap where results produced by well- designed and properly operated equipment of both types are essentially equivalent. Spurious defect indications contribute directly to increased component costs, so an evaluation of these effects for both the ultrasonic and the electromagnetic test methods is included for several commonly encountered sources of spurious defect signals. The experience in the application of these methods at Argonne National Laboratory on relatively large quantities of tubing from various sources are recounted from the standpoint of the lowest possible inspection cost per unit length of tubing. This section also summarizes experience gained at Argonne with the newer pulsed electromagnetic test methods. The critical but generally unappreciated role of tube diameter and wall thickness on tube inspection cost is discussed. Since the question of economical inspection is closely related to allowable defect levels, defect levels and standards in use at Argonne are covered. Finally, the practical and theoretical barriers to reduced component inspection costs are enumerated and a projection of what possible reductions in cost might be attainable in the future with the ultrasonic and electromagnetic test methods is attempted. (author) [French] Le reacteur ideal aurait entre autres caracteristiques celle de ne pas exiger de controles non destructifs. Cet ideal, comme tant d'autres, ne sera probablement jamais atteint. Dans l'etude de tout reacteur pour lequel le prix de revient constitue un facteur important, il faudrait envisager la question de savoir si les pieces de ce reacteur pourront etre essayees de facon economique en meme temps que l'on examine les possibilites de fabrication. Cette partie du memoire contient quelques considerations a ce propos ainsi qu'un expose de l'importance des essais non

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)

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.

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

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

Energy Technology Data Exchange (ETDEWEB)

Roche, R.; Gaudez, J.C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration

Techniques d'inspection par ondes guidees ultrasonores d'assemblages brases dans des reacteurs aeronautiques =

Science.gov (United States)

Comot, Pierre

L'industrie aeronautique, cherche a etudier la possibilite d'utiliser de maniere structurelle des joints brases, dans une optique de reduction de poids et de cout. Le developpement d'une methode d'evaluation rapide, fiable et peu couteuse pour evaluer l'integrite structurelle des joints apparait donc indispensable. La resistance mecanique d'un joint brase dependant principalement de la quantite de phase fragile dans sa microstructure. Les ondes guidees ultrasonores permettent de detecter ce type de phase lorsqu'elles sont couplees a une mesure spatio-temporelle. De plus la nature de ce type d'ondes permet l'inspection de joints ayant des formes complexes. Ce memoire se concentre donc sur le developpement d'une technique basee sur l'utilisation d'ondes guidees ultrasonores pour l'inspection de joints brases a recouvrement d'Inconel 625 avec comme metal d'apport du BNi-2. Dans un premiers temps un modele elements finis du joint a ete utilise pour simuler la propagation des ultrasons et optimiser les parametres d'inspection, la simulation a permis egalement de demontrer la faisabilite de la technique pour la detection de la quantite de phase fragile dans ce type de joints. Les parametres optimises sont la forme de signal d'excitation, sa frequence centrale et la direction d'excitation. Les simulations ont montre que l'energie de l'onde ultrasonore transmise a travers le joint aussi bien que celle reflechie, toutes deux extraites des courbes de dispersion, etaient proportionnelles a la quantite de phase fragile presente dans le joint et donc cette methode permet d'identifier la presence ou non d'une phase fragile dans ce type de joint. Ensuite des experimentations ont ete menees sur trois echantillons typiques presentant differentes quantites de phase fragile dans le joint, pour obtenir ce type d'echantillons differents temps de brasage ont ete utilises (1, 60 et 180 min). Pour cela un banc d'essai automatise a ete developpe permettant d'effectuer une analyse similaire

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

Optimal sizes and siting of nuclear fuel reprocessing plants; Tailles et localisations optimales des usines de retraitement des combustibles nucleaires

Energy Technology Data Exchange (ETDEWEB)

Thiriet, L; Deledicq, A [Commissariat a l' Energie Atomique, Siege (France). Centre d' Etudes Nucleaires

1967-07-01

traite le probleme des usines de traitement de l'uranium naturel irradie associees a des centrales nucleaires a uranium naturel-graphite CO{sub 2}. La localisation et la production annuelle des reacteurs, les sites possibles d'usines et les fonctions de cout (transport et retraitement) sont supposes connus. La methode consiste a traiter d'abord le probleme des usines de traitement comme un probleme de programmation dynamique, des tranches croissantes de programmes (production totale des reacteurs) etant explorees sequentiellement. Lorsque les quantites d'uranium naturel irradie a retraiter sont fixees, la minimisation du cout de transport est alors effectuee, elle aussi comme un probleme de programmation dynamique. On explore le voisinage de l'optimum du cout de traitement pour trouver le minimum de la somme d'un cout de traitement sous-optimal et du cout de transport optimal correspondant. Le probleme de retraitement etant representable sur un graphe sequentiel, l'algorithme utilise pour calculer les sous-optima est 'l'algorithme a reflexion' que nous avons elabore. La methode s'interprete comme un mecanisme general de determination de l'optimum lorsque, a un probleme dynamique sequentiel (par exemple un programme d'equipement), se superpose un probleme complementaire (par exemple de transport). Elle permet en outre d'evaluer les pertes resultant du choix, pour des raisons autres qu'economiques, d'une politique non optimale. (auteur)

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.

Long term developments in irradiated natural uranium processing costs. Optimal size and siting of plants; Perspectives a long terme des couts de traitement de l'uranium naturel irradie. Tailles et localisations optimales des usines

Energy Technology Data Exchange (ETDEWEB)

Thiriet, L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Oger, C; Vaumas, P de [Saint-Gobain Nucleaire, 92 - Courbevoie (France)

1964-07-01

processing plants are shown, different from those in part two. The indirect effect of these reprocessing programmes on the availability of plutonium, and therefore on the possibility.of undertaking plutonium burning reactor programmes, must be taken into account. (authors) [French] L'objet de cette communication est d'apporter une contribution a la solution du probleme du choix des tailles et des localisations optimales des usines de traitement des combustibles nucleaires irradies, associees a des programmes de puissance electrique installee. Dans une premiere partie, on etudie la structure des couts d'investissements et d'exploitation des usines de traitement de l'uranium naturel irradie, l'influence de la taille des usines sur ces couts et ces structures de couts. Au cout de traitement de l'uranium naturel irradie s'ajoute d'autre part le cout du transport des combustibles irradies des lieux de production aux sites des usines de traitement. La recherche du cout minimum pour la production d'un pays ou d'un ensemble de pays fait donc intervenir a la fois la taille et la localisation des usines. On indique les couts de transport previsibles pour l'uranium naturel irradie et la structure de ces couts (transport, assurance, couts et amortissement des containers). Dans une deuxieme partie, et pour differents echeanciers de combustibles irradies a traiter chaque annee, on determine les tailles et les localisations optimales des usines de traitement et la sensibilite de ces resultats, aux hypotheses de base concernant le cout du traitement, le cout du transport, l'annee de demarrage du programme d'usines, l'horizon choisi. - le probleme de nature combinatoire, assez complexe, est resolu par l'application des methodes de la programmation dynamique. - on montre que les methodes sont egalement applicables au probleme du choix des tailles et des localisations optimales des usines de traitement des elements du type MTR, associees aux programmes de reacteurs de recherche ainsi qu

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.)

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

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)

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

Statistics on the production and the use of the artificial radioelements in France; Statistiques sur la production et l'emploi des radioelements artificiels en France

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

The CEA is, in France, the unique producer of artificial radioelements for public uses. These products have been provided to the users since 1949. They include until now only radioelements formed in nuclear reactors. The following aspects of use in France of the artificial radioelements will be described: - Consumption of the artificial radioelements in France, - French production and import, - Teaching and study of applications. (M.B.) [French] Le Commissariat a l'Energie Atomique est, en Franoe, le seul producteur de radioelements artificiels pour l'utilisation publique. Ces produits ont ete fournis aux utilisateurs des 1949. Ils ne comprennent jusqu'a present que des radioelements formes dans des reacteurs nucleaires. Les aspects suivants de l'utilisation en France des radioelements artificiels seront decrits: onsommation des radioelements artificiels en France, Production francaise et importation, - Enseignement et etudes d'applications. (M.B.)

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

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)

Measurement of the thermal utilisation factor of the reactor G1; Mesure du facteur d'utilisation thermique du reacteur G1

Energy Technology Data Exchange (ETDEWEB)

Roullier, F; Schmitt, A P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The thermal utilisation factor of the lattice of the reactor G1 has been measured by applying the autoradiographic technique to thin detectors irradiated in the cell. The experimental apparatus is described, and the results compared with those obtained by calculation based on various formulae. The results of the study of the thermal flux distribution in a cell containing a thorium rod of the same diameter as the uranium rods in the lattice are also given. The precision of the measurements is discussed. Value found: f diameter 26 = 0.8949 {+-} 0,005. (author) [French] Le facteur d'utilisation thermique du reseau du reacteur G1 a ete mesure en appliquant la technique de l'autoradiographie a des detecteurs minces irradies dans la cellule. Les dispositifs experimentaux sont decrits et les resultats sont compares a ceux obtenus par le calcul a partir de diverses formules. Les resultats de l'etude de la distribution du flux thermique dans une cellule contenant une barre de thorium de meme diametre que les barres d'uranium du reseau sont egalement indiques. La precision des mesures est discutee. Valeur trouvee: f diametre 26 = 0,8949 {+-} 0,005. (author)

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.

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.

Space-time dependent impulse response of a subcritical cylindrical reactor; Reponse impulsionnelle spatio-temporelle d'un reacteur cylindrique en regime sous-critique

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

In this paper, a new formulation of the spatial dependent impulse response of a subcritical reactor in a cylindrical geometry is proposed. An expression of the transfer function between a point source at the center of coordinates and the flux at a given point (r,z) is obtained by solving: by means of Laplace transform, the one group diffusion equation. In this transfer function, variables r and p (p being the Laplace variable) remain linked within a modified Bessel function. Taking the inverse Laplace transform is done by two different ways: - using the Mellin-Fourier method which separates variables r and t. This method makes it possible to establish that there is identity between the classical formulation and the new one. - using an inverse Laplace transform which keeps variables r and t linked. This method requires to approximate the inverse Laplace transform of the end factor. It is then possible to replace the radial harmonics modes series of the classical expression by a single function. This new formulation seems to be of particular interest when dealing with reactors of large size and lifetime. It is also interesting each time the harmonics play an important role. (author) [French] Dans le present rapport, on propose une nouvelle formulation de la reponse impulsionnelle spatio-temporelle d'un reacteur sous-critique, en geometrie cylindrique. Une expression de la fonction de transfert entre une source ponctuelle placee au centre des coordonnees et le flux au point courant (r,z) est obtenue en resolvant, par transformation de Laplace, l'equation de la diffusion a un seul groupe d'energie. Dans cette fonction de transfert, les variables r et p (variable de Laplace) demeurent groupees dans une fonction de Bessel modifiee. Le retour a l'original est effectue de deux manieres: - la methode de Mellin-Fourier qui separe les variables r et t, permet d'etablir l'identite entre la nouvelle formulation et la formulation classique. - un original conservant les variables

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)

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.)

CO{sub 2} direct cycles suitable for AGR type reactors; Cycles directs de gaz carbonique applicables aux reacteurs du genre AGR

Energy Technology Data Exchange (ETDEWEB)

Maillet, E [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1967-10-01

The perspectives given by the gas turbines under pressure, to build simple nuclear power plants and acieving significantly high yield, are specified. The CO{sub 2} is characterised by by good efficiency under moderate temperature (500 to 750 Celsius degrees), compactness and the simpleness of machines and the safe exploitation (supply, storage, relief cooling, thermosyphon). The revision of thermal properties of the CO{sub 2} and loss elements show that several direct cycles would fit in particular to the AGR type reactors. Cycles that would diverge a little from classical models and able to lead to power and heat generation can lead by simple means to the best results. Several satisfying solutions present for the starting up, the power regulation and the stopping. The nuclear power plant components and the functioning safety are equally considered in the present report. The conclusions stimulate the studies and realizations of carbon dioxide gas turbines in when approprite. [French] Les perspectives offertes par la turbine a gaz sous pression, pour construire des centrales nucleaires simples et de rendement progressivement eleve, se precisent actuellement. le CO{sub 2} se distingue par sa bonne efficacite a temperature moderee (500 a 750 degres celsius), la compacite et la simplicite des machines, et la surete qu'il apporte a l'exploitation ( approvisionnement, stockage, refroidissement de secours, thermosiphon). La revision des proprietes thermophysiques du CO{sub 2} et des elements de pertes montre que divers cycles directs conviendraient en particulier aux reacteurs agr ou derives. Des cycles s'ecartant peu des modeles classiques, et se pretant ulterieurement a la production simultanee d'electricite et de chaleur, peuvent conduire par des moyens simples aux meilleurs resultats d'ensemble. Plusieurs solutions satisfaisantes se presentent pour le demarrage, le reglage de la puissance et l'arret. Les composants de la centrale et la surete de fonctionnement sont

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)

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

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

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

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.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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.

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

Development and testing of the EDF-2 reactor fuel element; Essais et mise au point de l'element combustible pour le reacteur EDF-2

Energy Technology Data Exchange (ETDEWEB)

Delpeyroux, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Furhmann, R [Societe Industrielle de Combustible Nucleaire (France)

1964-07-01

rassemble les etudes qui ont ete necessaires pour mener a bien la definition de l'element combustible EdF 2. Apres un bref rappel des caracteristiques du reacteur EdF 2 et des options preliminaires ayant permis de fixer un avant-projet d'element combustible, on aborde les etudes proprement dites: - Etudes uranium: essais de passage d'une couronne interne du tube en phase {beta}, flechage du tube sous l'action d'une force concentree, soudage des pastilles d'extremites et verification de leur etancheite. La tenue du tube a l'ecrasement et la resistance des pastilles a l'enfoncement sous l'action de la pression externe sont etudiees en detail dans un autre rapport CEA - Etudes gaine: rappel des conditions de fabrication et verification de l'etancheite de la gaine, tenue des ailettes au fluage sous l'action du courant gazeux - Etudes d'extremites: fluage en compression et soudage des bouchons a la gaine. - Etudes cartouche: determination des caracteristiques des gorges d'ancrage gaine-combustible et des conditions de gainage, verification de la tenue au cyclage thermique de l'element combustible, determination de la chute de temperature au contact gaine-combustible traitee en detail dans un autre rapport CEA, - Etudes de l'ensemble: les etudes se rapportant a la chemise de graphite, au support et aux vibrations de la cartouche ont ete traitees par le service des Etudes Mecaniques et Thermiques (Section de Mecanique), Dans ce domaine, la Section d'Etude d'Elements Combustibles a etudie la tenue des centreurs sous l'action du courant gazeux. L'aboutissement des etudes est constitue par le dessin de l'element combustible, le schema de fabrication et les normes de fabrication. La validite de l'ensemble de ces essais hors pile sera confirmee par des assais en pile qui sont en cours et par l'irradiation des elements dans le reacteur EdF 2 lui-meme. En conclusion, on donne l'orientation des etudes pour l'amelioration de l'element combustible et la definition d'un element combustible

Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

Measurements of reactivity of reactor G1; Mesures de reactivite sur reacteur G1

Energy Technology Data Exchange (ETDEWEB)

Bernot, J; Koechlin, J C; Portes, L; Teste du Bailler, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [French] Nous exposons et discutons diverses methodes utilisees, lors de l'etude physique du reacteur G1, pour determiner les variations du facteur de multiplication effectif consecutives a un changement donne dans la geometrie du milieu multiplicateur. La comparaison des resultats obtenus par diverses methodes nous a permis de tester leur validite et d'en preciser les conditions d'emploi. Dans une premiere partie, nous exposons les principes utilises et leurs domaines de validite. Dans une seconde partie nous donnons les resultats experimentaux obtenus avec quelques indications sur leur comparaison avec les estimations theoriques. (auteur)

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)

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)

Some particular problems put by operating experimental reactors; Quelques problemes particuliers poses par le fonctionnement des piles laboratoires

Energy Technology Data Exchange (ETDEWEB)

Candiotti, C; Mabeix, R; Uguen, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [French] Les redacteurs se basant sur six annees d'experience dans l'exploitation de reacteurs de recherche, exposent tout d'abord les differences d'utilisation entre ces engins et d'autres appareils fonctionnellement similaires et font ressortir, par voie de consequence, les servitudes correspondantes. Ces servitudes posent des problemes tres particuliers dans les domaines de l'exploitation proprement dite, de l'entretien, des modifications ou adjonctions apportees a l'ensemble. (auteur)

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

Materials Control in the Fabrication of Enriched Uranium Fuels; Controle des Matieres au Cours de la Fabrication des Combustibles a Base d'Uranium Enrichi; Uchet materialov pri izgotovlenii topliva na obogashchennom urane; Control de Materiales en la Elaboracion de Combustibles de Uranio Enriquecido

Energy Technology Data Exchange (ETDEWEB)

Cardwell, Jr., R. G. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

1966-02-15

measurement were successfully used where alloy fuel content was critical. Scrap handling had an important effect on the materials balance, by which fuel content was confirmed and good accountability was assured. Records and handling procedures, including batching and physical marking methods, were formulated in a manner that assisted the fabricator in criticality control. (author) [French] Grace aux efforts intenses qui ont ete accomplis au cours des 15 dernieres annees dans le domaine de la technologie des elements de combustible par le Laboratoire national d'Oak Ridge, il a ete possible d'etablir des methodes rationnelles de fabrication et de controle des combustibles eraichis, qui trouvent une iaige application dans la fabrication industrielle des elements de combustible a l'heure actuelle. Des techniques eprouvees de manipulation du combustible enrichi en alliages, en dispersion et sous forme d'oxyde en vrac ont ete mises au point et appliquees a l'etude et a la' fabrication des prototypes d'elements combustibles utilises pour le demarrage du reacteur d'essai de materiaux, du reacteur a protection constituee par la masse du ra- lentisseur ou reacteur piscine, du reacteur de puissance transportable construit pat V, du reacteur protection en tour, du reacteur expose a la Conference de Geneve, du reacteur a haut flux pour la production de radioisotopes et du reacteur experimental refroidi par un gaz. L'experience acquise est la base du present memoire qui traite essentiellement des problemes de controle des matieres qui se posent au cours de la fabrication de differents types d'elements de combustible a base d'uranium enrichi et montre comment ils ont ete resolus. Les objectifs principaux d'un systeme rationnel de controle des matieres sont les suivants: 1. reduire le plus possible le nombre des postes matiere a controler; 2. etablir des releves distincts pour chacune des phases principales des operations et les coordonner de maniere a pouvoir relever les ecarts avec un

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

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

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

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.)

Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors; Formulaire pour le calcul de la mecanique des empilements des reacteurs graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-07-01

This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [French] Le domaine de ce formulaire est strictement limite aux effets mecaniques, pour les empilements, des deformations, thermiques ou autres, des structures metalliques de soutien (aire - support et corset). On propose un ensemble de relations qui ont ete etablies a la suite des essais de CHINON sur des maquettes de grande taille. Ces relations permettent le calcul des mouvements, des deformations et des contraintes dans les empilements du type EDF, a reseau horizontal triangulaire regulier. (auteurs)

Study of transient states in thermo-ionic converters; Etude des regimes transitoires des convertisseurs thermoioniques

Energy Technology Data Exchange (ETDEWEB)

Landrot, J [Commissariat a l' Energie Atomique, 91 - Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

In order to control a thermo-ionic reactor, it is necessary to know the dynamic influence of four fundamental parameters: the injected thermal power, the electrical charge resistance, the temperature of the cesium and the thermal exchange coefficient of the collector cooling circuit. The principles of the thermo-ionic converter are briefly exposed. The over-riding influence of the first two parameters is shown with the help of experimental static readings. These two parameters are then made to vary in turn. The laws of variation as a function of the time, of the electrical power produced and of the temperature of the various parts of the converter are deduced. (author) [French] Pour envisager le controle et la regulation d'un reacteur thermoionique, il est necessaire de connaitre l'influence dynamique de quatre parametres fondamentaux: puissance thermique injectee, resistance electrique de charge, temperature de cesium et coefficient d'echange thermique du circuit de refroidissement du collecteur. On rappelle brievement les principes du convertisseur thermoionique. A l'aide de releves statiques experimentaux, on montre l'influence preponderante des deux premiers parametres. On fait ensuite varier successivement ces deux parametres. On met en evidence les lois de variation en fonction du temps de la puissance electrique produite et de la temperature des differents points du convertisseur. (auteur)

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)

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

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

EURATOM's Programme of Participation in Power Reactor Construction; Le programme de participation d'Euratom aux reacteurs de puissance; Programma uchastiya v razrabotke ehnergeticheskikh reaktorov Evratoma; El programa de participacion de la Euratom en la construccion y explotacion de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Ramadier, R. C.; Parker, E. [Communaute Europoenne de l' Energie Atomique, Bruxelles (Belgium)

1963-10-15

-years during which operating problems will become decisive for the development of atomic power. (author) [French] L'un des moyens mis en oeuvre par la Commission de l'Euratom en vue d'assurer le developpement d'une industrie nucleaire europeenne est un programme dit de ''participation communautaire''. Ce programme permet a la Commission de participer a concurrence de 32 millions d'u.c. AME a des realisations dans le domaine des reacteurs de puissance. La contrepartie est l'acquisition des informations relatives a la conception, la construction, le demarrage et le fonctionnement de ces reacteurs. Jusqu'a present des propositions emanant de trois societes ont donne lieu a la signature de contrats. Il s'agit de: a) la Societa Elettronucleare Nazionale (SENN) qui fait construire en Italie une centrale de 150 MW(e) nets equipee d'un reacteur a eau bouillante a double cycle; b) la Societa Italiana Meridionale Energia Atomica (SIMEA) qui a entrepris en Italie la construction d'une centrale de 200 MW(e) nets equipee d'un reacteur du type uranium naturel-graphite-gaz carbonique; c) la Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA) qui a entrepris a la frontiere franco-belge la construction d'une centrale equipee d'un reacteur a eau pressurisee d'une puissance qui pourra atteindre et probablement depasser 242 MW(e) nets. En outre, la Commission a e te saisie de demandes de participation a deux autres reacteurs de puissance presentees respectivement par le Groupement Rheinisch-Westfalisches Elektiizitatswerk-Bayernwerke (RWE-BW), et par la N.V. Samenwerkende Electriciteits-Productiebedrijve; la premiere pour un reacteur de 237 MW(e) a eau bouillante a double cycle, la seconde pour un reacteur de 50 MW(e) a eau bouillante a simple cyc le et circulation naturelle. La participation communautaire peut prendre des formes diverses. Elle peut en particulier prendre celle d'une participation au deficit eventuel de la production d'electricite des centrales pendant les premieres

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)

Two further years of operation of the reactor G1 (july 1958 - july 1960); Deux nouvelles annees de fonctionnement du reacteur G1. (juillet 1958 - Juillet 1960)

Energy Technology Data Exchange (ETDEWEB)

Mathot, P; Bauzit, J; Cante, R; Hebrard, L [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

observations ont pu etre faites sur l'empilement de graphite, en meme temps qu'etait accru le nombre de points de mesure des temperatures des gaines du combustible. - Du 25 septembre 1959 au 9 decembre 1959: preparation et execution du deuxieme recuit. A l'issue du recuit, le reseau de thorium a ete modifie et des thermocouples supplementaires donnant la temperature de la masse du graphite ont ete mis en place. Un appareillage permettant la mesure du flux radial a ete realise. - Du 9 decembre 1959 a juillet 1960: campagne de fonctionnement continu, avec le minimum d'arrets. Les resultats d'experience sont regroupes, independamment de toute chronologie sous trois grandes rubriques qui president a la vie du reacteur: - Fonctionnement continu, - Dechargements, - Recuits du reacteur. (auteur)

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.

Some particular problems put by operating experimental reactors; Quelques problemes particuliers poses par le fonctionnement des piles laboratoires

Energy Technology Data Exchange (ETDEWEB)

Candiotti, C.; Mabeix, R.; Uguen, R. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [French] Les redacteurs se basant sur six annees d'experience dans l'exploitation de reacteurs de recherche, exposent tout d'abord les differences d'utilisation entre ces engins et d'autres appareils fonctionnellement similaires et font ressortir, par voie de consequence, les servitudes correspondantes. Ces servitudes posent des problemes tres particuliers dans les domaines de l'exploitation proprement dite, de l'entretien, des modifications ou adjonctions apportees a l'ensemble. (auteur)

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.

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

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

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)

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)

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.)

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

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

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

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)

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

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

A critical summary of microscopic fast-neutron interactions with reactor structural, fissile and fertile materials; Apercu critique des interactions microscopiques des neutrons rapides avec les materiaux de construction et les matieres fissiles et fertiles utilisees dans les reacteurs; Kriticheskij obzor mikroskopicheskog o vzaimodejstviya bystrykh nejtronov s konstruktsionnymi, rasshcheplyayushchimis ya i vosproizvodyashchim i reaktornymi materialami; Resumen critico de las interacciones microscopicas de los neutrones rapidos con los materiales estructurales fisionables y fertiles utilizados en los reactores

Energy Technology Data Exchange (ETDEWEB)

Smith, A B [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

Prevailing knowledge of fast-neutron-induced reactions utilized in the nuclear design of reactor systems is reviewed. Principal emphasis is placed upon microscopic experimental methods, results and precisions. Fast-neutron scattering is considered in detail, including the results of experimental determinations of scattering from oxygen, iron, zirconium, niobium, tungsten, thorium and uranium. Representative results of experimental studies of fast-neutron capture and fast-neutron-induced fission are given. The measurements discussed not only provide results of considerable applied usefulness but axe also examples of the application of advanced experimental nuclear techniques. Areas of limited, conflicting or non-existent experimental information are outlined. A prognosis of future knowledge of fast-neutron reactions is made, with emphasis on the fulfillment of reactor requirements for basic nuclear data. (author) [French] L'auteur fait le point des connaissances sur les reactions provoquees par les neutrons rapides sur lesquelles on tend a fonder les projets de reacteurs. Il met en relief les methodes, les resultats et la precision de mesures experimentales a l'echelle microscopique. Il etudie en detail la diffusion des neutrons rapides, et donne les resultats de mesures experimentales de diffusion dans l'oxygene, le fer, le zirconium, le niobium, le tungstene, le thorium et l'uranium. Il donne les resultats les plus significatifs d'etudes experimentales sur la capture des neutrons rapides et sur la fission provoquee par des neutrons rapides. Les mesures etudiees, non seulement fournissent des renseignements d'une utilite pratique considerable, mais aussi constituent des exemples de l'application de techniques experimentales nucleaires a la pointe du progres. L'auteur indique les domaines ou les donnees experimentales sont limitees, contradictoires ou inexistantes. Il se livre a des pronostics sur le developpement des connaissances experimentales en matiere de

Rapsodie; Rapsodie

Energy Technology Data Exchange (ETDEWEB)

Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; D' Ayguevives, Ch [Groupement Atomique Alsacienne Atlantique (France); Sahl, W [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France)

1964-07-01

interesting studies were carried out on the safety of Rapsodie, and in particular scale models (1/3 and 1/10) were used to estimate the damage caused by the theoretical maximum nuclear accident. The construction of the reactor was started in the autumn of 1961. The industrial architects are the Groupement Atomique Alsacienne Atlantique (G. AAA). A brief outline of the current situation in the construction will be given. (authors) [French] Rapsodie, le premier reacteur a neutrons rapides construit on France dans le cadre d'une Association entre Euratom et le CEA, repond a un triple but: - Il servira d'abord de reacteur experimental dont on etudiera en detail le comportement en regime statique et dynamique, - Les enseignements retires de sa construction et de son fonctionnement serviront a developper la technologie des futurs reacteurs industriels a neutrons rapides. - Son flux de neutrons rapides sera suffisant pour lui permettre de servir aux essais sous irradiation d'elements combustibles pour les reacteurs a neutrons rapides suivants. Rapsodie ayant deja fait l'objet de descriptions, on se contentera d'en faire ici une presentation tres breve en insistant sur les points particulierement significatifs et sur les recentes modifications du projet. Seront successivement evoques: - les principales caracteristiques neutroniques et thermiques, - les assemblages combustibles et fertiles, - le bloc pile et les circuits de refroidissement, - les principaux moyens de manutention des assemblages, - les principes et moyens qui regissent la conduite et la surete du reacteur. La construction du reacteur a ete precedee par la realisation de maquettes en vraie grandeur de ses parties essentielles comprenant notamment: - un circuit complet de sodium do 10 MW, prototype des deux circuits qui equiperont le reacteur, - une maquette du bloc pile (cuve, fermeture superieure, structures internes du coeur et de la couverture), munie d'un circuit de sodium special permettant d'effectuer des essais d

Spatial flux instabilities, and their control in the graphite gas power reactors; Les instabilites spatiales du flux et leur controle dans les reacteurs de puissance graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Cailly, J L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Radial-azimuthal and axial spatial flux instabilities in graphite-gas reactors are studied by means of an analytical approach. Results are checked with those which are given by two dimensional (r, z and r, {theta}) kinetic models programmed for an IBM 7094 computer. At least, conclusions on the control of instabilities obtained from these models are reported. (author) [French] Les instabilites spatiales du flux dans les reacteurs graphite-gaz, radiales et azimutales d'une part, axiales d'autre part, sont etudiees au moyen d'une formulation analytique. Les resultats sont confrontes avec ceux que fournissent des modeles cinetiques a deux dimensions (r, z et r, {theta}) programmes sur IBM 7094. On donne enfin les conclusions relatives au controle de ces instabilites que ces modeles ont permis de degager. (auteur)

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

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.)

Study of transient states in thermo-ionic converters; Etude des regimes transitoires des convertisseurs thermoioniques

Energy Technology Data Exchange (ETDEWEB)

Landrot, J. [Commissariat a l' Energie Atomique, 91 - Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

In order to control a thermo-ionic reactor, it is necessary to know the dynamic influence of four fundamental parameters: the injected thermal power, the electrical charge resistance, the temperature of the cesium and the thermal exchange coefficient of the collector cooling circuit. The principles of the thermo-ionic converter are briefly exposed. The over-riding influence of the first two parameters is shown with the help of experimental static readings. These two parameters are then made to vary in turn. The laws of variation as a function of the time, of the electrical power produced and of the temperature of the various parts of the converter are deduced. (author) [French] Pour envisager le controle et la regulation d'un reacteur thermoionique, il est necessaire de connaitre l'influence dynamique de quatre parametres fondamentaux: puissance thermique injectee, resistance electrique de charge, temperature de cesium et coefficient d'echange thermique du circuit de refroidissement du collecteur. On rappelle brievement les principes du convertisseur thermoionique. A l'aide de releves statiques experimentaux, on montre l'influence preponderante des deux premiers parametres. On fait ensuite varier successivement ces deux parametres. On met en evidence les lois de variation en fonction du temps de la puissance electrique produite et de la temperature des differents points du convertisseur. (auteur)

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

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

Recent progress in the determination of gases in metals by emission spectrography (1963); Progres recents dans la determination des gaz dans les metaux par spectrographie d'emission (1963)

Energy Technology Data Exchange (ETDEWEB)

Rossi, G; Melamed, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

The Fassel method for analysis of bases in metals was used. By employing a smaller, completely cooled chamber greater sensitivity and lover blank values were obtained. A straight calibration curve, independent of the sample nature was established for oxygen using graphite pellets containing known amounts of ZnO and SnO{sub 2}. The calibration curve for hydrogen was obtained from previously analysed Zircaloy samples. In bath cases a lower limit of sensitivity of 15 {mu}g was attained. (authors) [French] On a utilise la methode Fassel pour l'analyse des metalloides dans les metaux. En employant une chambre de volume reduit avec refroidissement des parois, ou a pu abaisser la limite de sensibilite et reduire le blanc a une valeur negligeable. Une courbe de dosage pour l'oxygene, lineaire jusqu'a environ 15 {mu}g et independante de la nature de l'echantillon, a ete obtenue a partir des melanges oxyde de zinc + graphite et oxyde d'etain + graphite. Une courbe de dosage pour l'hydrogene lineaire jusqu'a 15 {mu}g a ete obtenue en partant d'echantillons de Zircaloy a teneur connue en hydrogene. (auteurs)

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.)

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

Single Pellet String Reactor for Intensification of Catalyst Testing in Gas/Liquid/Solid Configuration Réacteur catalytique de type “filaire” pour l’intensification de tests catalytiques en configuration gaz/liquide/solide

Directory of Open Access Journals (Sweden)

Hipolito A.I.

2010-09-01

has been shown that the pressure drop is controlled by the liquid/solid friction surface and that the pressure drop is not a limiting parameter in the reactor’s operation (values always lower than 0.1 bar. So, from a hydrodynamic point of view, this new reactor exhibits characteristics suitable for its use in catalytic tests. Finally, this reactor was implemented under reaction conditions to study hydrogenation reactions with a real industrial catalyst. The selective hydrogenation of allene was studied. The string reactor was shown to run isothermal kinetic tests with a very small amount of industrial-sized catalyst particles (less than 2 cc and to explore kinetics of fast reaction at high space velocities impossible to achieve in standard fixed bed units with appropriate hydrodynamic conditions. For constant residence time, the allene conversion does not vary with pressure and feed flow rate, which confirms that the string reactor allows one to perform catalytic tests with such a fast reaction without external mass transfer resistance. L’optimisation du catalyseur est une etape cle pour l’optimisation d’un procede catalytique du point de vue des rendements, de l’efficacite energetique et de la selectivite des reactions. La strategie de developpement d’un catalyseur comprend des tests effectues sur des reacteurs pilotes avec des charges reelles ou modeles. Cette etape a fait l’objet de nombreuses etudes ces dernieres decennies portant sur le dimensionnement des reacteurs, l’amelioration des outils d’analyses et les procedures operatoires. La plupart des etudes ont pour but de determiner l’activite catalytique de catalyseur sous forme de grain dans des conditions isothermes de facon a pouvoir determiner les parametres cinetiques de la reaction. Avec l’optimisation des catalyseurs, les flux de transfert de matieres externes aux grains peuvent devenir l’etape limitante, dans les reacteurs de laboratoire standard, par rapport aux flux de reaction

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

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)

Fast flux measurements by means of threshold detectors on the reactor 'Melusine'; Mesures de flux rapides a l'aide de detecteurs a seuil sur le reacteur 'Melusine'

Energy Technology Data Exchange (ETDEWEB)

Leger, P; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

Using existing data on the (n,p) and (n,{alpha}) threshold reactions we have carried out fast flux measurements on the swimming pool type reactor 'Melusine'. Four common elements: P, S, Mg, Al were chosen because from the point of view of fast spectrum analysis they represent a fairly good energy range from 2.4 MeV to 8 MeV. The fission flux value found in the central element at a power of 1 MW is 1.4 x 10{sup 13} n/cm{sup 2}/s {+-} 0.14. (author) [French] A l'aide des donnees actuelles sur les reactions a seuil (n,p) et (n,{alpha}) nous avons realise des mesures de flux rapide dans le reacteur du type piscine 'Melusine'. Quatre corps courants: P, S, Mg, Al, ont ete choisis parce qu'ils constituent au point de vue de l'analyse du spectre rapide un assez bon etalement en energie de 2,4 MeV A 8 MeV. La valeur du flux de fission trouve dans l'element central a une puissance de 1 MW est de 1,4.10{sup 13} n/cm{sup 2}/s {+-} 0,14. (auteur)

The influence of the (n, 2n) and (n, {alpha}) reactions of beryllium on the neutron balance in a BeO or Be moderated reactor and its consequences on the long term reactivity changes; Influence des reactions (n, 2n) et (n, {alpha}) du beryllium sur le bilan neutronique d'un reacteur modere a l'oxyde de beryllium ou au beryllium. Consequences sur l'evolution a long terme de la reactivite

Energy Technology Data Exchange (ETDEWEB)

Sahai, K; Benoist, P; Horowitz, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The reaction probabilities in an infinite and homogeneous medium of BeO or Be have been calculated from neutron cross-section curves, for a neutron produced with an energy distribution similar to a fission spectrum; the calculation shows that, after several elastic collisions, the neutron has yet an appreciable probability to undergo a reaction, in spite of the energy degradation in the spectrum due to each collision. This degradation has been calculated, taking into account of anisotropy of the collisions. The gain of the reactivity in a reactor has been obtained after correcting these probabilities for the attenuation of the flux of fission neutrons due to the inelastic scattering in the uranium. Finally, the calculation shows that in a power reactor, this gain of reactivity is in practice destroyed in a few years by the accumulation of poisonous nuclei such as Li{sup 6} and He{sup 3} following (n, {alpha}) reaction. (author) [French] Les probabilites de reaction en milieu infini et homogene de glucine (BeO) ou de beryllium ont ete calculees a partir des courbes de section efficace pour un neutron naissant suivant un spectre de fission; le calcul montre qu'apres plusieurs diffusions elastiques le neutron a encore une probabilite appreciable de subir une reaction, malgre la degradation du spectre a chaque diffusion; cette degradation a ete calculee en tenant compte de l'anisotropie du choc. Le gain de reactivite dans un reacteur a ensuite ete obtenu en corrigeant les probabilites en milieu homogene de l'effet l'attenuation du flux des neutrons de fission par les chocs inelastiques dans les barres d'uranium. Enfin, le calcul montre que, dans un reacteur de puissance, ce gain de reactivite est pratiquement detruit en peu d'annees par l'accumulation de noyaux poisons Li{sup 6} et He{sup 3} consecutive a la reaction (n, {alpha}). (auteur)

Study of the long-term values and prices of plutonium; a simplified parametrized model; Etude des valeurs et des prix du plutonium a long terme; un modele parametre simplifie

Energy Technology Data Exchange (ETDEWEB)

Gaussens, J; Paillot, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

The authors define the notions of use values and price of plutonium. They give a 'simplified parametrized model' simulating the equilibrium of the offer and the demand in time, concerning the plutonium and the price deriving from the relative scarcity of this metal, taking into account the technical and economic operating parameters of the various reactors confronted. This model is simple enough to allow direct computations and establish clear relations between the various parameters. The use of the linear programmes method allows on the other hand a wide extension of the model. This report includes three main parts: I - General description of the study (without detailed calculations) II - Mathematical development of the simplified parametrized model and application (the basic data and the results of the calculations are given) III - Appendices (giving the detailed computations of part II). (authors) [French] Les auteurs definissent les notions de valeurs d'usage et de prix du plutonium. Ils donnent un 'modele parametre simplifie' simulant l'equilibre de l'office et de la demande dans le temps concernant le plutonium et le prix qui decoule de la rarete relative de ce metal, compte tenu des parametres techniques et economiques de fonctionnement des divers reacteurs en presence. Ce modele est suffisamment simple pour permettre des calculs manuels et etablir des liaisons claires entre les divers parametres. L'utilisation de la technique des programmes lineaires permet par ailleurs une extension considerable du modele. Cette note comprend trois parties: I - Expose general de l'etude (sans expose du detail des calculs) II - Developpement mathematique du modele parametre simplifie et application (on precise les donnees de base et le resultat des calculs) III - Annexes (donnant le detail des calculs de la partie II). (auteurs)

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)

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

Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

Energy Technology Data Exchange (ETDEWEB)

Audinet, M [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

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)

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

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)

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

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)

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

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

Non-Destructive Testing in Reactor Pressure-Vessel Fabrication; Essais non Destructifs dans la Fabrication des Caissons Etanches de Reacteurs; Nedestruktivnoe ispytanie pri izgotovlenii reaktornykh bakov vysokogo davleniya; Ensayo no Destructivo Durante la Fabricacion de Recipientes de Presion para Reactores

Energy Technology Data Exchange (ETDEWEB)

McGonnagle, W. J. [Fluids Dynamics Research, Iit Research Institute, Chicago, IL (United States)

1965-09-15

of the pressure vessel are discussed. (author) [French] Le memoire a pour objet d'exposer les grandes lignes d'un programme de controle de la qualite dans la fabrication d'un caisson etanche de reacteur qui satisfera a toutes les specifications du point de vue nucleaire et de la securite, et de mettre en evidence le role et l'importance des essais non destructifs dans ce programme. Les defauts constates dans les materiaux, les elements et leur assemblage montrent que les methodes actuelles de fabrication ne permettent pas en elles-memes d'assurer le maintien de la qualite des elements critiques. 11 se produit des pailles et des heterogeneites memes lorsque l'on utilise les meilleurs procedes de fabrication et que l'on applique des methodes et techniques dument controlees. C'est pourquoi, afin d'obtenir la qualite requise pour un caisson de reacteur, il faut executer un programme approprie et coherent d'essais non destructifs. Les principales methodes d'essais non destructifs appliquees par les fabricants de caissons de reacteurs sont les suivantes: inspection visuelle, radiographie par les rayons X ou gamma, ultrasons, particules magnetiques et penetration de liquides. Le programme d'essais non destructifs comporte le controle des materiau', du forgeage, du moulage, du gainage et des soudures. L'auteur etudie les problemes particuliers que posent les essais non destructifs des caissons etanches. Il decrit et discute les techniques speciales propres aux essais non destructifs des caissons et de leurs elements. Le memoire donne un apercu des reglements et specifications applicables, notamment du reglement de fabrication des bouilleurs et caissons etanches publie par la Societe americaine des ingenieurs mecaniciens. L'auteur etudie la mesure dans laquelle les essais non destructifs peuvent contribuer a repondre aux specifications imposees par les institutions de normalisation, ainsi que la mesure dans laquelle les normes admises pour ces essais sont appropriees et

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)

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

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.)

Experimental study of heat transfer and pressures drops for cans with spiral herring-bone fins; Etude experimentale du transfert de chaleur et des pertes de charges des gaines a ailettes helicoidales en chevron

Energy Technology Data Exchange (ETDEWEB)

Pelce, J; Francois, S; Houseaux, O; Pierre, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Cans fitted with herring-bone fins are used for cooling uranium in certain nuclear reactor. By herring-bone is meant a staggered arrangement of the fins which have a plane of symmetry parallel to the general direction of liquid flow. The main geometrical parameter are then: the number of fins, the number of herring-bones, the angle of inclination of the fins with respect to the can axis, the dimensions of the fins, the can diameter and the channel diameter. The research is essentially experimental. The test are of three types: full size tests, in conditions approaching those in the reactor (constant flux, CO{sub 2} under pressure); full size tests but with a constant wall temperature, much easier to set up, and intended to distinguish rapidly between the merits of the various types of can; large-scale tests with air at atmospheric pressure for studying the phenomena in more detail. For each can tried out there is a corresponding pressure drop coefficient, a mean thermal exchange coefficient Mo-bar and a minimum exchange coefficient Mo{sub min} and Mo-bar are related by the expression Mo{sub min} = Mo-bar * f{sub c} * f, where f{sub c} and f are respectively circumferential and longitudinal singularity factor determined from a statistical study of all the temperatures measured for each can. The results are presented in about thirty tables and figures the most noteworthy results being summarized in the conclusion. (authors) [French] Les gaines a ailettes en chevron sont utilisees pour le refroidissement de l'uranium dans certains reacteurs nucleaires. Par chevron, on entend une disposition alternee des ailettes ayant un plan de symetrie parallele a la direction generale de l'ecoulement fluide. Les principaux parametres geometriques sont alors: le nombre des ailettes, le nombre de chevrons, l'angle d'inclinaison des ailettes par rapport a l'axe de la gaine, les dimensions des ailettes, le diametre de la gaine et le diametre du canal. L'etude est essentiellement

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)

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)

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)

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)

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

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

Magnox Fuel Cycles; Cycles des combustibles gaines de magnox; Toplivnye tsikly magnoks; Ciclos de combustible magnox

Energy Technology Data Exchange (ETDEWEB)

Johnson, A. [United Kingdom Atomic Energy Authority, Risley, Warrington, Lancs (United Kingdom)

1963-10-15

The interaction between reactivity flux and temperature distributions and irradiation patterns caused by different refuelling policies is considered and present calculation methods outlined. Various refuelling schemes for both batch and continuous discharge systems are compared. The problem of the efficient irradiation of the first charge is considered together with delayed onset refuelling and shuffling schemes. The economic advantages and problems of using non-natural uranium in flattened reactors are discussed. The practical consideration of on-load refuelling schemes on new reactors are considered and reference is made to the experience gained on Bradwell and Berkeley. The effect of the variation of fuel cost and endurance on fuel-cycle economics is outlined. (author) [French] L'auteur etudie en premier lieu l'interaction entre les distributions de la temperature du flux et de la reactivite, d'une part, et le regime de l'irradiation, d'autre part, dans le cas de differents programmes de rechargement du combustible et il decrit brievement les methodes de calcul actuelles. Il compare ensuite differents programmes de rechargement du combustible pour le dechargement par lots et le dechargement continu. Il etudie le probleme de l'irradiation effica ce de la premiere charge dans le cadre de programmes de remplacement et de deplacement des cartouches a action retardee. Il analyse les avantages economiques de l'utilisatio n d'uranium non naturel dans les reacteurs a flux aplati et les problemes qu'elle pose. Il examine les aspects pratiques des programmes de rechargement en marche pour les nouveaux reacteurs, en se referant a l'experience acquise au moyen des reacteurs de Bradwell et de Berkeley. Enfin, il decrit brievement les effets des variations du cout et de la resistance du combustible sur l'economie des cycles de combustible. (author) [Spanish] La memoria estudia la interaccion entre el flujo de reactividad y la distribucion de temperaturas, asi como los

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)

Detection of radioactive gases in the CO{sub 2} cooling the reactors G 2 - G 3; Detection des gaz radioactifs dans le CO{sub 2} de refroidissement des piles G2 - G3

Energy Technology Data Exchange (ETDEWEB)

Pouthier, J; Rossi, J [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

1968-07-01

The carbon dioxide cooling the reactors G2 - G3 contains activation gases and fission gases. It is of interest to know their concentration, for example to be able to deduce rapidly the norms which would have to be applied in the case of an incident in the circuit. Gas-phase chromatography is applied daily for carrying out analyses. The chromatogram has separate peaks due to tritium, argon 41, krypton 85 and the 133 and 135 isotopes of xenon. By integrating each peak it is possible to calculate the specific activity of each product. The construction of an apparatus for carrying out continuous measurements is under consideration. (authors) [French] Le gaz carbonique, refroidissant les reacteurs G2 - G3, contient des gaz d'activation et des gaz de fission. Il est interessant de connaitre leur teneur par exemple pour etre en mesure de deduire rapidement les normes qu'il y aurait lieu d'appliquer en cas d'incidents sur le circuit. La methode de chromatographie en phase gazeuse est employee quotidiennement pour faire des analyses. Le chromatogramme se presente sous forme de pics distincts dus au tritium, a l'argon 41, au krypton 85 et aux isotopes 133 et 135 du xenon. L'integration de chaque pic permet de calculer l'activite specifique de chaque compose. Il est envisage de construire un appareil pour des mesures en continu. (auteurs)

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

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

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

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

Simulation of power excursions - Osiris reactor; Simulation des excursions de puissance - pile Osiris

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author) [French] A la suite des essais effectues aux U.S.A. sur BORAX 1 et SPERT 1 et de l'accident survenu a SL 1, le Commissariat a l'Energie Atomique a lance un programme d'etudes sur la surete de ses reacteurs piscines vis-a-vis des excursions de puissance. Les premieres etudes ont abouti A la mise au point de charges programmees capables de simuler une excursion de puissance. On trouvera dans le present rapport l'application de ces methodes a l'etude de la surete d'OSIRIS. (auteur)

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)

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

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.

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)

Calculation of actual cross sections and thermalization of neutrons; Calcul des sections efficaces effectives et thermalisation des neutrons

Energy Technology Data Exchange (ETDEWEB)

Naudet, R.

1963-05-15

This report gathers and presents in a simple way results of studies performed at the CEA on issues of spectra in thermal reactors. It is in fact a synthesis of results eventually published in different documents. It first presents the notion of actual cross section as it was introduced by Westcott to characterize the dependence of neutron behaviour on speed distribution. It addresses the case of a homogeneous medium with a conventional model, with the heavy gas model, and with the hydrogen gas model. It generalizes the approach by the differential model. The next part addresses the case of a heterogeneous medium, and the case of presence of moderator nuclei within the fuel [French] Le present rapport a pour objet de rassembler et de presenter de maniere simple les resultats des etudes effectuees au CE.A. sur les problemes de spectres dans les reacteurs thermiques. Ces resultats se trouvaient disperses dans plusieurs documents, ou n'etaient pas encore rediges, et bien que les etudes se poursuivent, il a paru utile d'en faire une synthese provisoire. On a cherche d'autre part a en donner une presentation elementaire, accessible aux lecteurs peu familiarises avec les problemes de thermalisation; dans cet esprit l'expose a une forme didactique, et comporte des rappels de notions bien connues comme par exemple le formalisme de Westcott. (auteur)

Testing of a reactimeter for a light water reactor in the range + 500 to - 5000 pcm; Essai d'un reactimetre pour reacteur a eau legere dans la gamme + 500, - 5000 pcm

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

calcul de la reactivite ne repose pas sur un asservissement. Un de ses inconvenients est de ne pas pouvoir fonctionner en dehors d'une plage de variation de la puissance excedant 2,5 decades. Mais la mesure d'un echelon negatif de reactivite entre 0 et 3000 pcm est immediate. Il mesure la reactivite en ne la deduisant pas de la periode; il indique donc la reactivite d'une maniere precise en divergence aussi bien qu'en convergence, regime ou il n'existe pas, a proprement parler, de periode. Il permet donc un etalonnage tres rapide des barres de controle d'un reacteur (methode de rod-drop), la mesure de la reactivite d'une manipulation inseree dans le coeur, la mesure de certains effets de temperature. En inserant 'au moteur' une barre de controle dans le coeur, on peut tracer directement sa courbe d'efficacite. (auteur)

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

Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

Energy Technology Data Exchange (ETDEWEB)

Gaussens, J; Tanguy, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

reduction in investment costs can be obtained without relying on fuel enrichment, and that this development is accompanied moreover by improvements in the operational safety of the reactor. The economic aspects of the main technical problems entailed by these developments are discussed: loading and unloading machines, blowers etc... (authors) [French] Dans une premiere partie, on situe l'interet economique de l'utilisation de l'uranium naturel comme combustible. Cet interet reside a la fois dans le nombre limite et la simplicite relative des operations de mise en forme des elements combustibles, dans le faible cout du produit fini par kwh et dans les immobilisations modestes en capital qu'implique ce cycle par rapport ou cycle de l'uranium enrichi. Tous ces elements permettent de reduire le caractere aleatoire des evaluations des couts, particulierement marque dans le cas de l'uranium enrichi, en raison de la complexite de son cycle et des incertitudes concernant le prix du plutonium. Enfin, la diversite des sources d'approvisionnement en concentre d'uranium naturel opposee au quasi monopole actuel de la separation isotopique, et le faible cout du stockage de ce concentre, offrent des garanties en matiere de securite d'approvisionnement et d'independance economique et politique appreciables par rapport a l'uranium enrichi. En ce qui concerne l'ensemble des capitaux immobilises, on montre que si le cout des centrales au graphite-gaz est plus eleve que celui des centrales eau legere pour certaines gammes de puissance, ce resultat est fortement nuance des que l'on fait intervenir dans un souci d'independance nationale le cout de l'equipement de production des combustibles de l'une et l'autre filiere. Enfin, le cout marginal de la puissance du reacteur au graphite est faible, ses limitations technologiques ont considerablement recule (grace en particulier a l'utilisation du beton precontraint). On sait que la tendance actuelle est a l'accroissement de la puissance unitaire des

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.

Professional Nuclear Materials Management; Gestion Industrielle des Matieres Nucleaires; Obrashchenie s yadernymi materialami na professional'nom urovne; Administracion Eficiente de Materiales Nucleares

Energy Technology Data Exchange (ETDEWEB)

Forcella, A. A.; O' Leary, W. J. [Allis-Chalmers Manufacturing Company, Bethesda, MD (United States)

1966-02-15

. (author) [French] Le memoire expose en quoi consiste la gestion des matieres nucleaires dans une centrale nucleaire type, aux Etats-Unis. Comme le reacteur en question est partiellement finance a l'aide de capitaux prives, l'une des principales obligations de l'exploitant est d'assurer la protection et la rentabilite des investissements. Etant donne que les matieres nucleaires sont d'une valeur intrinseque elevee, il faut constamment proceder a des controles appropries allant au-dela des mesures de securite et de la comptabilite interessant les matieresnucleaires proprement dites afin de reduire les pertes au minimum. Il faut faire preuve de clairvoyance et planifier judicieusement pour prevenir toute perte supplementaire de capital provenant de frais inutiles ou d'un manque a gagner dans divers secteurs de l'exploitation. C'est ainsi que le gestionnaire de matieres nucleaires doit prendre des dispositions pour garantir la bonne marche des operations et assurer le respect des plans d'execution par une liaison et un controle constants, dans les domaines suivants? a) acquisition du combustible et des elements combustibles, b) utilisation des elements combustibles dans le reacteur et c) recuperation, dans le combustible irradie, des produits et matieres presentant de l'interet. Pendant la periode qui precede l'utilisation du reacteur, il faut faire une place importante dans la planification et les travaux preparatoires aux considerations d'economie dans la conception des elements combustibles, en ce qui concerne leur fabrication, leur manutention, leur transport et leur remplacement. Les differentes etapes de la fabrication doivent etre planifiees de facon a reduire au minimum le manque a gagner du a des periodes improductives d'entreposage de matieres tres couteuses. Pendant la marche du reacteur, il faut assurer une combustion maximale de la matiere fissile par des redistributions appropriees du combustible dans le coeur du reacteur. Parallelement, les temps morts dus a

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)

Alize 3 - first critical experiment for the franco-german high flux reactor - calculations; Alize 3 - premiere experience critique pour le reacteur a haut flux franco-allemand. Calculs

Energy Technology Data Exchange (ETDEWEB)

Scharmer, K [Commissariat a l' Energie Atomique, Dir. des Piles Atomiques, Saclay (France). Centre d' Etudes Nucleaires

1969-07-01

The results of experiments in the light water cooled D{sub 2}O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k{sub eff} was smaller than 0.5 per cent {delta}k/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D{sub 2}O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [French] Les resultats des experiences faites dans la maquette critique ALIZE III, refrigeree a l'eau legere et reflechie par l'eau lourde, ont ete compares aux calculs. On a utilise un modele de la theorie de diffusion a trois groupes rapides et epithermiques et deux groupes thermiques qui se recouvrent. Ce modele a permis de calculer la distribution de puissance dans le coeur en bon accord avec les mesures, meme dans le cas d'une forte variation du spectre des neutrons dans le coeur. L'erreur entre k{sub eff} calcule et mesure etait inferieure a 0,5 pour cent {delta}k/k. Le coefficient de vide et des materiaux de structure, la reactivite des barres 'noires', les variations du spectre (rapport Cd, rapport Pu/U) et la fraction des photo-neutrons retardes sont egalement calcules. Les mesures de reactivite et de perturbation de flux dans le reflecteur, dues aux canaux, ont ete interpretees du point de vue d'un arrangement optimum des canaux pour le Reacteur a Haut Flux Franco-Allemand. (auteur)

The Application of Non-Metallic Core Materials in a High-Temperature Reactor Experiment; Utilisation de materes non metalliques dans le coeur d'un reacteur experimental a haute temperature; Ispol'zovanie nemetallicheskikh materialov dlya aktivnoj zony vysokotemperaturnogo opytnogo reaktora; Empleo de materiales no metalicos en el nucleo de un reactor experimental de alta temperatura

Energy Technology Data Exchange (ETDEWEB)

Huddle, R. A.U.; Shepherd, L. R. [Organization for Economic Co-Operation and Development, Dragon Project, Atomic Energy Establishment, Winfrith, Dorset (United Kingdom)

1963-11-15

The OECD High-Temperature Reactor Project (DRAGON) was set up to develop the technology of high-temperature gas-cooled reactors and, as part of this development, to construct and operate a 20-MW(t) reactor experiment. The reactor, which is now nearing completion, is a helium-cooled system with a coreoutlet temperature of 750{sup o}C; it employs U{sup 235} fuel with thorium as a fertile material. A particular feature of this system is the absence of any metals in the core. Because of the high temperatures involved, namely, up to 1050{sup o}C at fuel element surfaces and above, 1500{sup o}C in-the hottest regions of the fuel, refractory nonmetallic materials are employed. All the core material is incorporated within the fuel element which leads to a high ratio of heat transfer surface area to core volume and hence permits a high average power density leading to a relatively compact system. Each fuel element consists of a cluster of graphite tubes, containing the fissile and fertile materials as carbides incorporated in graphite pellets. A purge flow of the helium coolant passing through the centre of each fuel rod is extracted from the base whence it passes into a helium processing plant to remove fission products and other impurities before being returned to the reactor. This procedure reduces the escape of fission products from the very hot ceramic fuel into the primary coolant stream. Problems associated with the development and production of ceramic fuel bodies and graphite for this reactor, and the behaviour of these materials under operating conditions are outlined. Some experience from irradiation and in-pile loop investigations are reported. The main emphasis in this programme is on the development of the high-temperature gas-cooled reactor for application as an economic power producing system. (author) [French] Les objectifs du Projet DRAGON de l'OCDE (reacteur a haute temperature) sont les suivants: ameliorer la technologie des reacteurs a haute temperature

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

Fission gas pressure build-up and fast-breeder economy; Accumulation de la pression des gaz de fission et economie des reacteurs surgenerateurs a neutrons rapides; Nakoplenie davleniya gazov produktov deleniya i ehkonomika reaktorov-razmnozhitelej na bystrykh nejtronakh; Aumento de la presion de los gases de fision y economia de los reactores reproductores rapidos

Energy Technology Data Exchange (ETDEWEB)

Engelmann, P [Kernforschungszentrum, Karlsruhe (Germany)

1962-03-15

Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [French] Le cout du cycle de combustible et la periode de doublement des reacteurs surgenerateurs a neutrons rapides dependent etroitement du taux de combustion. En utilisant pour combustible un oxyde ou un carbure, on peut atteindre un taux de combustion de 100 000 MW j/t. Avec des combustibles gaines, l'accumulation de la pression des gaz de fission est un facteur limitatif. Pour le fort taux de combustion envisage, une fraction non negligeable des gaz de fission penetre dans les interstices et contribue ainsi a la pression sur la gaine. A partir des rendements en produits de fission et des chaines de desintegration connus, l'auteur a calcule la production de gaz et l'accumulation de pression. Pour calculer la pression

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

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.

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)

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)

Partial combustion of a fuel cartridge in reactor G1; Combustion partielle d'une cartouche de combustible dans le reacteur G 1

Energy Technology Data Exchange (ETDEWEB)

De, Rouville; Leduc,; Segot, [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

-devices, some null regulating tension systems, annealing the background due to continuous pollution. This event has been fruitful. A grid trap has been set right ahead the reactor. Stricter instructions have been given for rising power operations and automatic burst slug sy (already improved as said above) has been duplicated by a human control. At last, the fault has pointed out that the reactors with gap had the disadvantage of facilitating the contamination of channels from one to another. On the other hand, graphite stores the radioactive dusts and hinders an easy decontamination. (author) [French] Le 26 octobre 1956, le reacteur G1 etait remis en marche apres un arret de quelques jours. L'installation de detection de rupture de gaines donna un premier signal de prealerte a 19h07 cote chargement, un second a 19h13 cote dechargement, puis d'autres. Le chef de quart ordonna a 19h15 une baisse rapide de la puissance mais voulant reperer le canal fautif avec precision la fit remonter ensuite a 2 puis a 5 MW. Bientot, par crainte de contamination exterieure, on dut arreter l'exploration et c'est par detection {gamma} a l'exterieur des tuyaux de detection de rupture de gaine qu'on identifia la cartouche endommagee dans le canal 19-13. Les enregistrements des stations de sante montrerent que les pointes observees etaient restees notablement inferieures aux limites maxima admissibles. L'examen methodique et le degagement du canal accidente occuperent trois semaines. On put apercevoir cote chargement les billettes d'uranium nues sur un lit de poudre de magnesie; cote dechargement, la gaine etait intacte mais l'extremite de la cartouche 'pendait' a l'interieur de la fente d'arrivee d'air. Repoussee cote chargement d'environ 30 cm, la cartouche se bloqua. Apres des essais divers, toujours sous injection d'argon, et avec des protections severes du personnel, on mit en oeuvre un tube fraise, analogue a ceux utilises pour les forages. On nettoya le canal par aspiration, sans toutefois

Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

in order to show the advantages resulting from such developments in gas-graphite natural metallic uranium reactor systems; these are: a doubling of the specific and volume powers, and a three-fold reduction in the number of channels. The research now under way will make it possible to calculate the reduction in capital costs which will result from these important technical advances. (authors) [French] Le programme francais de centrales a graphite et uranium naturel s'est developpe, d'EDF 1 a EDF 4 - dans la voie d'un accroissement de la puissance unitaire des installations, de la puissance specifique et de la puissance volumique, et d'une amelioration des conditions de securite de fonctionnement. La puissance elevee d'EDF 4 (500 MWe) et l'integration du circuit primaire dans le caisson, lui-meme en beton precontraint, permettent ainsi de tirer le meilleur parti des elements combustibles tubulaires utilises des EDF 1, et d'arriver ainsi a une solution tres satisfaisante. L'emploi d'un element combustible refroidi interieurement (element annulaire) permet de faire un nouveau pas en avant: il devient alors possible d'augmenter la pression du gaz de refroidissement sans craindre le fluage du tube d'uranium. L'emploi d'un caisson en beton precontraint permet une telle augmentation de pression, et l'integration du circuit primaire elimine les risques d'une depressurisation rapide qui aurait presente dans ce cas un risque majeur. On aborde dans ce rapport les principaux problemes poses par ce nouveau type de centrale et on indique les grandes lignes des recherches et etudes effectuees en France: - Les etudes de neutronique et thermique ont permis d'envisager l'emploi d'elements combustibles de grandes dimensions (diametre interne = 77 mm, diametre externe = 95 mm), tout en conservant l'uranium naturel. - Les problemes de fabrication de ces elements, et de leur comportement en pile, font l'objet d'un programme important, tant hors pile que dans les piles de puissance (EDF 2

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.

Burn up physics; Physique des combustibles irradies

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

requires samples of the order of several kilograms only. The relationships between these measurements and the investigations of lattices are discussed, and an outline is given of the way of carrying out the systematic study of fuels of various compositions. The method has been successfully applied to the systematic study of irradiated fuels (analysed independently by the methods mentioned above) thus giving the possibility of measuring in situ the absorption of fission products. (author) [French] Cette communication expose un ensemble d'etudes theoriques et d'experiences, effectuees au CEA et destinees a faire progresser la connaissance de l'evolution de la reactivite (au cours de l'irradiation du combustible) dans les reacteurs a uranium naturel ou faiblement enrichi.,. On rappelle les difficultes de l'experimentation directe sur des masses importantes de combustible irradie - en particulier dans les reacteurs de puissance en fonctionnement - et on souligne la necessite d'experiences a caractere fondamental distinguant: d'une part l'evolution de la composition des combustibles (chaines de noyaux lourds, produits de fission), d'autre part l'effet des modifications de composition sur le bilan de neutrons. Avant de presenter trois categories d'experiences que l'on est conduit a entreprendre, on rappelle l'importance des problemes lies aux spectres de neutrons et on decrit rapidement les methodes pratiques de calcul utilisees. L'irradiation systematique de quelques types de combustibles, suivie de leur analyse chimique et isotopique est en cours depuis plusieurs annees. On donne un apercu de l'ensemble du programme experimental et on decrit les moyens et les methodes mis en oeuvre: chaine {alpha}, {beta}, {gamma} pour la preparation des echantillons, dosage du Plutonium par coulommetrie et double dilution isotopique, separation du Bore utilise dans certains cas pour la mesure des densites de neutrons integrees. On discute sur quelques exemples l'interpretation des mesures

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

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

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

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

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

Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

Energy Technology Data Exchange (ETDEWEB)

Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

1962-04-15

In search for semiconductors, which can be used in high-flux reactors in order to measure flux distributions, we irradiated SiC p-n junctions in the Belgium BR-1 reactor. Two types of SiC-diodes of different origin have been irradiated. These junctions are grown in the Lely-furnace. The change in forward and reverse characteristics have been measured during and after irradiation up to temperatures of 150{sup o}C, while measurements up to a temperature of 500{sup o}C are in progress. It has been found that one type resists BR-1 neutrons up to an integrated flux of 10{sup 15} n/cm{sup 2}, while the other resists irradiation up to a flux of 10{sup 17} n/cm{sup 2}. The changes in characteristics are given as well as the result of some annealing experiments. (author) [French] En recherchant des semi-conducteurs pouvant servir a mesurer les distributions de flux dans les reacteurs a haut flux de neutrons, les auteurs ont irradie des jonctions p-n au SiC dans le reacteur belge BR-1. Deux types de diodes a SiC d'origines differentes ont ete ainsi irradies. Les jonctions en question sont preparees par etirage dans le four Lely. Les auteurs ont mesure les modifications subies par les caracteristiques I-V apres et pendant l'irradiation a des temperatures allant jusqu'a 150{sup o}C; ils poursuivent leurs mesures dans la gamme des temperatures allant de 150{sup o}C a 500{sup o}C. Us ont constate que l'un des types de diode a SiC resiste aux neutrons du reacteur BR-1 jusqu'a 10{sup 15} n/cm{sup 2}, tandis que l'autre type resiste a l'irradiation jusqu'a 10{sup 17} n/cm{sup 2}. Les auteurs indiquent les modifications subies par les caracteristiques, ainsi que le resultat de certaines experiences de recuit. (author) [Spanish] Los autores estan tratando de encontrar semiconductores con los que sea posible medir distribuciones de flujo en reactores de flujo elevado, y con este fin irradiaron uniones p-n del SiC en el reactor BR-1 de Belgica. Irradiaron dos tipos de diodos de SiC de

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)

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.

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)

A study of some radioprotection apparatuses used in the case of pool reactors; Etude de quelques dispositifs de radioprotection en service aupres des piles piscines

Energy Technology Data Exchange (ETDEWEB)

Robien, E de; Choudens, H de; Delpuech, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1965-07-01

necessite de mesurer avec une bonne precision des doses {gamma} a l'aide d'un appareil autonome a conduit a la realisation d'un dosiinetre {gamma} a scintillateur plastique; d) l'obligation de condamner l'ouverture des portes de certains locaux ou il risque d'y avoir une intensite de rayonnement forte a conduit a realiser un asservissement de l'ouverture de ces portes a l'intensite de rayonnement a l'interieur des locaux; e) les degagements d'iode radioactif ont ete mesures a l'aide de cartouches de charbon actif entourant un scintillateur relie a un monocanal; f) enfin, la commande de chute des barres des piles, en cas d'accident radioactif, a ete assuree par une chaine dont le detecteur est une chambre immergee dans la piscine, ce qui offre, dans le cas particulier des reacteurs piscines a couche chaude, le double avantage de pouvoir 'piloter' la couche chaude et de s'affranchir des rayonnements transitoires apparaissant dans le hall du reacteur au moment du defournement des dispositifs experimentaux. (auteurs)

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

Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

in order to show the advantages resulting from such developments in gas-graphite natural metallic uranium reactor systems; these are: a doubling of the specific and volume powers, and a three-fold reduction in the number of channels. The research now under way will make it possible to calculate the reduction in capital costs which will result from these important technical advances. (authors) [French] Le programme francais de centrales a graphite et uranium naturel s'est developpe, d'EDF 1 a EDF 4 - dans la voie d'un accroissement de la puissance unitaire des installations, de la puissance specifique et de la puissance volumique, et d'une amelioration des conditions de securite de fonctionnement. La puissance elevee d'EDF 4 (500 MWe) et l'integration du circuit primaire dans le caisson, lui-meme en beton precontraint, permettent ainsi de tirer le meilleur parti des elements combustibles tubulaires utilises des EDF 1, et d'arriver ainsi a une solution tres satisfaisante. L'emploi d'un element combustible refroidi interieurement (element annulaire) permet de faire un nouveau pas en avant: il devient alors possible d'augmenter la pression du gaz de refroidissement sans craindre le fluage du tube d'uranium. L'emploi d'un caisson en beton precontraint permet une telle augmentation de pression, et l'integration du circuit primaire elimine les risques d'une depressurisation rapide qui aurait presente dans ce cas un risque majeur. On aborde dans ce rapport les principaux problemes poses par ce nouveau type de centrale et on indique les grandes lignes des recherches et etudes effectuees en France: - Les etudes de neutronique et thermique ont permis d'envisager l'emploi d'elements combustibles de grandes dimensions (diametre interne = 77 mm, diametre externe = 95 mm), tout en conservant l'uranium naturel. - Les problemes de fabrication de ces elements, et de leur comportement en pile

Activity report 1999; Rapport d'activites 1999

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

The aim of this report is to outline the main developments of the ''Departement des Reacteurs Experimentaux'', (DRE) during the year 1999. DRE is one of the Department of the ''Direction des Reacteurs Nucleaires'', itself depending of the CEA Institution. After a presentation of the year highlights, this report gathers the main research and development programs. The second part concerns the production of radioisotopes, the silicon doping, the neutron radiography, the Orphee experiments and the activation analysis. The installations management, the closed reactors improvement program and the effluents and wastes processing of Grenoble, are presented in the other parts. Data on staff, budget and safety are also provided. (A.L.B.)

Activity report 1998

International Nuclear Information System (INIS)

1998-01-01

The aim of this report is to outline the main developments of the ''Departement des Reacteurs Experimentaux'', (DRE) during the year 1998. DRE is one of the Department of the ''Direction des Reacteurs Nucleaires'', itself depending of the CEA Institution. After a presentation of the year highlights, this report gathers the main research and development programs. The second part concerns the production of radioisotopes, the silicon doping, the neutron radiography, the Orphee experiments and the activation analysis. The installations management, the closed reactors improvement program and the effluents and wastes processing of Grenoble, are presented in the other parts. Data on staff, budget and safety are also provided. (A.L.B.)

Activity report 1999

International Nuclear Information System (INIS)

1999-01-01

The aim of this report is to outline the main developments of the ''Departement des Reacteurs Experimentaux'', (DRE) during the year 1999. DRE is one of the Department of the ''Direction des Reacteurs Nucleaires'', itself depending of the CEA Institution. After a presentation of the year highlights, this report gathers the main research and development programs. The second part concerns the production of radioisotopes, the silicon doping, the neutron radiography, the Orphee experiments and the activation analysis. The installations management, the closed reactors improvement program and the effluents and wastes processing of Grenoble, are presented in the other parts. Data on staff, budget and safety are also provided. (A.L.B.)

Sodium purification in Rapsodie; La purification du sodium a Rapsodie

Energy Technology Data Exchange (ETDEWEB)

Giraud, B [Commissariat a l' Energie Atomique, Dir. des Piles Atomiques, Cadarache (France). Centre d' Etudes Nucleaires

1968-07-01

This report is one of a series of publications presenting the main results of tests carried out during the start-up of the first french fast neutron reactor: Rapsodie. The article presents the sodium purification techniques used in the reactor cooling circuits both from the constructional point of view and with respect to results obtained during the first years working. (author) [French] Ce rapport fait partie d'une serie de publications presentant l'essentiel des resultats des essais effectues a l'occasion du demarrage du premier reacteur francais a neutrons rapides: RAPSODIE. Cet article expose les techniques de la purification du sodium utilise dans les circuits de refroidissement du reacteur tant au point de vue de leur realisation technologique, que des resultats obtenus pendant la premiere annee de fonctionnement. (auteur)

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

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

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.

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

Indium-Gallium Radiation Contour of the IRT Nuclear Reactor; Circuit d'activation d'indium-gallium dans le reacteur nucleaire IRT; Indij-gallievyj radiatsionnyj kontur yadernogo reaktora IRT; Circuito de radiaciones de indio-galio del reactor IRT

Energy Technology Data Exchange (ETDEWEB)

Breger, A K; Ryabukin, Y S; Tulkes, S G; Volkov, E N

1960-07-15

Following on theoretical work already published, an indium-gallium radiation contour of the IRT nuclear reactor has been prepared, and represents a powerful new source of gamma-radiation. The first contour of this type ''RK-1'' was prepared on the IRT reactor at the Physics Institute of the Academy of Sciences of the Georgian SSR. The paper gives the activation calculations for indium-gallium alloy; the structural components of RK-1 and their arrangement in the reactor tank and the hot cell; the devise for feeding liquid and gaseous substances into the irradiation zone; and the conveyor for solid substances to be irradiated. When the IRT reactor is at a power of 2000 kW, the radiation strength of the contour is equivalent to that of a gamma-emitter having an activity of 20,000 g. Ra equivalent. The prospects for the use of the indium-gallium radiation contour for research and semi-industrial purposes are discussed. (author) [French] A la suite de la publication d'un ouvrage theorique, on a etabli autour du reacteur nucleaire IRT un circuit d'activation d'indium-gallium qui represente une nouvelle source de rayonnements gamma de grande intensite. Le premier circuit de ce type ''RK-1'' a ete etabli sur le reacteur IRT a l'Institut de physique de l'Academie des sciences de la RSS de Georgie. Les auteurs donnent les calculs de l'activation pour l'alliage indium-gallium; ils indiquent les elements structurels du RK-1 et leur disposition dans le reservoir et dans la cellule de haute activite du reacteur; ils decrivent le dispositif permettant d'introduire des substances liquides et gazeuses dans la zone d'irradiation et le systeme qui transporte les substances solides a irradier. Lorsque le reacteur IRT fonctionne a 2 000 kW, la puissance de rayonnement du circuit equivaut a celle d'un emetteur gamma ayant une activite equivalente a 20 000 grammes de radium. Les auteurs examinent les perspectives d'emploi de ce processus pour la recherche et a des fins semi

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

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...

{gamma} activity and heating of rods in EL2 and EL3; Activitiy {gamma} et echauffement des barres de EL2 et EL3

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

A method is described for calculating the {gamma} activity of uranium rods, given the mean flux in which they are irradiated, the time they remain in the pile and the duration of deactivation. This calculation leads to numerical formulae which may be applied to the rods of the two reactors. It allows the saturation activities to be foreseen both for EL2 and for EL3, taking into recount the minimum times necessary for extraction. Measurements have been carried out, and the results are in good agreement with those foreseen by calculation. In the last section this method is used to calculate the heating of the irradiated rods. (author) [French] Une methode est indiquee ici, qui permet de calculer l'activite {gamma} des barres d'uranium connaissant le flux moyen dans lequel elles ont ete irradiees, leur temps de sejour en pile et la duree de la desactivation. Ce calcul conduit a des formules numeriques que l'on peut appliquer aux barres des deux reacteurs. Il permet de prevoir les activites atteintes a saturation, tant a EL2 qu'a EL3, compte tenu des temps minima necessaires a l'extraction. Des mesures ont ete faites: les resultats sont en bon accord avec les previsions du calcul. Enfin, en derniere partie, cette methode est utilisee pour calculer l'echauffement des barres irradiees. (auteur)

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)

The behaviour of some polyatomic gases in nuclear reactors; Le comportement de quelques gaz polyatomiques dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The chemical effect of ionizing radiations on a certain number of gaseous systems is described. Under the influence of radiations from a reactor, NH{sub 3}, is decomposed to nitrogen and hydrogen in stoichiometric proportions. Formation of N{sub 2}H{sub 3}, particularly could not be detected. Under a slow neutron flux the reaction {sup 14}N (n, p) {sup 14}C constitutes the main source of decomposition energy. Direct recombination of H, and N, has been brought about under the influence of radiation. The radiolysis of NH{sub 3}, occurs by a complex mechanism; and the kinetics follow a law of the order of about 2.5 which increases with the decomposition rate. The decomposition of hydrogen sulphide is appreciably faster than that of NH{sub 3}. Hydrogen is the only gaseous product of the reaction. The sulphur, which is deposited on the walls of the ampoules, is clearly visible to the naked eye. Up to the present decompositions up to 84 per cent have been obtained. The influence of the reaction {sup 32}S (n, p) {sup 32}P is considered. Radiochemical decomposition of nitrous oxide N{sub 2}O takes place with high yields. The reaction is complicated from the beginning by the formation of higher oxides of nitrogen which we identify and measure. Radiochemical decomposition of methane gives quantities of higher hydrocarbons. Certain of these gaseous systems could find applications in the measurement of high doses of radiation. This problem is discussed in the conclusion. (author)Fren. [French] L'effet chimique des rayonnements ionisants sur un certain nombre de systemes gazeux est decrit. Sous l'influence des rayonnements d'un reacteur, NH{sub 3} se decompose en azote et hydrogene en proportions stoechiometriques. En particulier aucune formation de N{sub 2}H{sub 4}, n'a pu etre detectee. Sous flux de neutrons lents, la reaction {sup 14}N (n, p){sup 14}C constitue la principale source d'energie de decomposition. La recombinaison directe de H{sub 2} et N{sub 2} a ete realisous l

Practical guide to dosimetry as applied in the research reactors of the Saclay and Grenoble nuclear research centers; Guide pratique de la dosimetrie mise en oeuvre dans les reacteurs de recherche du C.E.N./G et du C.E.N./S

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-07-01

Since the problems concerning neutron and gamma flux measurements which arise during irradiation experiments in the reactors in the Grenoble and Saclay Centres are of the same type, and since the solutions found are very often adopted in common, we have attempted to describe the methods we use at the present time. A brief description is given of the production of the detectors, the electronic apparatus; the formulae usually used for the interpretation of the measurements are given. A series of technical data cards give the most commonly used detector characteristics. These cards give the physical characteristics of the detectors, their nuclear constants, if any, the most suitable counting methods and the field of application. (authors) [French] Les problemes de mesures de flux de neutrons et de flux gamma qui se posent pour les experiences irradiees dans les reacteurs des Centres de Grenoble et de Saclay etant du meme type et les solutions trouvees, tres souvent adoptees en commun, nous avons cherche a decrire les methodes que nous pratiquons actuellement. On decrit tres brievement la fabrication des detecteurs, l'appareillage electronique; on rappelle les formules usuelles qui servent dans l'interpretation des mesures. Une serie de fiches techniques rassemble les caracteristiques des detecteurs les plus couramment utilises. Ces fiches indiquent les caracteristiques physiques des detecteurs, leurs constantes nucleaires s'il y a lieu, les methodes de comptage les mieux adaptees et le domaine d'utilisation. (auteurs)

Control Methods Used in the Department of Metallurgy for Structure and Fuel Elements; Methodes de Controle Utilisees au Departement de Metallurgie pour les Elements de Structure et les Elements Combustibles; Metody kontrolya struktury toplivnykh ehlementov v departamente metallurgii; Metodos de Control Utilizados en el Departamento de Metalurgia para los Elementos Estructurales y Combustibles

Energy Technology Data Exchange (ETDEWEB)

Destribats, Marie-Therese; Allain, C.; Prot, A.; Thome, P. [Centre d' Etudes Nucleaires de Saclay (France)

1965-09-15

nombreuses methodes de controle non destructif pour les differents materiaux entrant dans leur realisation, en particulier: radiographie et gammagraphie, methodes ultrasonores et courants de Foucault. Le memoire decrit les controles effectues au cours de la fabrication des reacteurs des filieres EDF (graphite- gaz) et EL4 (eau lourde) et des elements colamines. Les auteurs insistent sur quelques points caracteristiques de ces diverses methodes, dont certaines sont bien connues. Filiere EDF: mesure des epaisseurs des parois des tubes en uranium ou uranium allie, par ultrasons; recherche des cavites de ces tubes, par gammagraphie; controle des traitements thermiques de ces tubes, par ultrasons; recherche des defauts (inclusions, criques), par ultrasons dans les billettes et ebauches des gaines en MgZr; controle de l'etancheite des cartouches, par ressuage d'helium. Filiere EL 4: mesure des epaisseurs des parois des tubes de force, et de guidage en Zircaloy, par ultrasons et courants de Foucault; examen des tubes de force en Zircaloy et tube de gaine en acier inoxydable par ultrasons; radiographie sous vide des gaines en Be; controle d'etancheite des crayons, par ressuage d'helium. Elements colamines : mesure de l'epaisseur des gaines, par courants de Foucault puises; recherche de la position de l'ame sur tubes et plaques, par radiographie, comptage gamma et courants de Foucault puises; controle de l'homogeneite.du combustible, par comptage gamma; detection des defauts dans les billettes en ZrU, par ultrasons et gammagraphie; recherche des zones decollees des plaques, par ultrasons, courants de Foucault puises et mesure de resistivite. Ces mises au point ont ete faites en vue de leur utilisation industrielle. Plusieurs de ces controles sont effectues couramment en cours de fabrication, d'autres le seront prochainement et certains sont susceptibles de l'etre a plus longue echeance. (author) [Spanish] Los estudios del Commissariat a l'Energie Atomique sobre diversos tipos de

Neutronic study of the two french heavy water reactors; Etude neutronique des deux piles francaises a eau lourde

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [French] Les deux reacteurs francais - la pile de Chatillon, appelee ZOE, et la pile de Saclay, designee dans la suite par P2 - ont fait l'objet d'etudes neutroniques detaillees dont les principales sont exposees dans ce rapport. Ces etudes ont ete pour la plupart effectuees dans le cadre du Departement des Etudes de Piles (D.E.P.). Nous avons ainsi entre autre etudie la distribution du flux neutronique; les facteurs influencants la reactivite; le lien entre reactivite et divergence par la formule de Nordheim; le temps de vie moyen des neutrons; les spectres de neutrons de P2; l'effet xenon; ou encore l'effet des differents reglages des plaques et barres de controles. (M.B.)

Aspects of Reactor Physics Research at the Victoria University of Manchester; Quelques Aspects des Experiences de Physique des Reacteurs a l'Universite Victoria de Manchester; Aspekty ehksperimental'nykh issledovanij po fizike reaktorov v universitete viktorii v manchestere; Trabajos de Fisica Experimental con Reactores Efectuados en la Universidad Victoria de Manchester

Energy Technology Data Exchange (ETDEWEB)

Harris, M. J.; Walton, D. G. [Victoria University of Manchester (United Kingdom)

1964-02-15

constructed. Its mechanical design gives considerable flexibility so that, for instance, measurements parallel and perpendicular to the fuel rods are greatly facilitated. A programme of steady-state measurements is under way. Future work is outlined, and includes fine structure measurements, voidage effects and pulsed neutron studies. (author) [French] Le Departement du genie nucleaire de l'Universite de Manchester a ete cree en 1959. Depuis lors, les etudes post-universitaires de physique des reacteurs se sont progressivement developpees et elargies en partant virtuellement de zero; les travaux ont porte sur les reseaux a eau ordinaire et notamment sur les experiences exponentielles a uranium naturel et a eau ordinaire alimentees par un accelerateur de particules. Les auteurs passent en revue les travaux effectues, etudient les resultats obtenus, donnent des apercus sur les recherches futures et illustrent leur expose par la description de diverses techniques experimentales adoptees a Manchester, qui sont peu onereuses et ne necessitent qu'un personnel reduit. Les principaux sujets de recherches sont decrits ci-apres. Les auteurs ont etudie la diffusion des neutrons dans l'eau ordinaire en employant successivement la methode de la source puisee et celle de la source stationnaire. Avec la premiere methode, ils se sont astreints a faire une analyse harmonique complete, au point d'etudier effectivement les modes superieurs alors que, par le passe, ont cherchait seulement a les eliminer. Au moyen de la methode de la source stationnaire, ils ont cherche surtout a eliminer tous les effets dus a la dimension de la source finie et du detecteur, al'activation par resonance, a la perturbation du flux, etc. Ils discutent et comparent les resultats de ces deux etudes. Le memoire decrit ensuite une mesure tres precise des sections efficaces d'absorption, egalement en cours, par la methode des neutrons puises, en prenant soin d'eliminer les effets harmoniques et autres, generateurs d

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.

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

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

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)

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

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)

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)

Study of thermocouples for control of high temperatures; Etude de thermocouples pour le reperage des hautes temperatures

Energy Technology Data Exchange (ETDEWEB)

Villamayor, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Faculte des Sciences de l' Universite de Lyon - 69 (France)

1967-07-01

Previous works have shown that the tungsten-rhenium alloys thermocouples were a good instrument for control of high temperatures. From its, the author has studied the W/W 26 per cent and W 5 per cent Re/W 26 per cent Re french manufactured thermocouples and intended for control of temperatures in nuclear reactors until 2300 deg. C. In 'out-pile' study he determines the general characteristics of these thermocouples: average calibration curves, thermal shocks influence, response times, and alloys allowing the cold source compensation. The evolution of these thermocouples under thermal neutron flux has been determined by 'in-pile' study. The observations have led the author to propose a new type of thermocouples settled of molybdenum-columbium alloys. (author) [French] Des travaux anterieurs ont montre que les thermocouples des alliages tungstene-rhenium etaient susceptibles de reperer avec precision des hautes temperatures. A partir de la, l'auteur a etudie las thermocouples W/W 26 pour cent Re et W 5 pour cent Re/W 26 pour cent Re de fabrication francaise et destines au controle des temperatures dans les reacteurs nucleaires, jusqu'a 2300 deg. C Dans l'etude 'hors-pile' il a determine les caracteristiques generales de ces thermocouples: courbes d'etalonnage moyen, influence des chocs thermiques, temps de reponse, et alliages assurant la compensation de soudure froide. L'etude 'en-pile' a permis de rendre compte de l'evolution de ces thermocouples sous flux neutroniques. Les phenomenes observes ont conduit l'auteur a proposer un nouveau type de thermocouples constitues d'alliages molybdene-niobium. (auteur)

Activity report 1999; Rapport d'activites 1999

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

The aim of this report is to outline the main developments of the ''Departement des Reacteurs Experimentaux'', (DRE) during the year 1999. DRE is one of the Department of the ''Direction des Reacteurs Nucleaires'', itself depending of the CEA Institution. After a presentation of the year highlights, this report gathers the main research and development programs. The second part concerns the production of radioisotopes, the silicon doping, the neutron radiography, the Orphee experiments and the activation analysis. The installations management, the closed reactors improvement program and the effluents and wastes processing of Grenoble, are presented in the other parts. Data on staff, budget and safety are also provided. (A.L.B.)

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.

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.)

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)

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

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.

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

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

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.

Irradiation and development of the nuclear emulsions exposed to intense fluxes of thermal neutrons with {gamma} rays; Irradiation et developpement des emulsions nucleaires exposees a des flux intenses de neutrons thermiques, accompagnes de rayons {gamma}

Energy Technology Data Exchange (ETDEWEB)

Faraggi, H; Bonnet, A; Cohen, J [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1952-07-01

The thermal neutron fluxes provided by nuclear reactors permit the survey of relatively rare phenomenons, and dosage of very weak quantities of some elements. One of the most favorable detection technique are constituted by the use of the nuclear emulsions. one can mention: - the dosage of uranium by counting in the emulsion the number of traces due to fission fragments after irradiation. - The dosage of the lithium and the boron as trace amounts with the help of nuclear reactions (n, {alpha}) and thermal neutrons. - The research of reactions (n, {alpha}) or (n, p) of very weak cross section for middle or heavy elements. These different applications require however important neutrons fluxes. It had therefore obliged us to search for the most favorable irradiation and development of the emulsions conditions, to get the best visibility of the trajectories and decrease the phenomena of fog on the emulsion, which prevents any observation. (M.B.) [French] Les flux de neutrons thermiques fournis par les reacteurs nucleaires permettent l'etude de phenomenes relativement rares, et le dosage de tres faibles quantites de certains elements. Un des moyens de detection les plus favorables est constitue par l'utilisation des emulsions nucleaires. on peut citer: - le dosage de l'uranium par comptage dans l'emulsion du nombre de traces dues aux fragments de fission apres irradiation. - Le dosage du lithium et du bore a l'etat de traces a l'aide des reactions (n, {alpha}) sous l'action des neutrons thermiques. - La recherche de reactions (n,{alpha}) ou (n,p) de tres faible section efficace pour des elements moyens ou lourds. Ces differentes applications necessite cependant des flux de neutrons important. On a donc ete amene a rechercher les conditions les plus favorables d'irradiation et de developpement des emulsions, de maniere a obtenir la meilleure visibilite des trajectoires et diminuer les phenomenes de voile de l'emulsion, qui empeche toute observation. (M.B.)

Irradiation and development of the nuclear emulsions exposed to intense fluxes of thermal neutrons with {gamma} rays; Irradiation et developpement des emulsions nucleaires exposees a des flux intenses de neutrons thermiques, accompagnes de rayons {gamma}

Energy Technology Data Exchange (ETDEWEB)

Faraggi, H.; Bonnet, A.; Cohen, J. [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1952-07-01

The thermal neutron fluxes provided by nuclear reactors permit the survey of relatively rare phenomenons, and dosage of very weak quantities of some elements. One of the most favorable detection technique are constituted by the use of the nuclear emulsions. one can mention: - the dosage of uranium by counting in the emulsion the number of traces due to fission fragments after irradiation. - The dosage of the lithium and the boron as trace amounts with the help of nuclear reactions (n, {alpha}) and thermal neutrons. - The research of reactions (n, {alpha}) or (n, p) of very weak cross section for middle or heavy elements. These different applications require however important neutrons fluxes. It had therefore obliged us to search for the most favorable irradiation and development of the emulsions conditions, to get the best visibility of the trajectories and decrease the phenomena of fog on the emulsion, which prevents any observation. (M.B.) [French] Les flux de neutrons thermiques fournis par les reacteurs nucleaires permettent l'etude de phenomenes relativement rares, et le dosage de tres faibles quantites de certains elements. Un des moyens de detection les plus favorables est constitue par l'utilisation des emulsions nucleaires. on peut citer: - le dosage de l'uranium par comptage dans l'emulsion du nombre de traces dues aux fragments de fission apres irradiation. - Le dosage du lithium et du bore a l'etat de traces a l'aide des reactions (n, {alpha}) sous l'action des neutrons thermiques. - La recherche de reactions (n,{alpha}) ou (n,p) de tres faible section efficace pour des elements moyens ou lourds. Ces differentes applications necessite cependant des flux de neutrons important. On a donc ete amene a rechercher les conditions les plus favorables d'irradiation et de developpement des emulsions, de maniere a obtenir la meilleure visibilite des trajectoires et diminuer les phenomenes de voile de l'emulsion, qui

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

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)

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

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

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

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.

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)

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

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)

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

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.

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)

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)

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

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)

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

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

Contribution to the study of nuclear fuel materials with a metallic uranium base; Contribution a l'etude des materiaux combustibles nucleaires a base d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

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

1957-11-15

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

Fine structure and spectral index measurements in natural uranium - graphite lattices; Mesures fines dans des reseaux a graphite

Energy Technology Data Exchange (ETDEWEB)

Cogne, F; Journet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experiments described in this report have been carried out for the most part in the critical facility MARIUS, and a few during the start up of the EDF-1 power reactor. The first part deals with the fine structure measurements made in various lattices and with their analysis. Integration over the neutron spectrum of the mono-kinetic disadvantage factor derived by the A.B.H method yields results in good agreement with the experiments. The second part deals with spectral indexes measurements (Pu/U, In/Mn) made at room temperature in MARIUS. Comparison are made of experiments with calculations using various thermalization models. Experiments carried out at higher temperatures in EDF-1 are also described. (authors) [French] Les mesures decrites dans ce rapport ont ete faites pour la plupart dans l'empilement critique MARIUS sur des reseaux a graphite-uranium naturel. Une premiere partie traite des mesures de structure fine faites dans differents reseaux et de leur interpretation. On montre en particulier qu'une integration sur le spectre d'un calcul monocinetique type A.B.H. rend bien compte des experiences. Dans une deuxieme partie, on donne les resultats de mesures d'indices de spectre Pu/U et In/Mn faites sur des reseaux froids a MARIUS et leur comparaison avec les differents modeles de calculs de thermalisation. On donne egalement les resultats de quelques mesures en temperature effectuees lors du demarrage du reacteur EDF-1. (auteurs)

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.)

Use of cadmium in solution in the EL 4 reactor moderator irreversible fixing of cadmium on the metallic surfaces; Utilisation du cadmium en solution dans le moderateur du reacteur EL 4 - fixation irreversible du cadmium sur les surfaces metalliques

Energy Technology Data Exchange (ETDEWEB)

Croix, O; Paoli, O; Lecomte, J; Dolle, L; Gallic, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research into the poisoning of the EL-4 reactor by cadmium sulphate, measurements have been made by two different methods of the residual amounts of cadmium liable to be fixed irreversibly on the surfaces in contact with the heavy water. A marked influence of the pH has been noticed. The mechanism of the irreversible fixing is compatible with the hypothesis of an ion-exchange in the surface oxide layer. In a sufficiently wide range of pH the cadmium thus fixed causes very little residual poisoning. The stability of the cadmium sulphate solutions is however rather low in the conditions of poisoning. (authors) [French] Dans le cadre des etudes sur l'empoisonnement du reacteur EL-4 par le sulfate de cadmium, les quantites residuelles de cadmium susceptibles de se fixer irreversiblement sur les parois que mouillerait l'eau lourde, ont ete mesurees experimentalement par deux methodes differentes. On observe une influence nette du pH. Le mecanisme de la fixation irreversible est compatible avec l'hypothese d'un echange d'ions dans la pellicule d'oxyde superficielle. Dans des limites suffisamment larges de pH, la cadmium ainsi fixe n'occasionne pas d'empoisonnement residuel important. La stabilite des solutions de sulfate de cadmium dans les conditions de l'empoisonnement est cependant mediocre. (auteurs)

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)

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 −})

Thermal neutron flux measurements using neutron-electron converters; Mesure de flux de neutrons thermiques avec des convertisseurs neutrons electrons

Energy Technology Data Exchange (ETDEWEB)

Le Meur, R; Lecomte, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The operation of neutron-electron converters designed for measuring thermal neutron fluxes is examined. The principle is to produce short lived isotopes emitting beta particles, by activation, and to measure their activity not by extracting them from the reactor, but directly in the reactor using the emitted electrons to deflect the needle of a galvanometer placed outside the flux. After a theoretical study, the results of the measurements are presented; particular attention is paid to a new type of converter characterized by a layer structure. The converters are very useful for obtaining flux distributions with more than 10{sup 7} neutrons cm{sup -2}*sec{sup -1}. They work satisfactorily in pressurized carbon dioxide at 400 Celsius degrees. Some points still have to be cleared up however concerning interfering currents in the detectors and the behaviour of the dielectrics under irradiation. (authors) [French] On examine le fonctionnement de convertisseurs neutrons electrons destines a des mesures de flux de neutrons thermiques. Le principe est de former par activation des isotopes a periodes courtes et a emission beta et de mesurer leur activite non pas en les sortant du reacteur, mais directement en pile, utilisant les electrons emis pour faire devier l'aiguille d'un galvanometre place hors flux. Apres une etude theorique, on indique des resultats de mesures obtenus, en insistant particulierement sur un nouveau type de convertisseur, caracterise par sa structure stratifiee. Les convertisseurs sont tres interessants pour tracer, des cartes de flux a partir de 10{sup 7} neutrons cm{sup -2}*s{sup -1}. Ils sont utilisables pour des flux de 10{sup 14} neutrons cm{sup -2}*s{sup -1}. Ils fonctionnent correctement dans du gaz carbonique sous pression a 400 C. Des points restent cependant a eclaircir concernant les courants parasites dans les detecteurs et le comportement des dielectriques pendant leur irradiation. (auteur)

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)

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)

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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)

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

Experimental studies of some of the physical features of beryllium-moderated intermediate reactors; Etude experimentale de quelques particularites physiques des reacteurs a neutrons intermediaires, ralentis au beryllium; Ehksperimental'ny e issledovaniya nekotorykh fizicheskikh osobennostej promezhutochnykh reaktorov s berillievym zamedlitelem; Estudios experimentales de algunas caracteristicas fisicas de los reactores intermedios moderados con berilio

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A I; Kuznetsov, V A; Artyukhov, G Ya; Mogil' ner, A I; Prokhorov, Yu A; Steklovski, V M; Chernov, L A [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

1962-03-15

of neutrons absorbed by the uranium. The paper provides data, derived from the same assembly, on the efficiency of rods made of various absorbing materials. It gives the experimentally measured distribution of neutron density for neutrons of various energies in the neighbourhood of a boron-carbide rod, and the density of neutron captures by a 1/v detector within the rod. The paper also discusses methods used and the results obtained from experiments designed to assess the efficiency of recompensation, cylinders placed on the boundary between core and reflector. (author) [French] Le memoire analyse les resultats de plusieurs experiences effectuees sur l'ensemble critique PF-4, qui est destine a l'etude detaillee des particularites physiques des reacteurs a neutrons intermediaires. Les coeurs et les reflecteurs des differents esembles critiques etaient constitues par un assemblage compact de tubes en acier ou en aluminium dans lesquels etaient inseres des diques de diverses matieres. En combinant selon differentes proportions les disques d'uranium enrichi a 90% et les matieres de ralentissement et en introduisant dans le reflecteur des couches de ralentisseur de diverses epaisseurs, on a pu obtenir de grandes modifications du spectre des neutrons provoquant la fission. Le memoire decrit l'ensemble critique PF-4 et les differents assemblages qui le composent. Les auteurs analysent l'efficacite relative du ralentissement interieur et exterieur pour des reacteurs dans lesquels le rapport noyaux du ralentisseur noyaux d'uranium dans le coeur est tres peu eleve. Il ressort des experiences que, meme lorsqu'on emploie des reflecteurs tres epais, la faible dilution de l'uranium par le ralentisseur (le rap- port entre les noyaux du beryllium et de l'uranium-235 etant: {partial_derivative}Be/{partial_derivative}{sup 235}U{approx_equal}1) entraine un accroissement de la masse critique. Une partie importante du memoire est consacree a une analyse des effets hetero- genes produits

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)

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)

Neutron flux determinations in the reactors G2 and G3 during operation; Releves du flux neutronique dans les reacteurs G2 et G3 en puissance

Energy Technology Data Exchange (ETDEWEB)

Boulinier, C; Faurot, P; Sagot, M; Teste du Bailler, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

After demonstrating the sensitivity of the distribution of power in a production reactor to a deformation caused by dissymmetries of reactivity in the reactor, the authors describe the method of neutron flux determination devised for the reactors G2 and G3 under working conditions; the detector used is a tungsten or nickel wire, the {gamma} activity of which is measured with an ionisation chamber. Several flux determinations are given as examples to illustrate the sensitivity of the method. (author) [French] Apres avoir mis en evidence la sensibilite de la repartition de la puissance dans un reacteur de production a une deformation provoquee par de faibles dissymetries de reactivite dans le reacteur, les auteurs decrivent la methode de releve du flux neutronique mise au point pour les reacteurs G2 et G3 en puissance; le detecteur utilise est un fil de tungstene ou de nickel dont l'activite {gamma} est mesuree a l'aide d'une chambre d'ionisation. Quelques releves de flux illustrant la sensibilite de la methode sont donnes a titre d'exemple. (auteur)

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

Heat exchanges during the re-flooding of a water reactor core - within the framework of the 'reference accident'; Echanges thermiques lors du renoyage d'un coeur de reacteur a eau - dans le cadre de 'l'accident de reference'

Energy Technology Data Exchange (ETDEWEB)

Andreoni, Daniel

1975-11-28

After a brief presentation of reported studies made in different countries and regarding the so-called 'reference accident', this research thesis reports the study of reactor re-flooding when the reactor is completely dried and heating elements have reached a temperature between 300 and 900 C, with a constant water flow rate entering the test section, with a constant dissipated electrical power, and by using very simple geometries. After a first part addressing the experimental study, the author reports the development of conduction calculation codes used to compute the flow extracted by the two-phase flow, present the thermal-hydraulic code used to compute local values and to study the correlation of the upstream area exchange coefficient. The author finally reports an analysis of the different existing models and the study of a re-flooding model [French] La presente etude est consacree a l'un des aspects de la surete des reacteurs a eau sous pression, et plus precisement a l'accident tres important qui consiste en une perte de fluide caloporteur (Loss of Coolant Accident - 'LOCA'). Le but de l'etude est de fournir des renseignements necessaires a l'interpretation des experiences effectuees sur des grappes, de donner une correlation de coefficient d'echange dans la zone aval, et de donner aussi un modele de progression du front de trempe pour les analyses de surete. Une etude bibliographique preliminaire nous a permis de faire le point sur les experiences entreprises concernant le refroidissement de secours. Ensuite, les chapitres suivants seront decrits: 1) Le chapitre II, consacre a l'etude experimentale (boucle, sections d'essais, resultats globaux). 2) Le chapitre III ou seront presentes les codes de calcul de conduction, necessaires au calcul du flux extrait par le melange diphasique, le code de thermohydraulique necessaire au calcul des grandeurs locales et l'etude de la correlation du coefficient d'echange de la zone aval. 3) Enfin le chapitre IV ou, apres

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

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

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.)

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)

Cost of transporting irradiated fuels and maintenance costs of a chemical treatment plant for irradiated fuels; Cout de transport des combustibles irradies et cout d'entretien d'une usine de traitement chimique des combustibles irradies

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

other plants, an attempt has been made to express the maintenance costs as a percentage of the investments corresponding to each of the sections considered. The unequal aspect of maintenance in a plant of this type is demonstrated, this being due particularly to the inaccessibility of most of the equipment during operation. Finally some conclusions are drawn on what the maintenance costs could be and the means of reducing them in future plants. (author) [French] Le cout du cycle des combustibles a fait l'objet de nombreuses etudes mais beaucoup d'entre elles sont basees sur des etudes a priori et sont donc plus ou moins sujettes a caution. C'est ainsi que dans la partie ayant trait au traitement des combustibles irradies, des elements importants du cout n'ont que rarement ete precises a la suite d'experiences pratiques: le cout du transport des combustibles eux-memes et le cout d'entretien de l'usine. Les etudes relatives au cout du transport sont generalement basees sur des calculs faits a partir de donnees un peu arbitraires. Les etudes qui ont ete faites en France pour le transport d'uranium irradie entre les reacteurs EdF de Chinon et l'usine de retraitement de La Hague et l'uranium irradie des reacteurs de recherches jusqu'aux usines de retraitement etrangeres sont exposees et montrent qu'il a ete possible d'arriver a des types de chateaux de transport et des modalites d'expedition qui permettent de diminuer les couts dans des proportions tres importantes. Ceci a pu etre obtenu soit en combinant les transports par rail et par route soit par l'augmentation des capacites unitaires des chateaux de transport: on cite le cas d'un chateau de transport pour element des piles piscines qui est capable de transporter un coeur complet d'une pile a la fois entrainant une reduction substantielle du cout. Les etudes concernant les couts d'entretien d'usines de retraitement sont encore plus rares, or, dans les usines a entretien direct, ces couts sont une fraction non negligeable

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

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

Measurement of the in-pile core temperature of an EL-4 pencil element, first charge (can of type-347 stainless steel, 0.4 mm thick, UO{sub 2} fuel, 11 mm diameter). Determination of the apparent thermal conductivity integral of in-pile UO{sub 2}; Mesure de la temperature a coeur en pile d'un crayon EL-4 1er jeu (gaine acier inoxydable, nuance 347 - epaisseur 0,4 mm - combustible UO{sub 2} - diametre 11 mm). Determination de l'integrale de conductibilite thermique apparente de l'UO{sub 2} en pile

Energy Technology Data Exchange (ETDEWEB)

Lavaud, B; Ringot, C; Vignesoult, N [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-11-01

'un element combustible EL-4, type premier jeu, a gaine en acier inoxydable. On mesure cette temperature au coeur du crayon en pile a l'aide d'un thermocouple pour haute temperature: tungstene-rhenium a gaine tantale. Le crayon est place dans des conditions de fonctionnement analogues a celles de EL-4, tant en ce qui concerne la puissance specifique et la temperature sur gaine que la pression externe sur la gaine. La puissance specifique est obtenue dans le reacteur EL-3 avec un enrichissement de l'UO{sub 2} legerement superieur a celui normalement prevu pour EL-4. La temperature de gaine et la pression visees sont realisees a l'aide d'un conteneur d'irradiation en zircaloy-2 et a remplissage NaK, adapte, aux conditions du reacteur EL-3. - Les temperatures de l'UO{sub 2} a coeur, et de la surface de la gaine etant mesurees; - La puissance etant calculee a partir des echanges thermiques dans le conteneur etalonne en laboratoire; - La chute de temperature au contact UO{sub 2}-gaine etant deduite de mesures faites en laboratoire dans des conditions de flux calorifique comparables et sous une atmosphere de gaz correspondant au debut de la vie de l'element combustible; on peut tracer la courbe integrale de conductibilite. Les examens micro-graphiques de la structure de l'oxyde permettent de verifier la repartition des temperatures dans l'oxyde, deduite de l'integrale de conductibilite thermique. (auteurs)

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.)

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

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)

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)

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

Possibilities and limitations of analogue methods for studying the dynamics of nuclear power stations; Possibilites et limitations du calcul analogique pour les etudes dynamiques de centrales nucleaires

Energy Technology Data Exchange (ETDEWEB)

Caillet, C; Deat, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

1. Introduction: the present paper is devoted to analog simulation of problems related to nuclear reactors other than the simulation of the kinetic equations which is well known. 2. Thermodynamic problems: various problems relative to temperature evolution in a reactor, in a pipe, in an exchanger, in a turbine, are studied, and simulation techniques used by earlier authors are critically reviewed. 3. Pipe simulators: it is shown that this problem can be solved by the use of specialized simulators which will be described and analysed. 4. Rotating machine simulators: the particular aspect of rotating machine calculations introducing frequent use of diagrams is emphasized. A simulator requiring both digital and analogue methods is described. 5. The study of a nuclear power station: as an example it is proposed to discuss problems a rising in connection with the preceding elements (a, b, c, d) when simulating the behaviour of large nuclear plants. The part played by ordinary computing elements for the simulation of the different servomechanism transfer functions is considered and process of regulation is outlined. 6. Conclusion: the necessity of the use of high quality simulators and computers is underlined and the accuracy of the solutions is discussed. (author)Fren. [French] 1. Cinetique des reacteurs: la simulation des equations cinetiques d'un reacteur nucleaire ne pose desormais plus de probleme. II est donc possible de faire le point des differentes applications de la technique analogique dans ce domaine. 2. Les problemes thermodynamiques: on discute les differents problemes poses par l'evolution des temperatures dans un reacteur, dans une tuyauterie, dans un echangeur, dans une turbine, et on passe en revue les techniques de simulation proposees jusqu'a ce jour. 3s simulateurs de tuyauteries: on montre comment les differents problemes poses ci-dessus peuvent etre resolus, pour une classe tres vaste de reacteurs par l'emploi de simulateurs speciaux que l

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)

Slow Neutron Spectrometers at the Swedish Reactors; Spectrometres a Neutrons Lents des Reacteurs Suedois; 0421 041f 0415 041a 0422 0420 041e 041c 0415 0422 0420 042b 041c 0415 0414 041b 0415 041d 041d 042b 0425 041d 0415 0419 0422 0420 041e 041d 041e 0412 041d 0410 0428 0412 0415 0414 0421 041a 0418 0425 0420 0415 0410 041a 0422 041e 0420 0410 0425 ; Espectrometros para Neutrones Lentos en los Reactores de Suecia

Energy Technology Data Exchange (ETDEWEB)

Dahlborg, U.; Skoeld, K. [AB Atomenergi, Stockholm (Sweden); Larsson, K. -E. [Royal Institute of Technology, Stockholm (Sweden)

1965-06-15

is briefly discussed for illustrational purposes. A comparison between the light- and heavy-water moderated reactors for beam tube work shows the distinct advantages of the heavy-water type. (author) [French] Aux centres crees autour des deux, reacteurs de recherche suedois, Rl a Stockholm et R2 a Studsvik, on a maintenant la possibilite d'utiliser quatre spectrometres differents pour les experiences de diffusion inelastique des neutrons. A Stockholm, le reacteur Rl de 600 kW, ralenti a l'eau lourde, est equipe de deux spectrometres mecaniques a neutrons lents qui fonctionnent simultanement, Avec l'un, on utilise toujours un monochromateur a filtre en Be; avec l'autre, on peut employer soit le meme genre de monochromateur, soit un monochromateur a cristal. On a constate que pour les mesures de distribution angulaire, on obtient d'excellents resultats en combinant un monochromateur a cristal et un spectrometre mecanique, meme si l'intensite et le pouvoir de resolution sont relativement faibles. Recemment on a fait l'essai d'un selecteur de vitesse mecanique ayant un pouvoir de separation des longueurs d'onde de 4,2%. Cependant, cet instrument n'est pas encore utilise pour les experiences. Le spectrometre mecanique de Studsvik, avec lequel le reacteur R2 de 30 MW ralenti a l'eau legere est equipe, utilise pour la monochromatisation l'action combinee d'un monochromateur a filtre de Be et d'un hacheur a courbe de transmission etroite. Dans ce spectrometre, de meme que dans celui de Stockholm, le hacheur est place avant l'echantillon, ce qui permet l'enregistrement simultane de donnees pour des angles d'observation differents. Un spectrometre a cristal triaxial est aussi en service pres du reacteur R2. Les auteurs donnent certaines caracteristiques de ces instruments, notamment l'intensite, le pouvoir de resolution, et indiquent dans quelle mesure ils conviennent pour certaines operations. Ainsi, il ressort des donnees numeriques mentionnees qu'une amelioration assez

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)

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

Recent progress in the detection of bursts in the canning in French reactors; Progres recents de la detection des ruptures de gaines dans les reacteurs francais G1, EL2, G3, EL3

Energy Technology Data Exchange (ETDEWEB)

Goupil, J; Grenon, M; Raffailhac, J; Roguin, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

des produits de fission, 2) de la pollution d'uranium des gaines et de la pollution eventuelle des canaux apres ruptures de gaines rapides. L'evolumetre est constitue par une memoire qui stocke les valeurs de l'activite des canaux prises a un instant considere comme reference. A cette memoire, on vient comparer les valeurs de l'activite des canaux en cours de prospection. Une difference entre ces valeurs indique l'apparition ou l'evolution d'une fissure de gaine. Pour tenir compte des variations du regime thermodynamique dans les canaux, les valeurs extraites de la memoire sont corrigees par un signal provenant d'un detecteur d'activite place dans le circuit general de sortie du gaz de la pile. Dans le cas de la pile EL{sub 2}, egalement a refroidissement par CO{sub 2}, sous pression, une methode analogue a celle de G{sub 3} a ete utilisee. Des echantillons de gaz de refroidissement sont preleves dans chacune des 133 cellules de la pile successivement par l'ouverture d'electrovannes. Le gaz est filtre et les produits de fission sont extraits par une methode de collection electrostatique. Un scintillateur et une chaine electronique fournissent un signal specifique des produits de fission qui s'inscrit sur un enregistreur. Dans le cas d'un depassement du seuil d'activite, la cellule incriminee est isolee du systeme de prospection et prise en charge par un detecteur 'suiveur' qui permet de suivre l'evolution de la fissure. Une annee d'exploitation de la pile G1 qui est refroidie a l'air a la pression atmospherique a permis d'obtenir des resultats sur le fonctionnement du dispositif D.R.G. ce qui nous a amenes a perfectionner le dispositif initial en installant un evolumetre du type decrit ci-dessus pour G{sub 3}. Le reacteur EL{sub 3}, refroidi a l'eau lourde, utilise un systeme de detection base sur la mesure, au moyen de compteurs G.M., de l'activite des gaz de fission entraines par de l'helium dilue dans l'eau lourde puis extraits de celle-ci par des hydrocyclones. La

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

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.

Major accident analyses for experimental zero-power fast reactor assemblies; Analyse des accidents graves pouvant survenir dans les reacteurs experimentaux a neutrons rapides de puissance zero; Analiz krupnoj avarii dlya ehksperimental'ny kh reaktornykh ustanovok nulevoj moshchnosti na bystrykh nejtronakh; Analisis de los accidentes graves que pueden producirse en los reactores experimentales rapidos de potencia cero

Energy Technology Data Exchange (ETDEWEB)

Fischer, G.; Barts, E. W.; Kapil, S.; Tomabechi, K. [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

systems with the very soft neutron-energy spectra characteristic of large oxide power breeders. (author) [French] Les auteurs ont etudie la possibilite, le mecanisme et les consequences de la fusion et autres accidents nucleaires graves dans les reacteurs experimentaux a neutrons rapides de puissance zero, du type ZPR-III, a coeur divise. Cette etude a ete completee par une evaluation de l'importance de l'effet Doppler sur un grand nombre de reacteurs de ce type. Les auteurs demontrent qu'il est fort peu probable qu'une fusion se produise, du fait que la conjonction des circonstances qui pourraient la provoquer est difficilement realisable. L'expose du mecanisme de fusion est suivi de l'analyse des resultats de calculs couples neutronique-hydrodynamiqu e relatifs a deux reacteurs de puissance zero. On a choisi pour cette etude un coeur de 1200 l, qui correspond a un reacteur relativement grand a coeur normal. L'etude a egalement porte sur un coeur plus petit ayant un coefficient cavitaire plus important, qui pourrait presenter un plus grand danger. Chaque systeme a eu un comportement en fonction du temps tout a fait different. Si un accident grave survient dans un reacteur de puissance zero, les atomes de {sup 235}U, isoles dans les plaques d'uranium enrichi, s'echauffen t tres rapidement tandis que le reste du coeur demeure pratiquement froid; il y a ainsi formation d'un gaz du {sup 235}U qui donne lieu a la pression de rupture. Les auteurs expliquent l'adaptation qu'ils ont faite du code AX-I de neutronique-hydrodynamiqu e pour l'appliquer a un gaz de Van der Waals. Une autre modification importante de l'equation d'etat utilisee dans ce code consiste a employer une equation du type Mie-Grueneisen, derivee de la theorie de l'etat solide. Cette modification permet d'evaluer de facon plus satis- faisante le terme de pression pour les coeurs de composition variable. Du fait que les plaques en uranium fortement enrichi d'un reacteur de puissance zero s'echauffent plus

Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

International Nuclear Information System (INIS)

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

1995-01-01

A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

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.

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.

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

Studies on the Electrochemical Dissolution for the Treatment of 10 g-Scale Zircaloy-4 Cladding Hull Wastes in LiCl-KCl Molten Salts

International Nuclear Information System (INIS)

Lee, You Lee; Lee, Jang Hwa; Jeon, Min Ku; Kang, Kweon Ho

2012-01-01

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in 500 degree C LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

Comparison Of The Worth Of Critical And Exponential Measurements For Heavy-Water-Moderated Reactors; Valeur Relative des Mesures Critiques et Exponentielles pour l'Etude des Reacteurs Ralentis a l'Eau Lourde; Sravnenie tsennosti kriticheskikh i ehksponentsial'nykh izmerenij dlya reaktorov s tyazhelovodnym zamedlitelem; Valor Relativo de las Mediciones Criticas y Exponenciales para los Reactores Moderados por Agua Pesada

Energy Technology Data Exchange (ETDEWEB)

Graves, W. E.; Hennelly, E. J. [Savannah River Laboratory, E.I. Du Pont De Nemours and Co., Aiken, SC (United States)

1964-02-15

direct effects in mock-ups and as a test for heterogeneous and two-dimensional diffusion calculations. (6) Criticality studies of heavy-water lattice fuel in light water The SRL exponentials have proved particularly valuable for criticality studies to determine safe methods of handling enriched fuel in light water. High accuracy is not required in this case, and the generalized exponential buckling studies are definitely preferable to the more particularized critical studies. (author) [French] En regle generale, les experiences critiques et exponentielles sur des reseaux de reacteurs fournissent des renseignements qui font double emploi. Durant les dix dernieres annees, le Savannah River Laboratory (SRL) a fait fonctionner simultanement un ensemble critique a eau lourde (PDP) et un ensemble exponentiel (SE). Les auteurs exposent brievement l'experience acquise au SRL, indiquent les resultats obtenus et font des recommandations au sujet de la possibilite d'appliquer ces deux genres d'installations dans differentes experiences. Les auteurs examinent les six types d'experiences ci-apres: 1. Mesures du laplacien dans les reseaux isotropiques uniformes Le SRL a procede a de nombreuses comparaisons entre les mesures faites a l'aide d'ensembles critiques a une seule region, d'ensembles exponentiels, d'ensembles critiques a substitution et du reacteur d'essai des constantes physiques (PCTR). El semble que les seules difficultes que presentent les experiences exponentielles, resident dans les determinations du laplacien dans le sens radial. Si l'on reussit a faire ces determinations, les experiences exponentielles peuvent etre comparees favorablement aux experiences critiques. Les ensembles critiques a une seule region necessitent le plus de matieres; viennent ensuite les ensembles critiques a substitution et les ensembles exponentiels dont les besoins sont en gros comparables; enfin le PCTR ou les mesures en exigent le moins. 2. Effets anisotropiques et effets cavitaires Des

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)

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

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

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

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.

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.)

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

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)

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

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.

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)

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)

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

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.

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

Processing Th C{sub 2} - UC{sub 2} fuel extracted from high temperature reactors HTGCR; Etude du traitement des combustibles Th C{sub 2} - UC{sub 2} issus de reacteurs a haute temperature

Energy Technology Data Exchange (ETDEWEB)

Derrien, C; Lessart, P; Pianezza, E; Verry, C; Villain, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1969-07-01

The object of this investigation is solubilisation head-end (from crushing and grinding phase to non included first purification phase) of pulverulent ({sup 233}U/{sup 232}Th)C{sub 2} (200 - 500 microns diameter) contained in a graphite matrix extracted from a 4.10{sup 13} n.cm{sup -2}.s{sup -1} thermalized neutrons average flux with an irradiation of 80000 MWjT{sup -1} HTGCR reactor. After having succinctly described different bibliographic processes we have chosen the burn - leach of reactor fuel and graphite matrix containing it. The technology of burner is original in nuclear field and still more by utilizing ultra-sounds to intensify burning reaction and to minimize the weight of unburnables. The mixture of ThO{sub 2}, U{sub 3}O{sub 8}, and fission products oxides is solubilized by boiling HNO{sub 3} 13 M + HF 0.05 M. This process is profit-learning in a thorium recuperation and reprocessing point of view. In the contrary-case it would be interesting to consider a dry-process which would permit to separate solid ThF{sub 4} from gaseous UF{sub 6}. (authors) [French] Cette etude a pour objet le traitement initial de mise en solution ou 'head-end' (allant de la phase broyag-concassage a la phase de premiere purification exclue) d'un combustible ({sup 233}U/{sup 232}Th)C{sub 2} pulverulent (de 200 a 500 {mu} de diametre) contenu dans une matrice de graphite issu d'un reacteur HTGCR surgenerateur a neutrons thermiques de flux moyen 4. l0{sup 13} n.cm{sup -2}.s{sup -1} et taux d'irradiation 80000 MWjT{sup -1}. Apres exposition succincte des differents procedes bibliographiques decrits, nous avons finalement choisi le traitement par combustion-attaque ('Burn-Leach') du combustible et de la matrice etanche graphite qui le contient. La technologie du bruleur est originale dans le domaine nucleaire d'autant qu'elle utilise les ultra-sons pour ameliorer le rendement de la reaction de combustion et reduire au minimum le poids des imbrules. Le melange ThO{sub 2}, U{sub 3}O

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

Contribution to the study of nuclear fuel materials with a metallic uranium base; Contribution a l'etude des materiaux combustibles nucleaires a base d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

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

1957-11-15

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

Special concretes for protection in piles (1963); Les betons speciaux dans la protection des piles (1963)

Energy Technology Data Exchange (ETDEWEB)

Condat, M.J.; Lafore, P.; Rastoin, J. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1963-07-01

The report reviews the main concretes used for the protection of reactors. First are examined the general factors affecting the selection of concretes for protection; some of the general aspects of radiation attenuation (fast and thermal neutrons, {gamma}) are then recalled. After an enumeration of the qualities and defects of conventional concretes, desirable or possible improvements are proposed: introduction of heavy elements for increasing the efficiency with respect to fast neutrons and {gamma}; increase in the hydrogen content (as H{sub 2}O), addition of absorbent products (B, Li, rare earths) for thermal neutron capture. A list is given of the principal products used for this purpose. Finally there is a rapid review of the preparation, the stability with time, and the resistance to radiation and to heat. Appendices and tables give details and numerical values. Appendix 1 deals with the question of water in concretes. Appendix 2 gives a classification of concretes based mainly on their density. At the end, ten tables give the compositions and values of the physical and mechanical characteristics of sixteen particularly typical concretes. (authors) [French] Le rapport passe en revue les principaux betons utilisables pour la protection des reacteurs. On examine d'abord les facteurs generaux influencant la selection des betons de protection; on rappelle ensuite brievement les aspects generaux de l'attenuation des rayonnements (neutrons rapides, neutrons thermiques, {gamma}). Apres avoir rappele les qualites et les defauts des betons classiques, on fait ressortir les perfectionnements souhaitables ou necessaires: introduction d'elements lourds pour ameliorer l'efficacite contre les neutrons rapides et les {gamma}; augmentation de la teneur en H (en H{sub 2}O); addition de corps absorbants (B, Li, terres rares) pour capturer les neutrons thermiques. On enumere les principaux corps utilises a cet effet. On termine par un tres rapide apercu sur

Special concretes for protection in piles (1963); Les betons speciaux dans la protection des piles (1963)

Energy Technology Data Exchange (ETDEWEB)

Condat, M J; Lafore, P; Rastoin, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1963-07-01

The report reviews the main concretes used for the protection of reactors. First are examined the general factors affecting the selection of concretes for protection; some of the general aspects of radiation attenuation (fast and thermal neutrons, {gamma}) are then recalled. After an enumeration of the qualities and defects of conventional concretes, desirable or possible improvements are proposed: introduction of heavy elements for increasing the efficiency with respect to fast neutrons and {gamma}; increase in the hydrogen content (as H{sub 2}O), addition of absorbent products (B, Li, rare earths) for thermal neutron capture. A list is given of the principal products used for this purpose. Finally there is a rapid review of the preparation, the stability with time, and the resistance to radiation and to heat. Appendices and tables give details and numerical values. Appendix 1 deals with the question of water in concretes. Appendix 2 gives a classification of concretes based mainly on their density. At the end, ten tables give the compositions and values of the physical and mechanical characteristics of sixteen particularly typical concretes. (authors) [French] Le rapport passe en revue les principaux betons utilisables pour la protection des reacteurs. On examine d'abord les facteurs generaux influencant la selection des betons de protection; on rappelle ensuite brievement les aspects generaux de l'attenuation des rayonnements (neutrons rapides, neutrons thermiques, {gamma}). Apres avoir rappele les qualites et les defauts des betons classiques, on fait ressortir les perfectionnements souhaitables ou necessaires: introduction d'elements lourds pour ameliorer l'efficacite contre les neutrons rapides et les {gamma}; augmentation de la teneur en H (en H{sub 2}O); addition de corps absorbants (B, Li, terres rares) pour capturer les neutrons thermiques. On enumere les principaux corps utilises a cet effet. On termine par un tres rapide apercu sur la mise en oeuvre, la

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

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

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.)

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

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

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

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)

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

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.)

Methods for determining thermal stresses values. Some examples relating to nuclear reactors; Methodes de determination des contraintes thermiques. Quelques exemples d'application aux reacteurs

Energy Technology Data Exchange (ETDEWEB)

Bernard, J; Gautier, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Peres, A [Israel Institute of Technology, Dept. of Nuclear Science Technion (Israel)

1958-07-01

As modern techniques develop more elaborate machines, and make their way towards higher and higher temperatures and pressures, the thermal stresses become a matter of major importance in the design of mechanical structures. In the first part of this paper, the authors examine the problem from a theoretical standpoint, and try to evaluate the aptitude and limitation of mathematical techniques to attain the quantitative values of thermal stresses. This paper deals mainly with the experimental methods to measure thermal stresses. The authors show some examples relating to nuclear reactors. (author)Fren. [French] Au fur et a mesure que la technique moderne developpe des machines plus poussees et s'oriente vers des temperatures et des pressions toujours plus elevees, les contraintes thermiques deviennent un facteur d'importance capitale dans le calcul des structures mecaniques. Les auteurs examinent d'abord l'aspect theorique du probleme, ainsi que l'aptitude et les limites du calcul pour exprimer quantitativement la valeur des contraintes thermiques. Les auteurs exposent principalement, ensuite, les methodes experimentales qui permettent de mesurer ces contraintes, et illustrent cet expose de quelques exemples relatifs aux installations nucleaires. (auteur)

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)

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

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.

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)

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

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)

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

Study of the formation and of the distribution of dissolved gases and hydrogen peroxide in water from a swimming-pool reactor (triton) (1961); Etude de la formation et de la repartition des gaz dissous et de l'eau oxygenee dans l'eau d'un reacteur piscine (triton) (1961)

Energy Technology Data Exchange (ETDEWEB)

Chenouard, J; Rozenberg, J; Dolle, L; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

In order to determine experimentally the amount of radiolysis in the swimming-pool reactor Triton, direct measurements have been made of the quantity of radiolysis gas and hydrogen peroxide in the water, at the entry and exit of the core. The concentration distribution of these gases in the reactor was also determined. An explanation is given as to why no gases evolution is seen in the swimming-pool reactors of the C.E.A. The overall amount of radiolysis is zero, and a simple interpretation of this result is possible. The real amount of radiolysis occurring in the reactor core can be calculated. This is in satisfactory agreement with certain measurement mad elsewhere. (authors) [French] Pour determiner experimentalement le taux de radiolyse dans la pile piscine Triton, des mesures directes de la quantite de gaz de radiolyse et d'eau oxygenee dans l'eau a l'entree et a la sortie du coeur ont ete faites. La repartition de la concentration de ces gaz dans la piscine a egalement ete determinee. On explique pourquoi aucun degagement gazeux n'est observe dans les piles piscines du CE.A. Le taux de radiolyse global est nul, et une interpretation simple de ce resultat est possible. Un taux de radiolyse reel dans le coeur du reacteur peut etre calcule. Celui-ci est en accord satisfaisant avec certaines determinations faites ailleurs. (auteurs)

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.)

Adiabatic calorimeter with static vacuum for the measurement of the heating of in- pile materials; Calorimetre adiabatique a vide statique pour la mesure d'echauffements de materiaux en pile

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

After having reviewed the various interaction processes occurring between radiations present in nuclear reactors and matter, the author describes the different calorimetric methods which may be used for measuring the energy absorbed in the materials. He then gives a detailed description of the adiabatic calorimeter, the associated measurement device and the calibration methods which have been chosen. He finally gives values for the heating produced at various experimental positions in the reactors EL-2 and EL-3 for several materials currently used in reactor construction. (author) [French] Apres avoir passe en revue les differents processus d'interaction des rayonnements, existant dans les reacteurs nucleaires, avec la matiere, l'auteur decrit les differentes methodes calorimetriques qui peuvent etre utilisees pour mesurer l'energie absorbee dans les materiaux. II presente ensuite en detail le calorimetre adiabatique, le dispositif de mesure associe et les methodes d'etalonnage qui ont ete retenus. Enfin il donne des valeurs d'echauffement dans divers emplacements experimentaux des piles EL-2 et EL-3 pour differents materiaux d'utilisation courante dans les reacteurs. (auteur)

New Methods and Facilities for the Measurement of Physical Properties of Reactor Components and Irradiated Materials; Nouveaux Procedes et Instruments de Mesure des Proprietes Physiques des Elements de Reacteur et des Matieres Irradiees; Novye metody i sredstva izmereniya fizicheskikh s vojstv komponentov reaktora i obluchennykh materialov; Nuevos Metodos y Equipos para Medir Propiedades Fisicas de Componentes de Reactor y de Materiales Irradiados

Energy Technology Data Exchange (ETDEWEB)

Foerster, F.; Mueller, P. [Institut Dr. Foerster, Reutlingen, Federal Republic of Germany (Germany)

1965-09-15

zone 'chaude ' du reacteur. Ils discutent la relation entre la conductivite electrique et la dose d'irradiation. Les auteurs decrivent un instrument de mesure de la permeabilite, de la remanence et de la force co- ercitive en fonction des contraintes mecaniques, de la deformation elastique et inelastique et de la dose d'irradiation. Ils donnent des mesures de la variation des proprietes magnetiques en fonction des contraintes elastiques et de la deformation inelastique. Ils etudient les effets de l'irradiation sur la permeabilite et sur la force coercitive. Les auteurs decrivent un instrument permettant la mesure rapide et la lecture directe de la permeabilite des elements en acier inoxydable. Ils expliquent la correlation entre la permeabilite et la teneur en ferrite {Delta}. Us discutent certaines mesures du pourcentage de ferrite {Delta} dans les soudures de tubes en acier inoxydable ainsi que certaines mesures de precipitation de ferrite {Delta} en fonction de la deformation inelastique (forgeage a la main d'elements combustibles pour reacteurs). (author) [Spanish] Se describe un intrumento para medir y registrar en forma totalmente automatica el modulo de Young, el modulodecorte y la capacidad de amortiguamiento, en funcion de la temperatura y el tiempo. El modulo de Young se determina excitando muestras de diversos tamanos con sus frecuencias naturales, mientras que la capacidad de amortiguamiento se mide en funcion de la libre atenuacion de la vibracion, o bien por la anchura media de la curva de resonancia. Se presentan ejemplos de medidas de la recuperacion despues de provocar danos por irradiaciones y deformaciones plasticas asf como grado de grafitacion. Se describe la deteccion de fallas y variaciones de densidad en barras de grafito. Se explica, ademas, un metodo para investigar la retencion de pastillas de UO{sub 2} en tubos austenfticos de pared delgada. Se describe un horno especial para estudiar el comportamiento elastico e inelastico de muestras

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

Radiolysis of some aqueous solutions of neutron absorbers; Etude des effets de certains absorbeurs de neutrons en solution sur la radiolyse de l'eau

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-15

The initial yield of molecular hydrogen formed by radiolytic decomposition of water in reactor and {sup 60}Co gamma radiation is decreased by the presence of salts of polyvalent elements possessing only one stable valence, i.e cadmium, zinc, magnesium, gadolinium. This effect is favourable for the use of cadmium and gadolinium as soluble neutron absorber in heavy water reactors. Cations of these salts are not inert toward the primary products of water radiolysis. They have a high degree of reactivity toward the hydrated electron, which is the precursor of molecular hydrogen in neutral or alkaline aqueous media. The value of the rate constant for the reaction between cadmium ion and hydrated electron was shown to be (6.1 {+-} 1.8) 10{sup 10} M{sup -1} s{sup -1}. Boric acid at low concentration has no effect on the radiation chemistry of water. An isotope effect has been found in the radiolysis of heavy water, corresponding to a lowering of initial yield [G{sub 0}(D{sub 2}) < G{sub 0}(H{sub 2})]. additionally it was necessary to determine the influence of organic impurities, remaining after the purification of water, on the mechanism of its radiolytic decomposition. (author) [French] Le rendement initial de la formation d'hydrogene moleculaire dans la decomposition radiolytique de l'eau, sous l'effet du rayonnement des reacteurs nucleaires ou du cobalt 60, est diminue si le solute est un sel d'element polyvalent ne possedant qu'un seul etat stable de valence (cadmium, zinc, magnesium, gadolinium). Cet effet est favorable au choix des elements cadmium et gadolinium pour servir d'absorbeur soluble de neutrons dans un reacteur a eau lourde. Les cations de ces sels ne sont pas inertes vis-a-vis des produits primaires de la radiolyse. Ils ont une affinite notable pour l'electron solvate, precurseur de l'hydrogene moleculaire en milieu neutre ou alcalin. En particulier, la constante de vitesse de la reaction du cadmium ionise avec l'electron solvate a pu etre calculee. Sa

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

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.

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.

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.

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)

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)

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)

Neutron thermalization in absorbing infinite homogeneous media: theoretical methods; Methodes theoriques pour l'etude de la thermalisation des neutrons dans les milieux absorbants infinis et homogenes

Energy Technology Data Exchange (ETDEWEB)

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

1963-11-15

After a general survey of the theory of neutron thermalization in homogeneous media, one introduces, through a proper formulation, a simplified model generalizing both the Horowitz model (generalized heavy free gas approximation) and the proton gas model. When this model is used, the calculation of spectra is reduced to the solution of linear second order differential equations. Since it depends on two arbitrary functions, the model gives a good approximation of any usual moderator for reactor physics purposes. The choice of these functions is discussed from a theoretical point of view; a method based on the consideration of the first two moments of the scattering law is investigated. Finally, the possibility of discriminating models by using experimental informations is considered. (author) [French] Apres un passage en revue de generalites sur la thermalisation des neutrons dans les milieux homogenes, on developpe un formalisme permettant de definir et d'etudier un modele simplifie de thermaliseur. Ce modele generalise l'approximation proposee par J. HOROWITZ (''gaz lourd generalise'') et comporte comme cas particulier le modele ''hydrogene gazeux monoatomique''. Il ramene le calcul des spectres a la resolution d'equations differentielles lineaires du second ordre. Il fait intervenir deux fonctions arbitraires, ce qui lui permet de representer les thermaliseurs usuels de facon satisfaisante pour les besoins de la physique des reacteurs. L'ajustement theorique de ces fonctions est discute; on etudie une methode basee sur la consideration des deux premiers moments de la loi de diffusion. On envisage enfin la possibilite de discriminer les modeles d'apres des renseignements d'origine experimentale. (auteur)

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)

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)

Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

internal can of the annular element, has necessitated very much research work. - the exact temperature drop at the contact between the uranium and the can, and the strength of the lower end of the cartridge are points which are increasingly crucial in the case of the annular element. All in all the annular element thus calls for a great research effort. This effort is justified by the big step forwards in which it will result in the case of the EDF reactors thanks to its high specific power and to the high weight of uranium in each cartridge. (authors) [French] La politique choisie en France pour le developpement des elements combustibles destines aux reacteurs de l'Electricite de France, consiste a chercher, pour chaque pile nouvelle, a beneficier au maximum des progres techniques les plus recents en etudiant chaque fois un nouvel element combustible permettant une puissance par canal aussi elevee que possible. Les derniers elements combustibles ainsi etudies par le Commissariat a l'Energie Atomique sont de deux types differents: un element a tube d'uranium ferme aux deux extremites et refroidi exterieurement (ce type d'element, retenu pour les reacteurs EDF 2, EDF 3 et EDF 4 permet des puissances specifiques maximum de l'ordre de 6 MW/t). Un element a tube d'uranium ouvert, refroidi interieurement et exterieurement, appele clemont annulaire et dont on etudie la possibilite pour EDF5. Un tel element peut permettre des puissances specifiques superieures a 12 MW/t. Ces deux types d'elements possedent des caracteristiques communes: la gaine, pour le refroidissement externe, comporte des ailettes en chevron. Ce type de profil, qui a recu recemment des ameliorations notables augmentant son efficacite thermique, a l'avantage important d'eviter les vibrations de cartouche mais a pose des problemes technologiques de tenue au cyclage thermique qui ont necessite une etude approfondie. les cartouches sont placees a l'interieur de chemise en graphite, ce qui limite les efforts

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

Reactivity coefficients by perturbation theory; Calcul des coefficients de re activite par la theorie des perturbations; Koehffitsienty reaktivnosti, opredelennye pri pomoshchi teorii vozmushchenij; Determinacion, de coeficientes de reactividad con ayuda de la teoria de las perturbaciones

Energy Technology Data Exchange (ETDEWEB)

Webster, J W [International Atomic Energy Agency, Vienna (Austria)

1962-03-15

lethargie, en procedant par analogie a partir de l'equation differentielle a un groupe relative au flux adjoint. b) Montre comment appliquer la forme a deux groupes de la theorie des perturbations au cas d'un reacteur surgenerateur a neutrons rapides, refroidi au mercure. Lors de l'essai preliminaire du reacteur, on a constate que la variation de reactivite accompagnant une variation de densite du mercure est telle que le coefficient net de la reactivite est negatif pour certaines regions et positif pour d'autres. Il est negatif pour les regions de poids statistique le plus eleve et ou une variation de puissance entrainerait le changement de densite le plus important. Le coefficient global de densite du mercure est donc negatif, et par consequent le coefficient cavitaire est positif, ce qui est dangereux. On peut facilement voir, en employant la forme a deux groupes, quelles modifications doivent etre apportees aux plans du reacteur etudie pour obtenir un coefficient cavitaire negatif. Au cours de recherches ulterieures, il s'est revele possible d'apporter certaines de ces modifications, et l'on est finalement parvenu a etablir des plans tels que le coefficient cavitaire du reacteur soit negatif. (author) [Spanish] Las formulas de la teoria de las perturbaciones pueden establecerse empleando uno de los procedimientos mas importantes de la heuristica matematica, que consiste en partir de una expresion sencilla para llegar por analogia a una expresion mas compleja. La presente memoria: (a) Formula la teoria de las perturbaciones como metodo para calcular los coeficientes de reactividad. Consiste principalmente en desarrollar la ecuacion diferencial del flujo adjunto como funcion continua de la posicion y del letargo, procediendo por analogia a partir de la ecuacion diferencial de un solo grupo para el flujo adjunto. (b) Presenta una aplicacion de la forma de dos grupos de la teoria de las perturba ciones a un reactor reproductor rapido, refrigerado por mercurio hirviente

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)

Regularities in the Changes of Absorber Material Properties as a Function of Absorber Concentration; Regularite des Variations des Proprietes des Substances Absorbantes en Fonction de la Concentration de l'Absorbant; Zakonomernosti izmeneniya svojstv poglashchayushchikh materialov v zavisimosti ot kontsentratsii poglotitelya; Leyes de Variacion de las Propiedades de los Materiales Absorbentes en Funcion de la Concentracion del Absorbente

Energy Technology Data Exchange (ETDEWEB)

Portnoj, K. I.

1964-06-15

The paper presents regularities of the change in mechanical and heat-physical properties as well as in absorption capability as a function of absorber concentration for thermal and intermediate reactors. The thermal conductivity and the thermal expansion coefficient of absorber alloys containing boron and rare-earth element oxides is reduced with an increase of absorber concentration. Alloys with rare-earth element oxides have a linear law of the thermal expansion coefficient change, while for boron containing alloys this additive law of changes of properties is disturbed. This is caused by formation under high temperatures of boride phases with various crystal lattices. It is shown in the paper that absorption capability, being a function of absorber concentration, is changed along a curve with saturation and depends on the neutron spectrum. A hypothesis of the author on formation of absorption capability maximum under mutual alloying of absorbers is set forth. The hypothesis has got a wide experimental confirmation on a large number of metal and non-metal absorber system compositions in thermal and intermediate reactors. (author) [French] Le memoire expose la regularite des variations des proprietes mecaniques et thermiques ainsi que du pouvoir absorbant en fonction de la concentration de l'absorbant dans les reacteurs a neutrons thermiques et intermediaires. La conductibilite thermique et le coefficient de dilatation thermique des combinaisons absorbantes contenant du bore et des oxydes de terres rares diminuent a mesure qu'augmente la concentration de l'absorbant. Pour les combinaisons qui contiennent des oxydes de terres rares, la variation du coefficient de dilatation thermique est regie par une loi lineaire. Dans le cas des combinaisons contenant du bore, cette loi de variation des proprietes n'est pas rigoureusement applicable, du fait de la formation, a haute temperature, de phases 'borare' avec divers reseaux cristallins. Le memoire demontre que le

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)

Ultrasonic Water-Gap Measurements in MTR Fuel Elements; Mesure par Ultrasons des Espaces Intercalaires dans les Elements Combustibles des Reacteurs d'Essai de Materiaux; Izmereniya vodyanogo zazora v teplovydelyayushchikh ehlementakh dlya materialovedcheskogo reaktora s pomoshch'yu ul'trazvuka; Medicion Ultrasonica de la Capa de Agua en Elementos Combustibles para Reactores de Ensayo de Materiales

Energy Technology Data Exchange (ETDEWEB)

Deknock, R. [Metallurgy Department, S.C.K./C.E.N., Mol (Belgium)

1965-10-15

generalement eleves, il est indispensable de prevoir un transfert de chaleur uniforme et un refroidissement regulier empechant toute formation massive de vapeur. En outre, pour determiner le gonflement et le comportement general du combustible dans le reacteur, il faudra mesurer les espaces intercalaires dans les elements combustibles au cours de controles apres irradiation. A cette fin, on a mis au point une sonde fondee sur le principe des ultrasons, qui permet de mesurer les espaces intercalaires de 3 mm sur 1 m de long dans les elements combustibles du reacteur BR-2. Lorsqu'on procede a des experiences apres irradiation, la sonde doit pouvoir fonctionner dans un element combustible immerge dans un reservoir d'eau a une profondeur de 6 m au minimum. La sonde peut resister a une immersion prolongee dans l'eau et n'est pas endommagee par une irradiation gamma a des doses normales. Bien que le systeme soit fonde sur la methode classique de la reflexion des impulsions, il permet de separer les impulsions emises des impulsions reflechies au moyen d'un cristal ferroelectrique de 10 MHz a pouvoir eleve de dispersion de l'energie. Les resultats des mesures peuvent etre lus directement sur un oscilloscope: le temps est indique sur l'axe horizontal et la vitesse d'exploration est reglee de maniere a se trouver en relation directe avec la vitesse de propagation de l'onde, c'est-a-dire avec la distance intercalaire. Ce mode de lecture est satisfaisant lorsqu'on procede a un nombre limite de mesures, mais il est evidemment preferable d'enregistrer les resultats sur un graphique. Dans ce cas, les impulsions incidentes et les impulsions reflechies sont transmises a un convertisseur temps-tension au moyen d'un circuit logique transistorise. Cet appareil permet un ajustement continu du zero de sortie pour toute distance intercalaire choisie arbitrairement entre 2 et 4 mm, grace a quoi on peut obtenir un enregistrement autour d'un axe zero. En outre, toute variation de 100 {mu}m de la

Experimental Determination of the Neutron Characteristics of UO{sub 2}-PuO{sub 2}-H{sub 2}O Lattices; Determination Experimentale Des Caracteristiques Neutroniques De Reseaux UO{sub 2}-PuO{sub 2}-H{sub 2}O

Energy Technology Data Exchange (ETDEWEB)

Debrue, J.; Fabry, A.; Leenders, L.; Motte, F.; Van Den Broeck, H. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

1967-09-15

plutonium: fission density measurements by direct counting of the fission products formed in the mixed fuel as a result of irradiation, and measurements of spectral indices by means of fission detectors and detectors having resonances close to those of the isotopes of plutonium. The optimum conditions for applying these techniques are described and the effect on the fuel cycle of uncertainties in the values for the spectral indices (ratios of cross-sections) is discussed. A brief reference is finally made to a programme of sub-critical experiments conducted hitherto with two batches of UO{sub 2} fuel enriched to 5% and 7% {sup 235}U, and having the purpose of studying possible methods of exciting a sub-critical lattice by means of an external neutron source in pulsed and modulated operation. The aim of this programme is to ascertain to what extent these methods, which employ a quantity of fuel considerably less than the critical mass, can help in determining the characteristics of a lattice. The advantages of such methods are particularly obvious in the case of plutonium fuels. (author) [French] Au cours de l'etude dans l'installation critique VENUS du coeur a moderation variable du reacteur BR3/VULCAIN, un assemblage combustible constitue de 37 crayons UO{sub 2}-PuO{sub 2} (94% UO{sub 2} naturel, PuO{sub 2}) a ete substitue a un des assemblages combustibles UO{sub 2} enrichi a 7% {sup 235}U constituant le coeur de ce reacteur. Des experiences ont ete effectuees dans le but de permettre l'ajustement des modeles du calcul des performances de cet assemblage special: on a notamment mesure la distribution de la densite de fission ainsi que l'evolution du rapport des sections efficaces effectives de fission dans {sup 239}Pu et {sup 235}U. Une experience critique directement orientee sur les problemes de recyclage du plutonium dans les reacteurs thermiques a eau legere pressurisee est en cours d'execution dans la meme installation critique. A cette fin, on dispose de deux

Group cross-sections for fast reactors; Sections efficaces de groupes pour les reacteurs a neutrons rapides; Gruppovye secheniya reaktorov na bystrykh nejtronakh; Secciones eficaces de grupos para reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Zweifel, P P [University of Michigan, Ann Arbor, MI (United States); Ball, G L [Atomic Power Development Associates, Inc., Detroit, MI (United States)

1962-03-15

groupes. Ils montrent notamment que la section efficace moyenne de transport peut, avec une certaine approximation, s'exprimer en termes de libre parcours moyen. Le calcul de cette quantite prend beaucoup de temps, car elle ne peut se reduire en moyennes elementaires; neanmoins, on a demontre certaines inegalites, qui simplifient la methode de calcul des moyennes qui doit etre utilisee. Les auteurs analysent trois autres aspects des sections efficaces de groupes, que l'on neglige souvent, mais qu'il peut etre important de connaitre pour les etudes de reacteurs. a) Il est injustifie d'utiliser pour tous les reacteurs a neutrons rapides le meme ensemble de sections efficaces dont la moyenne par groupe a ete calculee si les spectres des differents reacteurs sont dissemblables et si les sections efficaces varient rapidement a l'interieur du groupe, comme c'est le cas le plus souvent. Les auteurs decrivent une methode d'iteration qui permet d'obtenir les valeurs moyennes correctes; ils determinent ensuite, a l'aide de cette methode, dans quelle mesure les calculs de reacteurs sont influences par les effets de spectre. b) Dans les calculs de transport (la methode S{sub n}, par exemple), les moyennes doivent etre calculees en tenant compte a la fois de tous les angles et de toutes les energies. Etant donne qu'on ne peut dissocier dans le flux une partie angulaire et une partie energetique, la plus grande attention est necessaire pour eviter les erreurs. Les auteurs etudient l'equation obtenue par la methode S{sub n} sous la forme d'un modele simple, et en tirent un critere qui pourrait aider a determiner l'importance de la non-separabilite angulaire dans les calculs de reacteurs. c) A partir de raisonnements fondes sur la conservation du nombre de neutrons, il est possible d'obtenir une relation consistant entre les coefficients de diffusion de groupe, le pouvoir de ralentissement et les sections efficaces d'absorption. Les auteurs montrent qu'il n'est pas exact de definir

Forming Limit Diagrams of Zircaloy-4 and Zirlo Sheets for Stamping of Spacer Grids of Nuclear Fuel Rods

International Nuclear Information System (INIS)

Seo, Yun Mi; Hyun, Hong Chul; Lee, Hyung Yil; Kim, Nak Soo

2011-01-01

In this work, we investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile and anisotropy tests were performed to obtain stress-strain curves and anisotropic coefficients. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following NUMISHEET 96. Theoretical FLD depends on FL models and yield criteria. To obtain the right hand side (RHS) of FLD, we applied the FL models (Swift's diffuse necking, M-K theory, S-R vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left hand side (LHS) of FLD. To consider the anisotropy of sheets, the yield criteria of Hill and Hosford were applied. Comparing the predicted curves with the experimental data, we found that the RHS of FLD for Zircaloy-4 can be described by the Swift model (with the Hill's criterion), while the LHS of the FLD can be explained by Hill model. The FLD for Zirlo can be explained by the S-R model and the Hosford's criterion (a = 8)

Study and Construction of the Metal Vessels for the Reactors of the EDF1 and EDF2 Sectors at Chinon; Etude et construction des caissons metalliques des reacteurs des tranches EDF1 et EDF2 de la centrale de Chinon; Izuchenie i konstruktsiya metallicheskikh korpusov reaktorov pervoj i vtoroj chasti programm ehlektrostantsij; Estudio y construccion de los recipientes metalicos de los reactores EDF1 y EDF2 de la central de Chinon

Energy Technology Data Exchange (ETDEWEB)

Lamiral, G.; Millot, R.; Passerieux, P. [Electricite de France, Clamart, Seine (France)

1963-10-15

The first two natural uranium-graphite-C0{sub 2} reactors at the Chinon station have metal vessels of thick manganese-molybdenum steel plate. The studies carried out on these vessels raised certain problems, particularly in connection with the design and dimensions of the port reinforcements. The reinforcements for the control-rod channels and fuel ports were studied on mock-ups and the results obtained were checked on the completed reactors during hydraulic tests. The type of construction initially used for the EDF1 vessel was relatively simple. The plates to be welded were locally preheated, and the vessel was not supposed to undergo more than one stress-relief heat treatment after completion of all the welding. Serious cracks developed, however, and it became necessary to alter the whole method of construction. In particular, the welding was now done after overall preheating and the vessel was subjected to multiple stress-relief treatments. This made it possible to fabricate the vessels for EDF1 and EDF2, but at the same time imposed certain limitations which considerably complicated work on the site. (author) [French] Les reacteurs a uranium naturel, graphite et gaz carbonique des deux premieres tranches de la Centrale de Chinon comportent des caissons metalliques realises a partir de toles de fortes epaisseurs, en acier au manganese-molybdene. Les etudes de ces paissons ont pose certains problemes, notamment en ce qui concerne les renforts d'ouvertures. Les renforts des passages des barres de controle et des orifices de chargement ont ete etudies sur maquette et les resultats obtenus ont ete controles sur les ouvrages termines lors des epreuves hydrauliques. Le mode de construction initialement utilise pour le caisson de la tranche EDF1 etait relativement simple; les toles a souder etaient prechauffees localement et le caisson ne devait subir qu'un seul traitement thermique de detente, apres execution de toutes les soudures. Une fissuration importante en cours

Critical heat flux on micro-structured zircaloy surfaces for flow boiling of water at low pressures