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Sample records for reacteurs industriels refroidis

  1. 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

  2. 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

  3. Containment for Heavy-Water Gas-Cooled Reactors; Le Confinement des Reacteurs a Eau Lourde Refroidis par Gaz

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    Verstraete, P.; Lehmann, D.; Lafitte, R. [Bonard et Gardel, Ingenieurs-Conseils, Lausanne (Switzerland)

    1967-09-15

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author) [French] Les principes de securite des reacteurs a eau lourde refroidis

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

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    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.)

  5. 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)

  6. 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

  7. Detection and location of can rupture in reactors cooled by a flow of water; Detection et localisation des ruptures de gaines sur les reacteurs refroidis par circulation d'eau

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

    1968-07-01

    This report brings together the principal methods of fission-product detection used for water reactors. The position, type and method of adjustment is given for each detector. The methods for localizing the defective elements are explained, in particular those using water sampling or decreases in the flux. A few installations are briefly described. They correspond to particular types of reactors using boiling, pressurized or cold water. Amongst the many methods used, it can be noted that when the fuel is resistant, the installations are fairly compact. In nuclear super-heated reactors on the other hand, the study of fuel behaviour calls for larger installations. An identification of defective elements exists when the reactor structure allows it. If this is not possible, a localization in a group of elements is obtained by a flux depression. (author) [French] Ce rapport rassemble les principales methodes de detection de produits de fission utilisees pour des reacteurs a eau. On indique pour les detecteurs leurs emplacements, leurs types, leurs reglages. On explique quelles sont les methodes de localisation des elements defectueux, en particulier celles utilisant des prelevements d'eau ou des depressions de flux. Quelques installations sont decrites sommairement. Elles correspondent a des types particuliers de reacteurs a eau bouillante, pressurisee ou froide. Parmi les nombreuses methodes utilisees, on constate que les installations sont peu importantes, lorsque le combustible est resistant. Par contre dans les reacteurs a surchauffe nucleaire l'etude du comportement du combustible necessite des installations plus importantes. Une identification d'elements defectueux existe lorsque la structure du reacteur le permet. A defaut une localisation dans un groupe d'elements est obtenue par depression de flux. (auteur)

  8. 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

  9. Development of Non-Metallic Fuel Elements for a High-Temperature Gas-Cooled Reactor; Mise au point d'elements combustibles non metalliques pour un reacteur a haute temperature, refroidi par un gaz; Razrabotka nemetallicheskikh teplovydelyashchikh ehlementov dlya vysokotemperaturnogo reaktora s gazovym okhlazhdeniem; Elementos combustibles no metalicos para un reactor de temperatura elevada refrigerado por gas

    Energy Technology Data Exchange (ETDEWEB)

    Liebmann, B.; Schafer, L.; Spener, G. [NUKEM, Nuklear-Chemie und -Metallurgie G.m.b.H., Wolfgang bei Hanau, Federal Republic of Germany (Germany)

    1963-11-15

    In connection with fuel element development work for the high-temperature gas-coolcd reactor of the Brown-Boveri/Krupp Reaktorbau G.m.b.H., two different fuel element concepts were considered and developed. In both cases the fuel element consists of a graphite ball of 6 cm in diam. which contains the fuel insert, a cylindrical pellet of about 20 mm in diam. and 16 mm in height. The two concepts differ in the type of the.fuel insert as well as in the preparation of the graphite ball. In the first concept the fuel insert consists of a mixture of UC{sub 2} and graphite which is prepared by blending U{sub 3}O{sub 8} and graphite, pressing them into pellets and reacting the two components in a vacuum furnace at 1800{sup o}C. The atomic ratio of U : C is 1:45. Since this type of fuel pellet does not retain the fission products completely the surrounding graphite sphere had to be made impervious to fission products by impregnation in order to obtain a fission-product retaining element. Permeabilities of the order of 10{sup -6}cm{sup 2}/s could be achieved. In the second concept the fuel insert consists of a solid solution of UC in ZrC and is coated with a layer of ZrC. The molar ratio of UC to ZrC is 1 : 20. The fuel pellet preparation was accomplished by the following procedure: UO{sub 2}, ZrO{sub 2}, and graphite were mixed and pressed into pellets. The pellets were reacted to the carbides. Ball milling of the carbides was followed by hot pressing at temperatures o f 2000{sup o}C. Densities of more than 95% of the theoretical density could be achieved. A full description of the preparation and of some physical properties of the fuel pellets is given in the paper. A sufficient fission gas retention behaviour of this type of fuel insert which allows it to be put into unimpregnated graphite balls is expected. Other advantages of this kind of fuel are discussed. (author) [French] Dans le cadre des etudes de combustibles destines au reacteur a haute temperature, refroidi par

  10. 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)

  11. Industrial Ultrasonic Inspection of Stainless-Steel Claddings for the EL4 Reactor; Controle Industriel par Ultrasons des Gaines en Acier Inoxydable du Reacteur EL4; Promyshlennyj kontrol' obolochechnykh trub iz nerzhaveyushchej stali reaktora dlya EL4 s pomoshch'yu ul'trazvukovogo metoda; Metodos Ultrasonicos para Control Industrial de las Vainas de Acero Inoxidable del Reactor EL4

    Energy Technology Data Exchange (ETDEWEB)

    Prot, A. C.; Foulquoer, H. E.; Peyrot, J. P. [Centre d' Etudes Nucleaires de Saclay (France)

    1965-09-15

    essentiel de rentabilite du reacteur. Il s'avere que le choix de la methode a utiliser et a mettre au point est delicat; le memoire en donne les elements essentiels. Ce choix etant fait, apres mise au point en laboratoire, deux nouveaux problemes se posent: - la transposition dans le domaine industriel; - la necessite de tenir compte de la qualite permise, a un instant determine, par les procedes de fabrication, en relation avec les normes de reception definies de maniere plus ou moins arbitraire. Ceci se traduit en fait par la necessite d'une etude statistique sur des lots de tubes de diverses provenances, et leur classement par rapport a des seuils plus ou moins severes. On verra que le nombre de tubes a controler est tres superieur a celui prevu initialement. Cela conduit a l'etude d'une machine de controle automatique, capable de satisfaire a la fois les exigences de cadence et celles propres au type de controle choisi: ces dernieres sont generalement d'ordre mecanique et necessitent une construction particulierement soignee. L'ensemble de ces considerations a conduit a concevoir une machine dont la cadence peut des maintenant couvrir sans difficulte les besoins d'une chaihe de fabrication d'elements combustibles. Les possibilites de cette machine sont etroitement liees aux caracteristiques du materiel de controle choisi, en particulier aux performances de l'electronique des appareils de controle par ultrasons et a celles des traducteurs utilises. Il resulte d'ailleurs de cette etude que le materiel standard ne repond que tres imparfaitement au probleme et que l'on doit envisager des maintenant un appareillage particulier pour ce type de controle. (author) [Spanish] Las mayores exigencias a que se someten los reactores obligan a utilizar materiales elaborados y controlados con sumo cuidado. Un aspecto de tal control se refiere a la calidad de las vainas empleadas, cuyas propiedades mecanicas ejercen una influencia decisiva sobre la rentabilidad del reactor. La eleccion

  12. Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR); Developpement du design d'un assemblage de controle et analyse dynamique des reacteurs a neutrons rapides de quatrieme generation refroidis au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Girardin, G.

    2009-07-09

    modeles neutroniques 2D et 3D du coeur du reacteur ont ete crees, bases sur le schema de calculs de reference ERANOS-2.0/ERALIB1. Pour l'analyse thermo-hydraulique, le code COPERNIC du CEA a ete utilise. Le travail de design a ete poursuivi par l'etude d'un schema de l'implantation des assemblages de controle (nombre et position dans le coeur). Des etudes detaillees de neutronique ont reveles l'existence de grands effets d'interaction entre les AC, appeles effets d'ombre/d'anti-ombre, conduisant a une amplification/reduction de l'antireactivite des AC. Les interactions entre les barreaux absorbants a l'interieur d'un AC, ainsi qu'entre les AC eux-memes, ont ete investiguees dans le detail, dans le but d'optimiser l'efficacite des AC (en terme de la fraction d'absorbant et la minimisation des effets d'heterogeneite associes). Resultant d'investigations detaillees, le diametre des pastilles absorbantes a ete choisi de maniere a minimiser l'influence 'barreau-a-barreau' a l'interieur de l'assemblage. En particulier, une partie centrale de l'assemblage a ete concue sans aucun barreau absorbant (zone remplie d'helium statique). Par ce biais, une reduction, a 13%, des effets d'heterogeneite, a ete obtenue. Les investigations neutroniques effectuees pour le coeur RNR-G de reference ('2004-Coeur'), specialement, celles liees a l'Etude des interactions entre les AC, ont directement contribue au nouveau design du coeur ('2007-Coeur'). Le rapport hauteur sur diametre a ete augmente a 0.6, compare a la valeur de 0.3 pour le coeur de reference. Pendant la troisieme phase, des modeles couples et detailles, cinetiques 3D et thermohydrauliques 1D, ont ete developpes pour le coeur RNR-G; le but etait d'arriver a une comprehension, en profondeur, du comportement 3D du coeur pendant des transitoires induits par le mouvement d

  13. Heavy water reactors physics; Physique des reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

  14. High-Temperature Gas-Cooled Reactor Critical Experiment and its Application to Thorium Absorption Rates; Experience Critique pour l'Etude d'un Reacteur a Haute Temperature, Refroidi par un Gaz et son Application a la Determination des Taux d'Absorption du Thorium; Kriticheskij opyt, postavlennyj na vysokotemperaturnom reaktore s gazovym okhlazhdeniem, i primenenie ego dlya opredeleniya stepeni pogloshcheniya toriya; Experimento Critico Efectuado en un Reactor de Elevada Temperatura Refrigerado por Gas y su Aplicacion para Calcular los Indices de Absorcion del Torio

    Energy Technology Data Exchange (ETDEWEB)

    Bardes, R. G.; Brown, J. R.; Drake, M. K.; Fischer, P. U.; Pound, D. C.; Sampson, J. B.; Stewart, H. B. [General Dynamics Corporation,San Diego, CA (United States)

    1964-04-15

    the fact that the thorium is dispersed in graphite and the usual cadmium-ratio technique is difficult to apply. Comparison of experimental and theoretical results shows excellent agreement over a range of variables. In addition, the results of both activation and reactivity measurements of Doppler coefficient are in agreement, a fact which is felt to be significant in view of the disparity between results from these two techniques in the literature. (author) [French] Lors de l'etude du reacteur HTGR a haute temperature refroidi par un gaz, et de son premier prototype a Peach Bottom, la General Atomic Division de la societe General Dynamics a decide qu'il fallait proceder a une experience critique pour obtenir certaines donnees d'entree necessaires pour l'analyse nucleaire. Aux fins de l'etude nucleaire theorique, les besoins particuliers en donnees d'entree relatives aux absorptions par le thorium ont amene les ingenieurs a concevoir un assemblage experimental critique compose d'un reseau central entoure d*une region tampon et d'une region de commande. Ce type.d'assemblage, dans lequel on peut creer le spectre a mesurer dans le reseau central relativement petit ayant la geometrie voulue, permet d'obtenir des donnees d'entree tres diverses pour les etudes de projets nouveaux, au point de vue de l'analyse nucleaire. Le memoire indique les avantages particuliers que presente cette methode par rapport a celle qiu consiste a construire une maquette, ainsi que le role de la theorie pour determiner quelles experiences sont le plus utiles et comment utiliser ensuite ces experiences dans la verification des procedes d'etude. Les auteurs ont mis au point deux methodes relativement nouvelles qui peuvent etre utilisees avec l'assemblage decrit ci-dessus: une methode d'oscillation de la reactivite pour determiner le coefficient Doppler pour le thorium; une methode d'activation pour determiner a la fois l'integrale de resonance pour le thorium disperse dans le graphite et ses

  15. Modelisation et simulation de pyrolyse de pneus usages dans des reacteurs de laboratoire et industriel

    Science.gov (United States)

    Lanteigne, Jean-Remi

    The present thesis covers an applied study on tire pyrolysis. The main objective is to develop tools to allow predicting the production and the quality of oil from tire pyrolysis. The first research objective consisted in modelling the kinetics of tires pyrolysis in a reactor, namely an industrial rotary drum operating in batch mode. A literature review performed later demonstrated that almost all kinetics models developed to represent tire pyrolysis could not represent the actual industrial process with enough accuracy. Among the families of kinetics models for pyrolysis, three have been identified: models with one single global reaction, models with multiple combined parallel reactions, and models with multiple parallel and series reactions. It was observed that these models show limitations. In the models with one single global reaction and with multiple parallels reactions, the production of each individual pyrolytic product cannot be predicted, but only for combined volatiles. Morevoer, the mass term in the kinetics refers to the final char weight (Winfinity) that varies with pyrolysis conditions, which yields less robust models. Also, despite the fact that models with multiple parallels and series reactions can predict the rate of production for each pyrolysis product, the selectivities are determined for operating temperatures instead of real mass temperatures, giving models for which parameters tuning is not adequate when used at the industrial scale. A new kinetics model has been developed, allowing predicting the rate of production of noncondensable gas, oil, and char from tire pyrolysis. The novelty of this model is the consideration of intrinsic selectivities for each product as a function of temperature. This hypothesis has been assumed valid considering that in the industrial pyrolysis process, pyrolysis kinetics is limiting. The developed model considers individual kinetics for each of the three pyrolytic products proportional to the global decomposition kinetics of pyrolysables. The simulation with data obtained in industrial operation showed the robustness of the model to predict with accuracy in transient regime, tires pyrolysis, with the help of model parameters obtained at laboratory scale, namely in regards of the trigger of production, the residence time of tires (dynamic production) and the amount of oil produced (cumulative yield). It is a novel way to model pyrolysis that could be extrapolated to new waste materials. The second objective of this doctoral research was to determine the evolution of specific tires specific heat during pyrolysis and the enthalpy of pyrolysis. The origin of this objective comes from a primary contradiction. With few exceptions, it is acknowledged that organic materials pyrolysis is globally an endothermic phenomenon. At the opposite, all experiments led with laboratory apparatuses such as DSC (Differential Scanning Calorimetry) showed exothermic peaks during dynamic experiments (constant heating rate). It has been confirmed by results obtained at the industrial scale, where no sign of exothermicity has been observed. The Hess Law has also confirmed these results, that globally, pyrolysis is indeed a completely endothermic process. An accurate energy balance is required to predict mass temperature during pyrolysis, this parameter being unbindable from kinetics. An advanced investigation of char first allowed demonstrating that specific heat of solids during pyrolysis decreases with increasing temperature until the weight loss peak is reached, around 400°C, and then starts increasing again. This observation, combined with the fact that the sample loses weight during pyrolysis is considered as the major cause of the apparition of an exothermic peak in laboratory scale experiments. That is, the control system of these apparatuses generates a bias and an unavoidable overheat of the samples producing this exothermic behavior. It would thus be an artifact. On the base of new data on the evolution of global specific heat during pyrolysis, a model of the energy balance has been developed at the industrial scale to determine the enthalpy of pyrolysis. The simulation has shown that a major part of the heat transferred to the pyrolized mass would make its temperature increase. Next, an enthalpy of pyrolysis dependent of weight loss was obtained. Finally, two other terms of enthalpy have been found, namely an enthalpy for the breakage of sulfur bridges and an enthalpy for the stabilization of char when conversion approaches completion. This research will have allowed establishing a novel general methodology to determine the enthalpy of pyrolysis. More particularly, new clarifications hasve been obtained in regards to the evolution of specific heat of solids during pyrolysis and new enthalpies of pyrolysis, all endothermic, could be obtained, in agreement with the theoretical expectations. The third research objective concerned the behavior of sulfur during tires pyrolysis. With as a premise that sulfur is an intrinsic contaminant of many types of waste, it is critical to clarify its fate during pyrolysis, in the present case for waste tires. It has been observed in the literature that some quantitative analyses had been presented, but generally, the mechanisms for the distribution of sulfur within the pyrolytic products remain unclear. Thus, it was then not possible to predict the transfer of sulfur to each of the tire pyrolysis products. The results taken form literature have been complemented with a series of TGA experiments followed by complete elemental analyses of the residual solids. Mass balances have been performed in order to characterize the distribution of elements within the three products (noncondensable gas, oil, and char). A novel parameter has been created during this research: the sulfur loss selectivity. This intrinsic selectivity is a prediction of the distribution of sulfur within the pyrolysis products as a function of temperature. Three phenomena has been identified that could affect the sulfur loss selectivity. First, the natural devolatilization of sulfur due to pyrolysis. Next, the sulfur devolatilization due to the desulfurization of the solid matrix by hydrogen and finally, the clustering of sulfur in the solid state due to metal sulfidation (zinc and iron). The results have shown that this selectivity reach a limit value of 1 when pyrolysis is limited by the kinetics and in the absence of metal. When the mass transfer is limiting at low temperature (<500°C) the selectivity will be greater than 1. At a temperature over 350°C with the presence of metals, the selectivity will be lower than 1. It is a useful tool for industrial pyrolysis processes, being a novel indicator for the distribution of contaminants during the pyrolysis of waste. A better comprehension of these mechanisms allows elaborating a better strategy when designing these industrial processes. For example, in light of this research, it could be preferable to pre-treat the tires at lower temperature to eliminate a significant part of sulfur before pyrolyzing them at high temperature. The resulting pyrolytic products would then necessitate a lighter purification post-treatment, being more efficient and more economical.

  16. Considerations concerning the reliability of reactor safety equipment; Considerations sur la fiabilite des ensembles de securite de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J; Guyot, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    donne les resultats d'une analyse de fiabilite previsionnelle lorsque l'on utilise les elements fonctionnels en boitiers modulaires developpes industriellement en France. L'amelioration de cette fiabilite semble assez limitee par une augmentation de la redondance, par contre, on montre comment elle peut etre tres nettement amelioree par l'emploi de systemes de recherche automatique de pannes a des frequences tres differentes pour les sous-ensembles de mesure et les sous-ensembles logiques. Enfin, on montre sur des exemples l'incidence sur la fiabilite de la complexite et de l'emploi de technologies differentes dans les ensembles de securite de reacteurs. (auteurs)

  17. 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)

  18. Industrial gases. Special issue; Industriele gassen. Thema

    Energy Technology Data Exchange (ETDEWEB)

    De Boer, A.; Voermans, F. (ed.)

    2004-07-01

    In three articles attention is paid to the use of synthesis gas in the Netherlands. In the first article the growing market for synthesis gas is discussed, in the second article it is argued that the energy efficiency of the production of nitrogen can be improved, and the third article is about cooling an ice track by means of carbon dioxide. [Dutch] In 3 artikelen wordt aandacht besteed aan het gebruik van industriele gassen. In het eerste artikel wordt de sterk groeiende markt voor synthese gas besproken, in het tweede artikel wordt aangegeven hoe de efficiency van stikstofproduktie kan worden verbeterd, en het derde artikele gaat in op het gebruik van kooldioxide in de koeling van een ijsbaan.

  19. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    operating power levels of reactor. The regulating system has brought about difficult problems; experimental examination, while operating, will solve them. Special meetings will be held concerning the burst slug system and fuel elements. (author) [French] La construction des reacteurs G2 et G3, dans le cadre du premier plan quinquennal francais, a ete confiee par le C.E.A. au groupement d'industriels FRANCE-ATOME. Bien que ces reacteurs restent essentiellement plutonigenes, on a accole a chacun d'eux une centrale electrique devant fournir 40 MW, dont la responsabilite a ete assumee par l'E.D.F. Le coeur du reacteur adopte la plupart des solutions du reacteur G1 (excepte la fente centrale): canaux horizontaux, empilement de briques parallelepipediques de graphite, protection thermique en acier. Le refroidissement est assure par du gaz carbonique sous 15 atmospheres. Cette pression est tenue par un caisson en beton precontraint, ayant la forme d'un cylindre horizontal. Des cables d'acier sous tension entourent le cylindre de beton, dont ils sont isoles par des patins. Les fonds du cylindre ont pose des problemes particuliers qui ont conduit a la forme hemispherique adoptee. L'etancheite du caisson est assuree par une tole de 30 mm liee a la face interne du beton. Un des aspects les plus originaux de ces reacteurs est la possibilite de charger et decharger en marche. Cote chargement, des sas a barillets, pesant chacun 50 tonnes; permettent de faire passer les cartouches neuves sous la pression de 15 atmospheres. Ces cartouches progressent de facon quasi continue dans le canal pour tomber finalement par des goulottes inclinees et des toboggans helicoidaux dans un nouveau sas. La circulation du gaz carbonique est assuree par trois turbo-soufflantes, actionnees elles-memes par la vapeur moyenne pression obtenue dans echangeurs, chaque reacteur alimente quatre echangeurs ayant pose de difficiles problemes de construction et de mise en place. Le cycle secondaire est un cycle

  20. [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

  1. impacts des rejets industriels sur les eaux souterraines

    African Journals Online (AJOL)

    USER

    pollution et sont nombreuses, notamment les usines sidérurgiques, métalliques et pétrochimiques. Leurs rejets ont des effets néfastes sur les eaux de la région. Actuellement, les effluents industriels situés dans la région de Berrahal, contiennent d'importantes quantités de produits chimiques organiques et inorganiques et ...

  2. 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

  3. 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

  4. Nouvelles approches ergonomiques de la cartographie des risques industriels

    Directory of Open Access Journals (Sweden)

    Éliane Propeck-Zimmermann

    2009-12-01

    Full Text Available Les cartes actuelles présentent des insuffisances pour une gestion territoriale des risques industriels. En redéfinissant le concept de «situation à risques» et en l'implémentant dans un SIG, on peut disposer à la demande d'informations riches et variées: cartes analytiques, de synthèse, typologies, requêtes spatiales diverses. Les recherches en cours visent à développer une interface ergonomique pour faciliter la concertation entre différents acteurs.

  5. 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)

  6. Elaboration et Suivi des Budgets de Marketing Industriel: le Système ADVISOR

    OpenAIRE

    Choffray, Jean-Marie; Delabre, Gilles

    1982-01-01

    Cet article revoit les problèmes posés par l'élaboration et le suivi des budgets de marketing dans un environnement industriel. Il présente le système ADVISOR pour lequel nous avons développé un programme interactif adapté aux besoins des entreprises Françaises.

  7. 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

    specifiques nucleaires et classiques. Ces conditions de rentabilite conduisent a admettre pour les reacteurs ainsi utilises certaines caracteristiques techniques et economiques hors desquelles la competition est improbable. On situe, d'autre part, ces resultats par rapport au marche potentiel de la vapeur et de l'electricite et on est ainsi conduit a examiner certaines utilisations de la chaleur des centrales mixtes telles que l'alimentation de complexes industriels, de divers types de chauffage urbain ou du dessalement des eaux de mer. (auteurs)

  8. 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

    combustibles et des investissements specifiques nucleaires et classiques. Ces conditions de rentabilite conduisent a admettre pour les reacteurs ainsi utilises certaines caracteristiques techniques et economiques hors desquelles la competition est improbable. On situe, d'autre part, ces resultats par rapport au marche potentiel de la vapeur et de l'electricite et on est ainsi conduit a examiner certaines utilisations de la chaleur des centrales mixtes telles que l'alimentation de complexes industriels, de divers types de chauffage urbain ou du dessalement des eaux de mer. (auteurs)

  9. 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

  10. 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

  11. 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

    refroidis par un gaz, construire et exploiter, dans le cadre de ce projet, un reacteur experimental de 20 MWth. Le reacteur - dont la construction touche a sa fin - est refroidi a l'helium; la temperature de sortie du coeur est de 750{sup o}C; il emploie de l'uranium-235 comme combustible et du thorium comme matiere fertile. Il est caracterise par l'absence de tout metal dans le coeur du reacteur. En raison'des hautes temperatures, qui peuvent atteindre 1050{sup o}C a la surface des elements combustibles et s'elever au dessus de 1500{sup o}C dans les regions les plus chaudes du combustible, on emploie des materiaux refractaires non metalliques. Le fait que tous les constituants du coeur sont reunis dans l'element combustible permet d'obtenir un rapport eleve entre la surface de transfert de la chaleur et le volume du coeur, d'ou une puissance specifique moyenne elevee pour un ensemble de dimension relativement faible. Chaque element combustible est constitue par un faisceau de tubes de graphite contenant les matieres fissiles et fertiles sous forme de carbure incorpore a des pastilles de graphite. Un courant refroidisseur d'helium traverse le centre de chaque barreau de combustible d'ou il ressort par la base pour etre conduit dans une installation de purification dans laquelle il est debarrasse des produits de fission et autres impuretes avant d'etre achemine de nouveau vers le reacteur. Cette methode permet de reduire la fuite des produits de fission qui, s'echappant du combustible ceramique porte a tres haute temperature, entrent dans le circuit de refroidissement primaire. Les auteurs exposent les problemes lies a la mise au point et a la fabrication du graphite et des elements combustibles ceramiques destines a ce reacteur ainsi que le comportement de ces materiaux dans les conditions de fonctionnement. Ils indiquent les resultats de recherches en pile et dans des boucles d'irradiation. Dans ce programme, tout l'effort se concentre sur la mise au point de reacteurs a

  12. Draught control by means of industrial air curtains. Background information on the design of industrial air curtains; Tochtbestrijding met industriele luchtgordijnen. Achtergronden bij ontwerp van industriele luchtgrodijnen

    Energy Technology Data Exchange (ETDEWEB)

    Cremers, B. [Biddle, Kootstertille (Netherlands); Traversar, R. [TNO Milieu, Energie en Procesinnovatie TNO-MEP, Apeldoorn (Netherlands)

    2008-02-15

    In industrial buildings a balanced ventilation is not common property. Due to increasing differences in pressure across the front much cold outside air enters the building, resulting in poor comfort near the door. A good air curtain can heat up large amounts of incoming cold outside air in such a way that energy saving is optimal and comfort remains high. (mk) [Dutch] Bij industriele gebouwen is gebalanceerde ventilatie geen gemeengoed. Door oplopende drukverschillen over de gevel komt veel koude buitenlucht naar binnen en is het comfort vlak achter de deur erg laag. Een goed luchtgordijn kan grote hoeveelheden binnenkomende koude buitenlucht zodanig opwarmen dat de energiebesparing optimaal is en het comfort hoog blijft.

  13. 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)

  14. Generative Fertigung im Maschinenbau - industrieller 3D-Druck auf dem Weg in die Serienproduktion

    OpenAIRE

    Müller, Bernhard

    2014-01-01

    3D-Druck ist aktuell medial omnipräsent, sein Potential für echte industrielle Anwendungen, v. a. im Maschinenbauumfeld, wird kontrovers diskutiert. Der Vortrag gibt einen fundierten Einblick in den Stand der Technik zum industriellen 3D-Druck (Generative Fertigung , Additive Manufacturing) und zeigt spezifische Potentiale mit industrieller Relevanz auf. Dabei werden ebenso Praxisbeispiele aus heutiger Anwendung in der Industrie gezeigt als auch Zukunftsszenarien für potentielle Anwendungen e...

  15. DESIGNOR: une méthode nouvelle d'aide à la conception des produits industriels

    OpenAIRE

    Choffray, Jean-Marie

    1980-01-01

    Cet article présente une méthode nouvelle, appelée DESIGNOR, d'aide à la décision marketing. Son objectif est d'accroître la créativité au cours du processus de développement d'un nouveau produit industriel et de réduire les risques d'échec commercial.

  16. Can rupture detector for water cooled piles; Detecteur de rupture de gaine pour piles refroidies a l'eau

    Energy Technology Data Exchange (ETDEWEB)

    Choudens, H de; Guitton, P

    1962-07-01

    The object of this study was to develop a simple, easy to-build, apparatus for showing the appearance of a defect on a fuel element can of a swimming pool reactor. The apparatus used consists of a coil of activated carbon around a NaI(Tl) crystal. Through this coil pass the gases obtained by degassing a sample of water from the reactor; the fission gases which appear when a can leaks are trapped in the carbon; the NaI(Tl) crystal is coupled with a photomultiplier followed by a single-channel selector fixed on a photo-electric peak characteristic of the {gamma} spectrum of fission gases. Preliminary experiments were carried out in laboratory; a more complete device was then built and is now working in the reactor Melusine. (author) [French] Le but de cette etude a ete la realisation d'un appareil simple et facile a realiser destine a indiquer l'apparition d'un defaut sur une gaine des elements combustibles d'une pile piscine. A cet effet, l'appareillage utilise est compose d'un serpentin de charbon active entourant un cristal de NaI (Tl). Ce serpentin est parcouru par les gaz provenant du degazage d'un prelevement d'eau de la piscine du reacteur; les gaz de fission apparaissant lors d'une rupture de gaine sont retenus dans le charbon; le cristal INa (Tl) est couple avec un PM suivi d'un selecteur monocanal cale sur un pic photoelectrique caracteristique du spectre {gamma} des gaz de fission. Des manipulations preliminaires ont ete faites en laboratoire, un dispositif plus complet a ete ensuite monte et fonctionne a la Pile Melusine. (auteurs)

  17. 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

  18. Study of isotopic exchange reactors (1961); Etude des reacteurs d'echange isotopique (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Grandcollot, P; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    A study is made of the general case of the theory of first-order isotopic chemical exchange between a gaseous and a liquid phase in a reactor, starting from fundamental reaction kinetics data, and without making any limiting hypothesis concerning the value of the separation factor. The cases of counter-current reactors and of co-current reactors are considered successively. The general deuterium conservation equation requires the definition of the quotient of the reactor; the performances of this reactor are characterised by its overall efficiency. The idea of the ratio is introduced because it represents a convenient intermediary in the calculations. The search for an additive value for reactors in series leads logically to the defining of an exchange capacity, and a total efficiency, or number of theoretical reactors. This method of expressing the performances of a reactor is more general than the efficiency due to Murphee which only has a physical significance in the particular case of homogeneous liquid reactors. The relationships between these various quantities are established, and the representation due to Mc Cabe and Thiele is generalized. The reactor performances are linked to the first - order reaction kinetics by the transfer number. The relationships are given for a certain number of concrete cases. Finally the application of these calculations is given, together with the approximations necessary in the case where, because of the presence of several components in each phase, the exchange reaction no longer obeys a single kinetic law. (authors) [French] On examine dans le cas general la theorie d'un reacteur quelconque pour l'echange chimique isotopique du premier ordre entre une phase gazeuse et une phase liquide, a partir des donnees fondamentales sur la cinetique de la reaction, sans faire aucune hypothese limitative sur le cas des reacteurs a contre ourant, puis celui des reacteurs a co-courant. L'equation generale de conservation du deuterium

  19. Technique of nuclear reactors controls; Technique des controles des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

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

    1953-12-15

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [French] Nous avons aborde le probleme de la ''Technique du Controle des reacteurs nucleaires'' dans le but de realiser le controle du reacteur de Saclay. C'est ainsi que nous avons ete amene a etudier le probleme dans son ensemble, tel qu'il se pose pour tout reacteur a uranium naturel. Ce travail traite principalement du domaine des mesures a caractere nucleaire et s'etend dans le domaine des mesures thermodynamque de niveaux, etc... mettant en relief les caracteristiques nouvelles exigees de la part des detecteurs du fait de leur utilisation dans le flux de neutrons thermiques. Dans le domaine de mesures nucleaires, nous indiquons principalement les realisations et les resultats obtenus pour les detecteurs de neutrons thermiques et pour la mesure de courants d'ionisations. Nous traitons egalement du probleme technique du demarrage d'un reacteur et du probleme de la mesure de la reactivite. Nous donnons les details necessaires a la comrehension de tous les schemas et plans de cablages essentiels mis au point, en particulier, pour le reacteur de Saclay. (auteur)

  20. Industrial treatment of solutions of fission products. Separation of caesium-137; Traitement industriel de solutions de produits de fission. Separation du cesium-137

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et confection de

  1. Industrial treatment of solutions of fission products. Separation of caesium-137; Traitement industriel de solutions de produits de fission. Separation du cesium-137

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    Two types of chemical treatment can be considered for the manufacture of solid sources for industrial uses from fission product solutions remaining after plutonium extraction: a) concentration of the solution and preparation of solid sources from the bulk material, without separation, b) separation of one or several fission products from which the sources are made. Examination of the radio-chemical composition of the mixture of fission products that will be available from the Marcoule reactors (G1, G2 and G3) shows that caesium-137 accounts for 30 per cent of the {gamma} energy available immediately after the plutonium separation, 70 per cent two years after and 100 per cent after five years. There is practically no advantage in making sources from bulk fission products, since the separation of caesium-137 is no more complicated and yet it results in a material with more potential uses. The separation of caesium-137 by a method based on the standard phospho-tungstate precipitation method has been considered. Previously, the precipitated caesium phospho-tungstate was dissolved and caesium was recovered from the solution by cation-exchange or by removal of phosphate and tungstate ions by anion-exchange. A study has now been made, of the metathesis of caesium phospho-tungstate to barium phosphate and tungstate by the action of barium hydroxide, the caesium being obtained in solution as the hydroxide. The advantages of this new procedure are: - greater decontamination of caesium-137 without further purification, - possibility of direct transformation to caesium sulphate, - general simplification of the procedure and, consequently, of the equipment. (author) [French] Deux types de traitement chimique peuvent etre envisages pour amener a l'etat de sources solides utilisables industriellement les produits de fission contenus dans les solutions residuaires de l'extraction du plutonium. Ces traitements sont les suivants: a) concentration des solutions et

  2. Les Parcs Industriels Fournisseurs ou le choix de la proximité géographique

    OpenAIRE

    Sonia Adam-Ledunois; Jérôme Guédon; Sophie Renault

    2008-01-01

    International audience; La proximité géographique est régulièrement présentée comme le remède à de nombreux maux, la solution à des relations distendues voire rompues, et cela tant sur un plan social qu'économique. Nombre de décisions ou démarches, civiles, politiques ou stratégiques, placent ainsi la proximité au cœur des dispositifs envisagés (emplois de proximité, police de proximité, fête des voisins, pôle de compétitivité, systèmes de production localisés, districts industriels, etc.). L...

  3. Operational experience of the Marcoule reactors; Experience d'exploitation des reacteurs de Marcoule

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [French] Les resultats atteints apres trois ans de fonctionnement des reacteurs G-2/G-3 permettent une accumulation considerable de l'experience d'exploitation de ces reacteurs. Les principales originalites: - caisson en beton precontraint - chargement en marche - surveillance automatique des temperatures sont largement justifiees par l'exploitation actuelle. L'auteur confirme l'interet de ces solutions d'avant-garde et en tire des conclusions pour les etudes de futures centrales nucleaires. (auteur)

  4. 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)

  5. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core; Recuperation de l'energie degagee dans G 1 pile a graphite refroidie a l'air

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [French] Le Commissariat a l'Energie Atomique (dans le cadre du plan quinquennal) a entre autres objectifs, la realisation des deux premiers reacteurs francais moderes au graphite. La construction du reacteur G-1 a Marcoule, premiere pile plutonigene francaise, est realise afin qu'il puisse diverger au debut de 1956 et atteindre sa pleine puissance au debut du second semestre de la meme annee. Dans ce rapport nous detaillerons les specificites du reacteur et en particulier son systeme de refroidissement et de recuperation d'energie. Le reacteur G-1 etant essentielement destine a permettre aux techniciens francais d'etudier le plus tot possible le comportement d'une installation productrice d'energie empruntant sa chaleur a une source nucleaire. (M.B.)

  6. 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

  7. 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

  8. Neutron noise in nuclear reactors; Le bruit neutronique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Blaquiere, A. [Institut National des Sciences et Techniques Nucleaires (France); Pachowska, R. [Universite Technique de Varsovie (Poland)

    1961-06-15

    The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

  9. Aerodynamic and thermal studies of cans of gas cooled fuel elements; Etudes aerodynamiques et thermiques de gaines d'elements combustibles refroidis par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    l'amelioration des grappes et s'est ensuite poursuivie dans les deux Laboratoires pour le refroidissement interne des elements combustibles annulaires. Au fur et a mesure de l'avancement de ces etudes, les moyens d'essais se sont amplifies tandis que les methodes experimentales n'ont cesse de s'ameliorer. Actuellement, les deux Laboratoires, qui travaillent en pleine collaboration, disposent de moyens puissants. Pour les grappes, l'effort a surtout porte sur les pertes de charge dues aux pieces d'assemblage et sur les variations de temperature autour des elements de la grappe. On est arrive ainsi a determiner d'une facon satisfaisante les points chauds de gaine, les deformations des crayons et les conditions de stabilite de ces deformations. Pour les gaines a chevrons, les etudes ont porte, d'une part sur l'evolution des performances en fonction des parametres geometriques, et, d'autre part sur les singularites aerodynamiques et thermiques creees tant par les ailettes que par les interruptions de cartouche. Ces etudes ont abouti a une connaissance tres complete des cartouches choisies pour les reacteurs EDF2 et EDF3, et ouvrent maintenant des perspectives tres encourageantes pour les reacteurs a venir, en particulier les reacteurs equipes d'elements annulaires; parmi les solutions convenant a la gaine interne de l'element annulaire, les corrugations et les ailettes longitudinales ont fait l'objet d'essais assez etendus dans une large plage de nombre de Reynolds. (auteurs)

  10. Unstable fluid flow in a water-cooled heating channel; Instabilites d'ecoulement du fluide dans un canal chauffant refroidi a l'eau

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    Experimental investigations of the instable behavior of a pressurized water flow in forced convection in a heating channel, with subcooled or bulk boiling have been carried. Tests were conducted at 1140, 850 and 570 psi. The test section was 35 in. high, surmounted by a 25.4 in. riser, these sections were by-passed by a pipe where the flow was between 1 and 4 times the flow in the test section. The water velocity (in the test section) was between 1.6 and 6.6 ft/s. Under certain conditions oscillations with a period of several seconds and perfectly stable have been observed. A mathematical model has been defined and a good agreement obtained for the main characteristics of the oscillations. It seems that the dimensions of the riser have a determining effect: the inception of bulk boiling gives an important variation of the driving head which can generate oscillations due to the non-zero delay for the system to reach its equilibrium. (author) [French] Une investigation experimentale des risques d'instabilite d'ecoulement d'eau sous pression dans un canal chauffant, avec ebullition sous refroidie ou en masse a ete faite. Les essais ont ete effectues a des pressions de 40, 60 et 80 kg/cm{sup 2}. La section d'essais d'une hauteur de 1,37m etait surmontee d'une cheminee de 1 m, le tout by-passe par une conduite ou le debit pouvait varier entre 1 et 4 fois le debit dans la section d'essais. La vitesse dans la section variait entre 0,5 et 2 m/s. Dans certaines conditions, des oscillations de periode de l'ordre de quelques secondes et parfaitement stables sont apparues. Un modele mathematique a permis de retrouver les principales caracteristiques de ces oscillations. Il semble que l'influence des dimensions de la cheminee soit determinante: en effet, l'apparition de l'ebullition de masse entraine une importante variation de la hauteur motrice qui peut engendrer des oscillations entretenues a cause du retard non nul a la mise en equilibre du systeme. (auteur)

  11. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

  12. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de fonctionnement et le

  13. 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)

  14. Some particular aspects of control in nuclear power reactors; Conception de la surete en france et influence des imperatifs de surete sur la conception des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Vathaire, F de; Vernier, Ph; Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [French] La presente communication propose une synthese de l'experience acquise en France en matiere de surete des reacteurs. Les reacteurs de la filiere graphite-gaz faisant l'objet d'une communication particuliere, on examine ici la surete des autres types de reacteurs etudies en France: - reacteurs eau lourde-gaz, - reacteurs a neutrons rapides, - reacteurs de recherche a eau des types piscines et tank. Les imperatifs de surete propres aux differentes filieres sont developpes, en mettant l'accent sur leur influence sur la conception des reacteurs et sur les limitations de puissance qu'ils entrainent. Les etudes de surete correspondantes sont presentees, en insistant plus particulierement sur les travaux originaux developpes dans ces domaines. On indique notamment les moyens d'essais qui ont ete construits pour ces etudes: le reacteur CABRI, boucle en pile pour essais de depressurisation, boucles hors pile, maquettes, etc. (auteurs)

  15. Underground seasonal storage of industrial waste heat; Saisonale Speicherung industrieller Abwaerme im Untergrund

    Energy Technology Data Exchange (ETDEWEB)

    Reuss, M.; Mueller, J. [Bayerische Landesanstalt fuer Landtechnik, TU Muenchen-Weihenstephan, Freising (Germany)

    1998-12-31

    The thermal efficiency of subject systems, especially at higher temperatures is influenced by heat and humidity transport underground. Thermal conductivity and specific thermal capacity depend on the humidity content of the soil. A simulation model was developed that describes the coupled heat and humidity transport in the temperature range up to 90 C. This model will be validated in laboratory and field tests and then be used for designing and analysing underground stores. Pilot plants for the storage of industrial waste heat were designed and planned on the basis of this simulation. In both cases these are cogeneration plants whose waste heat was to be used for space heating and as process energy. Both plants have a very high demand of electric energy which is mostly supplied by the cogeneration plant. The waste heat is put into the store during the summer. In the winter heat is supplied by both the store and the cogeneration plant. In both cases the store has a volume of approx. 15,000 cubic metres with 140 and 210 pits located in a depth of 30 and 40 metres. The plants are used to carry out extensive measurements for the validation of simulation models. (orig.) [Deutsch] Die thermische Leistungsfaehigkeit solcher Systeme wird insbesondere im hoeheren Temperaturbereich durch den Waerme- und Feuchtetransport im Untergrund beeinflusst. Sowohl die Waermeleitfaehigkeit als auch die spezifische Waermekapazitaet sind vom Feuchtegehalt des Bodens abhaengig. Es wurde ein Simulationsmodell entwickelt, das den gekoppelten Waerme- und Feuchtetransport im Temperaturbereich bis 90 C beschreibt. Dieses Modell wird an Labor- und Feldexperimenten validiert und dient dann zur Auslegung und Analyse von Erdwaermesonden-Speichern. Basierend auf diesen theoretischen Grundlagenarbeiten wurden Pilotanlagen zur saisonalen Speicherung industrieller Abwaerme ausgelegt und geplant. In beiden Faellen handelt es sich um Kraft/Waermekopplungsanlagen, deren Abwaerme zur Gebaeudeheizung und

  16. 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

  17. 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

  18. Sociologie du travail et critique du temps industriel The Sociology of Labor and a critical vision of industrial Time

    Directory of Open Access Journals (Sweden)

    Jens Thoemmes

    2009-07-01

    Full Text Available La sociologie du travail a développé un point de vue particulier sur l'analyse des temporalités sociales. L'objectif de cet article est de montrer l'élaboration progressive d'une critique d'un temps industriel unique et unifiant. Cette critique s'appuie sur trois éléments : la multiplicité des temporalités de l’activité professionnelle, le caractère déstructurant du travail industriel et l'émergence d'un temps des loisirs. Ce point de vue a été élaboré par la sociologie du travail en France après 1945, notamment par Georges Friedmann, Pierre Naville et William Grossin.  Avant d’aborder cette période nous voudrions  interroger le moment de la fondation de la sociologie. En mobilisant des travaux peu connus de Max Weber sur le travail, nous verrons en quoi il est un précurseur de l’analyse des attitudes diversifiées des individus et des collectifs à l’égard du temps. Cette perspective lie, dès les premières enquêtes sociologiques, le travail industriel à une interrogation sur les temporalités.The sociology of work and labor developed a particular viewpoint on the analysis of social times. The objective of this article is to show how a criticism of a unique and unifying industrial time progressively grew up. Such criticism leans on three elements: the multiplicity of social times linked to the occupational activity, the disintegrating nature of work in the manufacturing process and the emergence of leisure-time. This point of view was elaborated by the sociology of work in France after 1945, notably by Georges Friedmann, Pierre Naville and William Grossin. Before approaching this period we question the moment of the foundation of sociology. By mobilizing little known research by Max Weber on the manufacturing process, we shall see in what way he was a precursor of the analysis of diversified attitudes regarding time by individuals and groups.

  19. 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)

  20. New modelling method for fast reactor neutronic behaviours analysis; Nouvelles methodes de modelisation neutronique des reacteurs rapides de quatrieme Generation

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, P.

    2011-05-23

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author) [French] Les criteres de surete qui regissent le developpement de coeurs de reacteurs de quatrieme generation implique l'usage d'outils de calcul neutronique performants. Une premiere partie de la these reprend toutes les etapes de modelisation neutronique des reacteurs rapides actuellement d'usage dans le code de reference ECCO. La capacite des modeles a decrire le phenomene d'autoprotection, a representer les fuites neutroniques au niveau d'un reseau d'assemblages combustibles et a generer des sections macroscopiques representatives est appreciee sur le domaine des reacteurs rapides innovants respectant les criteres de quatrieme generation. La deuxieme partie de ce memoire se consacre a la modelisation des coeurs rapides avec reflecteur acier. Ces derniers necessitent le developpement de methodes avancees de condensation et d'homogenisation. Plusieurs methodes sont proposees et confrontees sur un probleme de modelisation typique: le coeur ZONA2B du reacteur maquette MASURCA

  1. 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.)

  2. Aerodynamic study of the fluid flow in the channel of a reactor filled with internally and externally cooled fuel elements; Etude aerodynamique de l'ecoulement fluide dans un canal de reacteur charge en elements combustibles annulaires refroidis interieurement et exterieurement

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author) [French] On etudie le probleme de la repartition des debits et des pressions le long des deux espaces, interne et externe, delimites par une succession d'elements annulaires non jointifs, disposes suivant l'axe d'un canal. On constate que le phenomene peut etre mis aisement en equation. Les relations theoriques obtenues prennent en particulier une forme simple lorsque les intervalles entre elements sont superieurs a une valeur minimum. L'etude experimentale a permis de constater que cette formulation theorique etablie etait alors valable avec une precision excellente. L'etude experimentale a ete egalement effectuee dans le cas de tres faibles intervalles entre elements consecutifs. On a pu alors verifier que l'hypothese adoptee dans l'etablissement des relations theoriques etait parfaitement justifiee. Dans un dernier chapitre enfin, nous abordons le point de vue pratique de l'evaluation des pertes de charge aux bornes d'un train d'elements annulaires. (auteur)

  3. Aerodynamic study of the fluid flow in the channel of a reactor filled with internally and externally cooled fuel elements; Etude aerodynamique de l'ecoulement fluide dans un canal de reacteur charge en elements combustibles annulaires refroidis interieurement et exterieurement

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author) [French] On etudie le probleme de la repartition des debits et des pressions le long des deux espaces, interne et externe, delimites par une succession d'elements annulaires non jointifs, disposes suivant l'axe d'un canal. On constate que le phenomene peut etre mis aisement en equation. Les relations theoriques obtenues prennent en particulier une forme simple lorsque les intervalles entre elements sont superieurs a une valeur minimum. L'etude experimentale a permis de constater que cette formulation theorique etablie etait alors valable avec une precision excellente. L'etude experimentale a ete egalement effectuee dans le cas de tres faibles intervalles entre elements consecutifs. On a pu alors verifier que l'hypothese adoptee dans l'etablissement des relations theoriques etait parfaitement justifiee. Dans un dernier chapitre enfin, nous abordons le point de vue pratique de l'evaluation des pertes de charge aux bornes d'un train d'elements annulaires. (auteur)

  4. 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

  5. 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)

  6. 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)

  7. 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)

  8. 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

    The introduction, at an increasing rate, of power reactors using slightly enriched uranium will inevitably lead to the production of considerable quantities of plutonium over the next decade. Fast reactors will not be capable of absorbing this material before 1980. The question thus arises of whether one should store the plutonium far future use in fast reactors, recycle it in existing thermal reactors, or try to sell it. The problem has been studied for an electric power generating system that does not foresee selling the plutonium produced by its reactors and does not buy plutonium outside, which enables a good approximation to be made and eliminates the major unknown quantity represented by the future market price of plutonium. Assuming within this system a programme that provides for the construction of power reactors of a given type and capacity at specific dates, the utilization of the plutonium produced can be optimized by linear programming techniques so as to minimize the discounted total cost of the power generated over a given period. A later stage consists in optimizing, by various techniques, not only the utilization but also the production of plutonium by appropriate selection of the power reactor types to be constructed. (author) [French] L'implantation, a un rythme croissant, de centrales nucleaires a uranium legerement enrichi entrainera la production ineluctable d'une quantite importante de plutonium au cours de la prochaine decennie. Les reacteurs a neutrons rapides ne seront capables d'absorber cette production qu'apres 1980. La question se pose donc de savoir s'il est preferable de stocker le plutonium en vue de son utilisation ulterieure dans les reacteurs a neutrons rapides plutot que de le recycler dans les reacteurs actuels a neutrons thermiques ou d'essayer de le vendre. Ce probleme a ete etudie dans le cadre d'un systeme de production d'energie electrique qui ne prevoirait pas la vente du plutonium produit par ses reacteurs nucleaires ni

  9. L’écologie industrielle : quand l’écosystème industriel devient un vecteur du développement durable

    Directory of Open Access Journals (Sweden)

    Arnaud Diemer

    2007-08-01

    Full Text Available L’écologie industrielle, définie par Robert Frosch (1995 comme « l’ensemble des pratiques destinées à réduire la pollution industrielle », nous amène à penser que l’écosystème industriel peut être un véritable vecteur du développement durable. L’ingénierie écologique et l’écotechnologie recommandent aux industriels de procéder à un ensemble d’opérations de rationalisation de la production (optimisation des consommations énergétiques et matérielles, minimisation des déchets à la source,réutilisation des rejets pour servir de matières premières à d’autres processus de production. Les symbioses industrielles et les parcs éco-industriels sont généralement présentés comme des modèles de rationalisation industrielle et des illustrations tangibles du développement durable.Industrial ecology is defined by Robert Frosch (1995 as practices intended to reduce industrial pollution. That leads us to think industrial ecosystem as a vector of sustainable development.Ecological engineering and ecotechnology recommend managers to rationalize the production process (optimization of material consumptions, minimization of bads…. Industrial symbiosis and industrial parks are generally presented as models of industrial rationalization and tangible illustrations of the sustainable development.

  10. La montagne, lieu de développement industriel : l’exemple du Languedoc-Roussillon

    Directory of Open Access Journals (Sweden)

    2007-03-01

    Full Text Available Michel WieninRiches en énergie (rivières, bois, puis houille, en matières premières minérales et en produits d’élevage, les régions de montagne ont développé tôt une industrie qui profite de l’essor des transports au XIXe siècle. Loin des grands bassins miniers, les vallées se transforment en une succession de bourgs industriels, accédant beaucoup plus vite à la modernité que les plaines restées agricoles ou que les villages d’altitude, à l’écart du progrès. Délimité par le Massif Central et les Pyrénées, le Languedoc-Roussillon présente une plaine évoluant vers une quasi-monoculture de la vigne et des vallées de montagne qui se tournent vers l’industrie textile et métallurgique.France’s mountain regions, which are rich in varied sources of energy (rivers, woods, then coal, in mineral resources and in products derived from livestock rearing, developed industries at an early date, benefiting from the development of new transport networks during the 19th century. Far from the major mining basins, the mountain valleys were transformed into a succession of industrial villages, reaching modernity more rapidly than the plains, still devoted to agriculture, or the villages high in the mountains and isolated from such progress. Bordered by the Massif Central and the Pyrenees, the region of Languedoc-Roussillon comprises a plain which evolved towards a mono-activity of wine production and the mountain valleys which witnessed the development of textile and metallurgical industries.

  11. Storage of plugs and experimental devices from reactors; Stockage des bouchons et dispositifs experimentaux en provenance des reacteurs (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Cerre, P; Mestre, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    - Within the general programme of storage and treatment of radioactive waste produced by the various operations carried out in an atomic center, it is useful to consider separately the problem of certain waste from reactors, which, because of its size and physical nature, has to be stored with a view to being later treated and finally evacuated. The solution which we propose for this storage problem is presented in this paper. (authors) [French] - Dans le cadre du stockage et du conditionnement des dechets radioactifs provenant des diverses manipulations effectuees dans un centre atomique, il y a lieu de considerer a part certains dechets des reacteurs qui, par leur dimension et leur nature physique doivent etre stockes en vue de leur reprise ulterieure pour un conditionnement et une evacuation definitifs. La solution que nous avons apportee a ce stockage fait l'objet de l'expose qui suit. (auteurs)

  12. Description of the french graphite reactor and of the experiments performed in 1956; Presentation du premier reacteur a graphite francais et des experiences effectuees en 1956

    Energy Technology Data Exchange (ETDEWEB)

    Bussac, J; Leduc, C; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [French] Ce rapport presente les experiences qui furent faites sur le reacteur G1 et dont la description en detail fait l'objet des rapports suivants (670 'B a P'). Les principaux resultats sont fournis ici et commentes. On trouvera en outre les caracteristiques neutroniques du coeur actif de la pile, une description des principales installations et une mention des essais qui ont conduit au fonctionnement normal du reacteur en puissance. (auteur)

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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

  20. Optimistaion énergétique d'un ensemble industriel Energy Optimization of an Industrial Installation

    Directory of Open Access Journals (Sweden)

    Raimbault C.

    2006-11-01

    Full Text Available Cet article porte sur la mise au point d'une méthode d'optimisation du système de production d'utilités d'un ensemble industriel et l'étude de dispositifs destinés à économiser l'énergie. Le programme d'optimisation fait appel à la programmation linéaire. Un programme générateur de matrice et un programme de traduction des résultats ont été mis au point. On dispose ainsi d'un programme d'optimisation adapté à tout système de production d'utilités. Différents dispositifs permettant d'économiser l'énergie ont été étudiés. L'étude a porté, d'une part, sur des dispositifs classiques tels que les dispositifs de récupération de chaleur sur les fumées et, d'autre part, sur des dispositifs nouveaux. Des solutions nouvelles ont été recherchées dans deux domaines qui sont apparus essentiels : production combinée de travail et de chaleur et valorisation de calories à bas niveau. Enfin la méthode d'optimisation a été appliquée au cas d'une raffinerie réelle dont l'étude avait été effectuée récemment. L'optimisation sur une base économique a permis de dégager une économie de 9,3 % sur la consommation d'énergie, mais a surtout démontré les larges possibilités de la méthode dans son application à un cas concret. This article describes the development of a method for optimizing utilities production systems for on industrial installation and the study of energy-saving systems. The optimization program makes use of linear programming. A matrix-generating program and a result-translating program were developed. The result is an optimization program suited for any utilities production system. Different energy-saving systems were examined, including conventional systems such as heat-recovery devices as well as new systems. New solutions were sought for in two fields which appear essential, i. e. the combined production of work and heat and the valorization of low-level calories. The optimization method was

  1. Master i Industriel Informationsteknologi

    DEFF Research Database (Denmark)

    Knudsen, Morten

    2000-01-01

    Målgruppen for denne åbne Masteruddannelse, som udbydes af Aalborg Universitet i samarbejde med IT-Vest, er personer, som dagligt arbejder med teknisk IT, og som har en videregående uddannelse på bachelorniveau. Uddannelsen, som er 3- årig på ½ tid, startede September 1998, og det samlede optag d...

  2. Master i Industriel Informationsteknologi

    DEFF Research Database (Denmark)

    Knudsen, Morten

    2001-01-01

    Målgruppen for denne åbne Masteruddannelse, som udbydes af Aalborg Universitet i samarbejde med IT-Vest, er personer, som dagligt arbejder med teknisk IT, og som har en videregående uddannelse på bachelorniveau. Uddannelsen, som er 3- årig på ½ tid, startede September 1998, og det samlede optag d...

  3. Industriel patriarkalisme som moderniseringsstrategi

    DEFF Research Database (Denmark)

    Rasmussen, Marianne Rostgård

    2003-01-01

    with hierachical social relations between industrialists and workers, specifying a set of duties and obligations for each of the two parties to meet. Industrialists were, among other things, expected to care for workers who were no longer able to provide for themselves due to illness or old age. The article...... discusses different schemes for taking care of ill and aged workers and shows that such schemes were relatively common around 1900. The discussion focuses on norms and practices, not effects. Whether the schemes actually resulted in lessened social tensions during the transformation to an industrial society...

  4. 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)

  5. 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

    A pulsed fast reactor (IBR) has been operating at rated capacity since December 1960 in the Joint Institute for Nuclear Research. This reactor is used as a pulsed neutron source for physical experiments carried out by the time-of-flight method. It is used for total cross-section and intermediate neutron capture cross- section measurements, for studying the interaction between slow neutrons and solids and liquids, and for measuring neutron spectra produced in various media. The paper describes the basic structural features of the reactor and the results of the experiments for which it has been used. The reactor's operating system is based on recurrent pulses. Power pulses are produced when the mobile part of the reactor core moves swiftly through the stationary part of the core. The mobile part of the core is fastened to a rotating disc and travels at a speed of 230 m/s. The frequency of power pulses can be altered by means of an auxiliary mobile zone which has a range of 2.3-88 pulses per second. The mean power of the reactor is 1 kW, and the half-width of the power pulse in 36 {mu}s. The reactor is provided with a control and safety system which ensures automatic maintenance of mean power and swift shutdown in the event of any operational irregularity. It is fitted with a system of evacuated-neutron-flight tubes used in time-of-flight experiments. The main tube is 1000 m in length. In the start-up process and during physical experiments carried out on the reactor, the influence on reactivity of displacing the controls and the mobile parts of the core was studied ; the length of the pulse was measured under various operating conditions, and power pulse amplitude fluctuations were studied. Further measurements were made to establish the lifetime of prompt neutrons, the effective fraction of delayed neutrons, and coefficients of reactivity. (author) [French] L'Institut unifie de recherches nucleaires dispose d'un reacteur puise a neutrons rapides (IBR), qui

  6. Églises et temples en bassins industriels : Belfort-Héricourt-Montbéliard (Franche-Comté (1944-2008

    Directory of Open Access Journals (Sweden)

    Yves-Claude Lequin

    2009-11-01

    Full Text Available Si l’on en juge par l’exemple de Belfort-Montbéliard, nombreux furent les lieux de culte - catholiques et luthériens - construits dans les bassins industriels français entre 1950 et 1968. Étapes : églises modernes « en majesté » vers 1950, églises sobres aux formes arrondies dans les ZUP vers 1960, préfabriqués après 1964, quasi arrêt en 1968, reconversion du bâti à partir de 1990. Raisons de cet essor puissant et du rapide déclin : les tensions sociales et les évolutions mentales des paroissiens.From the example of Belfort-Montbéliard, a fair number of places of worship - Catholic and Lutheran - were built in French industrial basins between 1950 and 1968. Stages: modern “in majesty” churches around 1950, sober buildings with curved shapes in priority development areas around 1960, prefabricated buildings after 1964, hardly any construction in 1968, conversion of buildings as from 1990. Reasons for this marked growth and for the swift decline: social tensions and evolution of parishioners ‘minds.

  7. 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)

  8. 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)

  9. 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)

  10. 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)

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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.

  16. Contribution to multi-agents modeling of the operation of industrial processes: application to the operation of a pressurized water reactor under accidental situation; Contribution a la modelisation multi-agents de la conduite de processus industriels: application a la conduite en situation accidentelle d`un reacteur nucleaire a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Elias, P.

    1996-11-13

    This work is related to the CEA `Escrime` project which concerns the reliability and functioning safety of nuclear reactors, and in particular the operation and supervision of nuclear installations. Its aim is the analysis and the formalizing of PWRs operation in order to define the collaboration and optimum sharing of tasks between human operators and automatized systems for an improved functioning safety. Chapter 1 describes the operation of nuclear reactors and the instrumentation and control activities. It focusses on the weaknesses of actual automatized systems and examines the interest of the multi-agents approach to build an improved automatized system. Chapter 2 presents the actual state of the art about multi-agent systems and about their application to reactor operation. Chapter 3 is devoted to the definition of the conceptual model of automatized systems developed in this work (distribution of operation activities, competition between agents, hierarchy, arbitration). Chapter 4 describes the computer model of the essential operating system elaborated according to the conceptual model defined above. Modeling is performed using Spirit and an application is described in chapter 5. (J.S.). 58 refs.

  17. Integration of industrial risk in regional policy management. Possibilities of evaluation; L'integration du risque industriel dans les politiques de gestion territoriales. Possibilites d'evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Hubert, Philippe; Pages, Pierre

    1990-02-01

    Since the responsibility for risk management depends more and more on regional factors, evaluation methods and management regulations are not developed enough. This study start from the fact that an important methods exist for evaluation transport of dangerous materials in cities, risk analysis and emergency plans related to classified installations, management of quality of water, 'chronic' industrial risks at local and regional level, probabilistic estimation for industrial plants. The objective is in fact to show what risk analysis could bring to the municipality, the city or the region. [French] Alors que la responsabilite de la maitrise du risque repose de plus en plus sur les collectivites territoriales, soit dans la mouvance naturelle de la decentralisation, soit par des textes specifiques, les moyens d'evaluation et les regles de gestion sont encore tres peu developpes. Cette etude part du fait qu'un materiel important existe cependant: evaluations sur le transport des matieres dangereuses dans les villes, etudes de danger et plans d'intervention associes a la legislation sur les installations classees, gestion de la qualite de l'eau par les agences de bassin, bilans des risques industriels 'chroniques' a l'echelle locale ou regionale, evaluations probabilistes sur des objets industriels. L'objet est donc de montrer ce que peuvent apporter ces analyses dans l'evaluation du risque accidentel dans la collectivite, la ville ou, a cause des implications des strategies de prevention, le departement ou la region. L'heterogeneite de la qualite et de la quantite des resultats que l'on peut attendre en fonction des divers types de risques et de consequences est d'abord mise en valeur. Malgre leurs incertitudes, des modeles existent qui permettent de calculer les victimes d'accidents industriels 'types'. Pour les installations classees et les transports de matieres dangereuses, le calcul du risque est loin d'etre systematique, mais il est pratique. Un second domaine

  18. 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

  19. 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)

  20. Description of methods for making activation detectors for use in nuclear reactors; Description des procedes de fabrication des detecteurs d'activation utilises dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Barbalat, R; Le Coguie, R; Leger, P; Salon, L; Thierry, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1968-07-01

    A brief description of methods currently used for making activation detectors, thin films and various deposits used in nuclear reactors. The thicknesses required vary from about a few tenths of a micron to a few tenths of a millimeter. Different techniques are used for fixing the large variety of elements: rolling, moulding, painting, electrolysis, vacuum deposition, thin films, wires, enamels, protective linings, etc. (authors) [French] Expose succinct des procedes actuellement mis en oeuvre pour la realisation des detecteurs d'activation, feuilles minces et depots divers utilises dans les reacteurs nucleaires. La gamme des epaisseurs necessaires s'etendant approximativement des dixiemes de micrometre aux dixiemes de millimetre. La diversite des elements a fixer justifiant les techniques differentes selon les cas: laminage, moulage, peinture, electrolyse, depot sous vide, couches minces, fils, emaux, revetements protecteurs, etc. (auteurs)

  1. 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)

  2. 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)

  3. Construction of the core of the 'heavy water-gas' reactor EL 4; Structures du coeur du reacteur 'eau- lourde-gaz EL 4'

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J L; Foulquier, H; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    problem of thermal insulation around a zirconium alloy liner tube. The neutron absorption equivalent is about 1, 1 mm of Al, and the mean loss around 2 p. 100 of the thermal power of the reactor. The methods proposed have proved practicable as a result of important research and developments on automatic remote control for all the operations which make up the sequences of mounting, demounting and repairing of the construction components. In particular the possibilities opened up by the new techniques of welding tubes from the inside have been extended to other problems connected with the assembling of a reactor. (authors) [French] Le coeur de ce reacteur est constitue par une cuve contenant l'eau lourde, cuve traversee d'une serie de tubes de force dans lesquels circule le gaz caloporteur sous pression de 60 at. Les specifications de depart qui ont joue un role important dans la conception de ces structures concernent des aspects de securite de fonctionnement (chargement du combustible par les deux faces du reacteur, remplacement des structures sur les deux faces du reacteur), des necessites neutroniques (absorption des structures minimum, pas du reseau, diametre des tubes de force) et des considerations thermiques (temperature de sortie 500 C). Ces specifications ont entraine une disposition horizontale des tubes de force et des problemes d'encombrement tres delicats qui ont elimine (pour les dimensions d'EL 4) toute possibilite de recourir a des compensateurs de dilatation sur les tubes de force. II s'ensuit un dessin de cuve semi-rigide dans lequel les tubes de force contribuent pour une part importante a la resistance mecanique de l'ensemble en jouant le role de tirant, d'ou des contraintes elevees sur les jonctions et tubes de force (et le choix des alliages de zirconium). Les structures comprennent le tube de force, les jonctions, l'isolement thermique et le tube de guidage. On expose brievement les moyens d'essais mis en oeuvre et les performances de ces diverses

  4. 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

  5. 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

  6. 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

  7. Some fundamental aspects of boiling in nuclear reactors; Quelques aspects fondamentaux de l'ebullition dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Mondin, H; Lavigne, P; Semeria, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    oscillation, the conditions of burnout are compared with those obtained under steady conditions. The burn-out flux following uniform 'stopped' heating has been studied in a channel containing still water. The flux shows a maximum as a function of unsaturation. The influence of the geometry and the nature of the metal was investigated. 4 - Output Oscillations: Using a low pressure (8 atm) loop, the influence of various parameters on the periods of output oscillations in a boiling channel on the thresholds at which they appear, was studied. Some new aspects of this complex phenomena were observed and are reported. (authors) [French] On indique les principaux resultats obtenus a Grenoble depuis quatre ans dans le domaine des mecanismes de l'ebullition et des phenomenes connexes dans les reacteurs nucleaires. 1 - OBSERVATION DE L'EBULLITION: Par photographie et cinematographie ultrarapide (8000 images par seconde maximum) on a observe l'ebullition en vase ou en canal jusqu'a 140 kg/cm{sup 2}. On a denombre les populations de germes (sites) generateurs de bulles et obtenu une correlation donnant leur nombre par unite de surface en fonction du flux thermique et de la pression. Le diametre des bulles se detachant de la paroi a ete etudie jusqu'a 140 kg/cm{sup 2}. On a mis en evidence trois types de bulles: - Les bulles en equilibre dont le diametre suit la formule de Fritz et Ende, - Les bulles d'ebullition dont le diametre diminue rapidement avec la pression (1/100 mm a 140 kg/cm{sup 2}), - Les coalescences apparaissant en liquide sature au-dessus de 15 W/cm{sup 2} et dont la proportion est independante de la pression. Par visualisation en strioscopie on observe les mouvements du film thermique associes a l'amorcage des germes, au depart et a la condensation des bulles; les mecanismes responsables de l'excellent transfert de chaleur ont pu ainsi etre precises. 2 - PERTES DE PRESSION EN ECOULEMENT DIPHASE: On a etabli un modele de variation continue du taux de vide dans un canal

  8. Global Industriel E-handel

    DEFF Research Database (Denmark)

    Rask, Morten

    Afhandlingen undersøger danske industrivirksomheders transformationsproces på vej mod en ny æra, hvor Internettet anvendes som et centralt redskab i samhandelen med andre virksomheder, og hvor virksomhedens konkurrenceevne og globaliseringsproces styrkes som følge af anvendelsen. Formålet med afh...

  9. 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

  10. [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

  11. Measurement and regulation of the level of a homogeneous plutonium reactor; Mesure et regulation du niveau d'un reacteur homogene au plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Berger, F; Bertrand, J

    1958-12-01

    Reactivity depends strongly on disturbances of the level of the plutonium solution In the homogeneous reactor. Proserpine has a small cylindrical core, 250 mm diameter, and 10 liters volume. With a view to reducing the dangers due to corrosion and contamination, the solution level in the core is raised by pneumatic pressure. The level is stabilized by means of a regulating system. During critical experiments the variations of the level are less than one hundredth part of a millimeter. (author) [French] Les variations du niveau de la solution de plutonium dans le reacteur homogene Proserpine ont une grosse influence sur la reactivite, car le coeur est petit (10 litres de solution dans un cylindre de diametre 250 mm). En vue de reduire les dangers dus a la corrosion et a la contamination, la commande du volume liquide est pneumatique. Nous avons realise la stabilite du niveau par une regulation qui, dans les essais en regime critique, limite les variations du plan liquide a une fraction de centieme de millimetre. (auteur)

  12. 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)

  13. 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)

  14. Ultrasonic testing of canning tubes in stainless steel of the EL 4 reactor; Controle par ultrasons des tubes de gaine en acier inoxydable du reacteur EL 4

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    From all the methods possible for controlling thin cans the one chosen, for numerous reasons, vas that making use of ultrasonic techniques. A method has been developed which should make it possible to carry out a rapid and efficient industrial control of canning tubes, The reasons for the choice of the ultrasonic method are given in detail, together with the principles of the method and the actual control parameters. In the present state of our research, it should be possible to control at least 50 000 tubes a year. Improvements brought about in the details of the control technique itself should make it possible to increase this rate considerably. (authors) [French] Parmi toutes les methodes possibles de controle des gaines minces, le procede retenu pour de multiples raisons a ete celui faisant appel a la technique des ultrasons. Une methode a ete mise au point qui doit permettre un controle industriel rapide et efficace des tubes de gaine. Sont exposes en detail, les raisons du choix de la methode par ultrasons, les principes de cette methode et les parametres du controle proprement dit. Dans l'etat actuel de nos etudes la cadence devrait permettre le controle de 50000 tubes par an au minimum. Des ameliorations de detail portant sur la technique de controle elle-meme, doivent permettre d'accelerer tres notablement cette cadence. (auteurs)

  15. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P. [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines

  16. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines comprennent la seismicite

  17. Production of a magnetic field of 1 600 000 At/m in a non-magnetic region of volume 0.5 cubic decimeters by a coil cooled by liquid hydrogen (1963); Production d'un champ magnetique de l 600 000 At/m dans un milieu non magnetique dont le volume est 0,5 decimetre cube. Enroulement refroidi a l'hydrogene liquide (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    The eventual application of the technique of cryogenic coils for the production of magnetic fields for liquid hydrogen bubble chambers has led us to the study and construction of two copper coils (cooled in a bath of liquid hydrogen). Part one treats certain generalities concerning the resistivity of several pure metals at very low temperatures, the gain in power, thermal isolation and the choice of conductor and power supply. Part two describes the experimental arrangement. The last part is devoted to a description of the two coils which were constructed, and to the results obtained. (author) [French] L'application eventuelle de la technique des ''bobines cryogeniques'' aux chambres a bulles a hydrogene liquide pour la production du champ magnetique nous a conduits a l'etude et a la construction de deux bobines en cuivre (refroidies dans un bain d'hydrogene liquide). La premiere partie traite de generalites concernant la resistivite de quelques metaux purs, a tres basse temperature, le gain de puissance, l'isolement thermique et le choix du conducteur et de l'alimentation electrique. Dans la seconde partie on decrit l'installation d'essais. La derniere partie est consacree a la description des deux bobines realisees et aux resultats obtenus. (auteur)

  18. Le néo-corporatisme réinterpellé : analyse comparée de deux politiques d'accès à l'emploi, l'apprentissage industriel en Belgique et le contrat de qualification "jeunes" en France

    OpenAIRE

    Levêque, Audrey

    2006-01-01

    The comparative study of the "apprentissage industriel" in French speaking Belgium and of the "contrat de qualification jeunes" in France shows the maintenance of the neo corporatist model of social State. The analysis of the emergence and of the implementation of these two employment public policies for low qualified young permit to understand how they are appropriated on the ground, especially in the metallurgy sector. If these policies are intended for those low qualified young, the implem...

  19. The development of a pilot industrial irradiation facility; Realisation d'une station d'irradiation {gamma} a caractere 'semi-industriel'

    Energy Technology Data Exchange (ETDEWEB)

    Balestic, F; Leveque, P; Prevost, C; Comte, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Saclay, utilisee pour le stockage des barres d'uranium apres leur defournement, a ete specialement amenagee pour permettre des irradiations sur des quantites semi-industrielles de produits chimiques et alimentaires. Des calculs de flux {gamma} ont montre qu'il etait possible de realiser des sources intenses et d'irradier de grands volumes avec des arrangements particuliers de barres dans la casemate. L'installation comporte ainsi deux parties ou different la repartition des barres et les dispositifs d'irradiation. Dans une premiere chambre betonnee et climatisee, 9 rangees verticales de barres d'uranium et de canaux a irradiation alternent avec un ecartement minimum de fa n a constituer un assemblage compact. Le diametre des canaux varie de 10 a 20 cm et la hauteur de chargement utile, 70 cm, correspond a la zone sensiblement uniforme du rayonnement. Dans une deuxieme chambre maintenue a -7 deg. C, des balancelles (h: 70 cm, Q: 35 cm) guidees par un monorail circulent autour d'une seule rangee de 20 barres. La carte des flux y a l'interieur de la casemate a ete etablie par une methode de calcul deja utilisee pour les barres de la pile EL2 a Saclay. Le champ du rayonnement varie dans le temps principalement au cours des premiers mois apres le defournement et suivant le motif geo etrique adopte pour les barres. Les intensites d onisation theoriques sont comprises entre dix mille et plusieurs millions de roentgens par heure. Ces valeurs sont comparees aux resultats obtenus avec les dosimetres chimiques au sulfate ferreux et a l'acide oxalique. En outre, des verreseurs etalonnes avec des sources de cobalt permettent un controle rapide de la dose delivree aux balancelles du convoyeur durant leur parcours dans la chambre froide. La casemate de desactivation de EL3 constitue de par sa capacite de traitement la premiere realisation en France d'une station d'irradiation {gamma} a caractere semi-industriel effectuee avec le concours des services du Commissariat a l

  20. Les hydrocarbures aromatiques polycycliques dans l'environnement : la réhabilitation des anciens sites industriels The Polycyclic Aromatic Hydrocarbons in the Environment : the Former Industrial Sites Remediation

    Directory of Open Access Journals (Sweden)

    Costes J. M.

    2006-12-01

    Full Text Available Les hydrocarbures aromatiques polycycliques ou HAP peuvent être d'origine naturelle mais ils proviennent principalement des processus de pyrolyse. On peut les retrouver dans les sols de certains anciens sites industriels. Cela peut être le cas des sites d'anciennes usines à gaz. Même si aucune conséquence sur la santé humaine n'a été signalée et même si les risques paraissent virtuels, le principe de précaution rend nécessaire de s'occuper des risques liés à ces anciens sites industriels. Gaz de France, propriétaire de 467 sites d'anciennes usines à gaz assume l'héritage industriel dans le cadre d'un protocole signé avec le ministère de l'Environnement. Après une étude des sols, une évaluation des risques est réalisée. En fonction des résultats de cette évaluation des risques et de l'usage du site (actuel et prévu, des solutions de traitement peuvent être mises en Suvre. Parmi les techniques applicables aux sols pollués par des HAP, un intérêt particulier s'est porté sur les traitements biologiques, en pleine évolution, qui offrent une solution économique bien adaptée au traitement de grands volumes de sols souillés par une pollution organique moyennement concentrée. Polycyclic aromatic hydrocarbons (PAHs can be found under natural conditions but they can be produced by pyrolysis processes. They can be found in former industrial sites subsoil, especially on Manufactured Gas Plant sites (MGP sites. Gaz de France has inherited the patrimony of former French gas companies on nationalisation in 1946; consequently, Gaz De France is still the owner of 467 of manufactured gas plants. Even if no impact on human health has been detected and even if the risks seem to be virtual, Gaz de France has to prevent any environmental consequence due to the possible presence of residues in the subsoil of the sites: a protocol has been signed with the French Ministry of Environment. Following the investigations on the site, a

  1. 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

  2. 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)

  3. 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)

  4. 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)

  5. 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)

  6. 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)

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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

    This apparatus is designed to measure instantaneously the positive or negative reactivity of a uranium reactor moderated by light water, on condition that the point of departure is the critical state of the reactor, or an already known sub-critical state. Slight modifications only are required to adapt it to another type of reactor. It is an analogue computer which simply inverses the transfer function of the reactor; it is not therefore a model reactor of which the output voltage is connected by a servo-mechanism to the power of the reactor to give the reactivity; the principle of the calculation of the reactivity does not depend on a servomechanism. One of its disadvantages is that it cannot operate outside a power variation range of 2.5 decades. However the measurement of a negative reactivity value between 0 and 3000 pcm is immediate. It measures the reactivity without deducting it from the period; it therefore gives the reactivity very precisely both for divergence and convergence even through in this latter case the period does not in fact exist. The equipment makes it possible to calibrate very rapidly the control rods of a reactor (the rod-drop method), to measure the reactivity of an experiment in the core, and to measure certain temperature effects. It is also possible by introducing a control into the core at a measured rate, to deduce directly its efficiency curve. (author) [French] Cet appareil est destine a mesurer instantanement la reactivite positive ou negative d'un reacteur a uranium modere a l'eau legere, a condition de partir de l'etat critique du reacteur, ou eventuellement d'un etat sous-critique deja connu. De legeres modifications permettent de l'adapter a un autre type de moderateur. C'est un calculateur analogique, qui inverse purement et simplement la fonction de transfert du reacteur; ce n'est donc pas un simulateur de pile dont la tension de sortie est asservie a la puissance du reacteur pour elaborer la reactivite; le principe du

  14. The Use of Research Reactors and Short-Lived Isotopes in the Study of Nuclear-Reactor Fuel Materials; Emploi de Reacteurs de Recherche et de Radioisotopes de Courte Periode dans l'Etude des Combustibles pour Reacteurs Nucleaires; ИСПОЛЬЗОВАНИЕ ИССЛЕДОВАТЕЛЬСКИХ РЕАКТОРОВ И КОРОТКОЖИВУЩИХ ИЗОТОПОВ ПРИ ИЗУЧЕНИИ ТОПЛИВНЫХ МАТЕРИАЛОВ ДЛЯ ЯДЕРННХ РЕАКТОРОВ; Empleo de Reactores de Investigacion y de Isotopos de Periodo Corto en el Estudio de Combustibles Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Elleman, T. S.; Townley, C. W.; Sunderman, D. N. [Battelle Memorial Institute, Columbus, OH (United States)

    1963-03-15

    recueillant les produits de filiation radioactifs non volatils sur un fil charge, pour les soumettre ensuite a une analyse radiochimique, tandis que les gaz de fission ayant une periode plus longue'(krypton-85m, krypton-87, krypton-88, iode-131, xenon-133 et xenon-135) sont absorbes sur les pieges a charbon de bois refroidis, separes par elution, sur une colonne chromatographique, en fractions d'iode, de krypton et de xenon, puis analyses par spectrometrie gamma. Les produits de fission non volatils, degages par l'echantillon, se deposent sur un piege voisin constitue par une feuille metallique qui peut etre retire, aux fins d'analyse, a n'importe quel moment pendant l'irradiation. La liberation des produits de fission, observee dans differentes conditions d'irradiation, peut etre ou non fonction de la concentration; il peut souvent y avoir liberation de certains elements predominants, liberation rapide de produits de fission lors de changements de temperature ou degagement de gaz de fission apres l'arret du reacteur. L'application de cette technique permet d'obtenir des renseignements fondamentaux sur le fonctionnement de prototypes de combustible, sans qu'il soit necessaire d'employer de grands reacteurs pour les essais ou des installations souterraines speciales pour la manipulation des echantillons irradies. (author) [Spanish] Los reactores de investigacion pueden ser muy utiles para estudiar la movilidad de los productos de fision en nuevos tipos de materiales combustibles para reactores porque permiten trabajar en condiciones ambientales analogas a las que reinan durante la utilizacion normal del combustible, a la vez que permiten regular exactamente los parametros y variar en gran medida los disenos experimentales. Si se alteran las condiciones de la irradiacion y analizan cuantitativamente los productos de fision de periodo corto que la muestra libera, es posible determinar los mecanismos de desprendimiento de los productos de fision y la relacion que guardan con las

  15. 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

  16. 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

    organic liquids under irradiation. The equipments of Melusine are now being modernized. Siloe, another swimming-pool reactor, has been operating at 15 MW since the end of 1963. The performances achieved constitute a considerable progress in the field of swimming-pool reactors, since the fluxes obtained with Siloe are in the same order of magnitude than those which needed till now a tank type structure, whose numerous disadvantages in the fitting of experiments are well known. Siloe will be used mostly in the study of structural materials, graphites, refractory fissile materials, and for solid state physics. The reactor Pegase, in service in the Cadarache Nuclear Centre since 1963, is intended solely for testing full-scale fuel elements of EDF and EL 4 types. The current programme, for the eight loops of the reactor, covers the elements for the reactors EDF 2, EDF 3 and EL 4. New loops are in the course of being studied for the fuel elements of the EDF 4 and EDF 5 reactors. The general line of the CEA programmes has shown up the considerable need for fast neutron irradiations. The reactor Osiris which is in the course of being constructed will serve to complement the CEA's equipment in this field and at the same time fill the gap which will be left in the near future in the Centre de Saclay by the closing down of EL 2. Osiris is a light water reactor whose special structure will allow it to function at 50 MW without the disadvantages usually associated with the presence of a heavy waterproof tank. This reactor, which should be put into service in 1966, is mainly intended for the investigation of structural materials, graphite and refractory fuels; it will also serve to increase the production of high specific activity isotopes, and to develop activation analysis techniques. (authors) [French] Les auteurs examinent successivement les differents reacteurs de recherche en service dans les Centres du Commissariat a l'Energie Atomique. Ils retracent brievement l'histoire de ces

  17. 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

  18. 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

  19. 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

  20. 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

  1. Materials flow management in the metal industry. Design and techno-economical analysis of industriyl recycling concepts; Stoffstrommanagement in der Metallindustrie. Zur Gestaltung und techno-oekonomischen Bewertung industrieller Recyclingkonzepte

    Energy Technology Data Exchange (ETDEWEB)

    Haehre, S.

    2000-07-01

    Materials flow management is an attempt to cope with excessive consumption of resources and with the large-scale release of pollutants, both of which are hazards of our industrialised society. It is necessary to analyse materials and energy flow along interconnected product and process chains to develop suitable concepts, e.g. for innovative environmental protection and recycling measures. A planning instrument is designed for this purpose which is based on a combination of flowsheeting models common in chemical process engineering and material flow networks based on Petri nets. Its application enables computer-assisted design and techno-economic evaluation of industrial recycling concepts, so that decision makers in enterprises and administrations will be given a tool for in-house and external material flow management. In the implementation stage, the flowsheeting system 'Aspen Plus' and the eco-banancing tool 'Umberto' were used, and exemplary problems of material flow management in the steel and zinc industry were investigated. [German] Zur Ueberwindung negativer Umwelteinfluesse der Industriegesellschaft, die inbesondere aus dem Ressourcenverbrauch und der Freisetzung grosser Schadstoffmengen resultieren, werden Massnahmen im Sinne des Stoffstrommanagements gefordert. Die Entwicklung geeigneter Konzepte, etwa zum Einsatz innovativer Umweltschutz- und Recyclingmassnahmen, macht die Analyse von Stoff- und Energiestroemen entlang vernetzter Produkt- und Prozessketten erforderlich. Hierfuer wird ein Planungsinstrument konzipiert, das auf einer Kombination verfahrenstechnischer Flowsheeting-Modelle und Petri-Netz-basierter Stoffstromnetze beruht. Sein Einsatz ermoeglicht computergestuetzt die Gestaltung und techno-oekonomische Bewertung industrieller Recyclingkonzepte, so dass Entscheidungstraeger in Unternehmen und Behoerden ein Hilfsmittel zum betrieblichen und betriebsuebergreifenden Stoffstrommanagement erhalten. Bei der Implementierung

  2. 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

  3. 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

    Experience of the CEA heavy water reactors and a systematic study of the radiolytic decomposition of water in the core of swimming-pool reactors are described. Setting up of reactivity control by physico-chemical methods. Reactivity control by homogeneous poisoning of the reactor A comparison of the evolution of xenon poisoning with the residual anti reactivity of the poison in solution during its nuclear consumption establishes the programme which must govern the variation in its concentration if the exact compensation is to be produced The behaviour of the poison towards the reactor materials under the particular operational conditions must be taken into account. Radiolytic decomposition of water in the reactors in the presence of soluble poisons: A study of the effect of certain chemically inert salts, present in small concentrations in the water, on its radiolytic decomposition rate, has led to some new results which are discussed. The choice of a soluble poison is justified on the basis of the above results. Reactivity control by the use of a gaseous absorbent The use of a gas control rod circuit for compensation purposes, in place of solid control rods is described. The use of soluble poisons in the moderator to compensate the xenon effect, and of a gaseous absorbent in a circuit known as a gas control rod form original aspects of the reactivity control in the reactor EL 4. (authors) [French] L'observation du comportement de l'eau et de l'eau lourde dans les reacteurs en exploitation, contribue au fonctionnement sur de ceux-ci et oriente certaines etudes relatives aux techniques de controle de la reactivite par mise en oeuvre de poisons solubles. L'utilisation de poisons nucleaires dissous dans l'eau du reacteur entraine une pollution chimique de celle-ci. Les conditions d'emploi permettant d'eviter les effets indesirables de cette pollution sont etudiees. Les problemes analytiques - bien qu'importants - ne sont pas abordes dans le cadre de la communication

  4. 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

  5. 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

  6. 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

  7. 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

  8. Contribution to the study and use of ionisation chambers for nuclear reactor control (1965); Contribution a l'etude et a l'utilisation des chambres d'ionisation pour le controle des reacteurs nucleaires (1965)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-02-15

    high-power reactors. (author) [French] Les chambres d'ionisation sont actuellement les detecteurs les mieux adaptes au controle des reacteurs nucleaires par des mesures neutroniques. Nous avons cru bon de rappeler quelques generalites concernant la dynamique des reacteurs, les differents procedes de detection des neutrons, le fonctionnement des chambres d'ionisation et les methodes de mesure utilisees. Notre contribution aux techniques de controle des reacteurs consiste d'une part en une tentative de synthese des facteurs intervenant dans le fonctionnement des chambres d'ionisation, l'etude de ces facteurs, et d'autre part l'elaboration de chambres d'ionisation a fission et a bore permettant de suivre la marche d'un reacteur du demarrage jusqu'a la puissance maximale. Dans le domaine des chambres a fission, nous avons en particulier ameliore les techniques de depot d'oxyde d'uranium sur l'aluminium et realise la mise au point de depots par electrolyse sur d'autres metaux: acier inoxydable, cuivre, molybdene, nickel, tantale, titane, kovar, tungstene et beryllium. Nous avons elabore plusieurs types de chambres a fission servant au demarrage des reacteurs: un type de performances moyennes actuellement utilise dans les piles francaises un type a haute sensibilite un type a haute temperature qui a fonctionne jusqu'a 600 deg. C. En ce qui concerne les chambres a bore, nous avons etudie les perturbations apportees dans les mesures par l'exposition des chambres a d'importants flux de neutrons et a un rayonnement {gamma} intense. Cette exposition produit une modification des proprietes des materiaux constitutifs et la production dans les chambres d'un bruit de fond qui peut gener considerablement les mesures neutroniques. Nous avons montre que la technique de compensation permettait de limiter l'importance de ce bruit de fond et d'augmenter ainsi la plage de fonctionnement des chambres d'ionisation classiques destinees aux mesures de puissance. Enfin, nous avons realise deux

  9. Study of radioactivity diffusion for bitumen-coated blocks produced by an industrial coating plant; Etude de la diffusion de la radioactivite de blocs d'enrobes bitumineux en provenance d'un atelier d'enrobage industriel

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    The solidification by bitumen of chemical coprecipitation sludges from the Marcoule waste treatment station has been studied in the laboratory and has led to the construction of an industrial coating plant. The quality of the coated material obtained has been controlled by the lixiviation test carried out with ordinary water and with sea-water on 45 ml laboratory samples and on industrial coated blocks of 150 litres. Tests on blocks of such a size have necessitated the installation of three special tanks. Two, each of 2000 litres capacity, contain ordinary and sea-water which was continuously recycled at a rate of 2.5 cm/hr and renewed periodically. In the third tank having a capacity of 11000 litres, the coated block was buried in earth and sprinkled with ordinary water with a view to studying the migration of radioelements in soil. The results of these tests confirm those obtained during the laboratory experiments. (authors) [French] La solidification par le bitume des boues de coprecipitation chimique de la station de traitement des effluents du Centre de Marcoule, etudiee en laboratoire, a conduit a la realisation d'une installation industrielle d'enrobage. La qualite de l'enrobe obtenu a ete controlee par le test de lixiviation qui a ete effectue en eau ordinaire et en eau de mer sur des echantillons de laboratoire de 45 ml et sur des blocs d'enrobe industriel de 150 litres. L'experimentation sur des blocs de telles dimensions a necessite l'installation de 3 cuves speciales. Deux, d'une capacite de 2000 litres, contiennent de l'eau ordinaire et de l'eau de mer recyclees en permanence a une vitesse de 2.5 cm/h et renouvelees periodiquement. Dans la 3eme cuve d'une capacite de 11000 litres, le bloc d'enrobe a ete enfoui dans de la terre et asperge d'eau ordinaire afin d'etudier la migration des radioelements dans le sol. Les resultats de ces essais confirment ceux obtenus au cours des tests de laboratoire. (auteurs)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. Le néo-corporatisme réinterpellé : analyse comparée de deux politiques d'accès à l'emploi, l'apprentissage industriel en Belgique et le contrat de qualification "jeunes" en France

    OpenAIRE

    Levêque, Audrey

    2006-01-01

    L'étude comparative de l'apprentissage industriel en Belgique francophone et du contrat de qualification "jeunes" en France permet de réinterpeller le modèle néo-corporatiste d'Etat social. L'analyse de l'émergence et de la mise en oeuvre de ces deux politiques publiques d'accès à l'emploi pour les jeunes peu qualifiés permet de comprendre comment elles sont réappropriées par les acteurs de terrain et ce, plus particulièrement dans la branche de la métallurgie. Si ces politiques s'adressent t...

  18. Militärisch-Industrieller Komplex?

    DEFF Research Database (Denmark)

    Nørby, Søren

    2015-01-01

    Aufgedeckte Fehlentwicklungen und Missstände bei der Beschaffung von Rüstungsgütern für die Bundeswehr schrecken die deutsche Öffentlichkeit immer wieder – wenn auch nur kurzfristig – auf. Dabei werden in den Diskussionen häufig Argumentationsmuster und Begriffe aus der Zeit des Kalten Krieges...... verwendet, die auch damals schon nur bedingt der Realität entsprachen. So ist zum Beispiel der weithin bekannte Begriff des »Militärisch-Industriellen Komplexes« für die Verhältnisse in Deutschland bis zum heutigen Tag unpassend. Dennoch ist weder in der Geschichts-, noch in der Politik- oder in den...

  19. Development of a power-period calculation unit for nuclear reactor Control; Etude et realisation d'un ensemble de calcul puissance periode pour le controle d'un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-10-01

    The apparatus studied is a digital calculating assembly which makes it possible to prepare and to present numerically the period and power of a nuclear reactor during operation, from start-up to nominal power. The pulses from a fission chamber are analyzed continuously, using real time. A small number of elements is required because of the systematic use of a calculation technique comprising the determination of a base 2 logarithm by a linear approximation. The accuracy obtained for the period is of the order of 14%; the response time of the order of the calculated period value. An approximate value of the power (30%) is given at each calculation cycle together with the power thresholds required for the control. (author) [French] L'appareil etudie est un ensemble de calcul digital permettant d'elaborer et d'afficher numeriquement la periode et la puissance, d'un reacteur nucleaire lors de son fonctionnement depuis le demarrage jusqu'a la puissance nominale. Il traite en temps reel, de facon continue, les impulsions en provenance d'une chambre de fission. Grace a l'utilisation systematique d'une technique de calcul, la determination d'un logarithme a base 2 par approximation lineaire, un nombre reduit d'elements est utilise. La precision obtenue sur la periode est de l'ordre de 14 pour cent, le temps de reponse de l'ordre de la valeur de la periode calculee. Un ordre de grandeur de la puissance (30 pour cent) est donne a chaque cycle de calcul ainsi que des seuils de puissance necessaires au controle. (auteur)

  20. 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

  1. 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

  2. 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

  3. 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)

  4. Initial Operating Experience with the ''NPD'' Reactor; Experience recueillie pendant les premiers mois de fonctionnement du reacteur NPD; Pervyj opyt po ehkspluatatsii reaktora NPD; Experiencia inicial de funcionamiento del reactor NPD

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, L. G. [Hydro-Electric Power Commission of Ontario, Toronto, Ontario (Canada)

    1963-10-15

    Canada's first nuclear power station, the Nuclear Power Demonstration station (NPD), is intended to serve as a means of proof-testing the performance of the Canadian type of station using natural uranium as fuel and heavy water as moderator and coolant. It reached full power on 28 June 1962. Although designed for base-load operation it will, during the early stages, be operated part of the time on high-capacity.runs and part of the time on improvement periods. Progress has been favourable so far; the first high-capacity run of six weeks'duration yielded a capacity factor of 70%. Improvements already made have increased safety, improved performance and demonstrated potential methods of capital-cost reduction for future stations. For example, shaft seals on primary coolant pumps have been modified for better performance, freezer-type vapour recovery equipment has been replaced in favour of absorption columns to reduce heavy-water vapour loss, and flow limiters are being installed in sample lines to reduce losses of heavy water in the event of joint failures. During December 1962 two simultaneous leaks from the on-power refuelling machine led to an unusual sequence of events in which a considerable amount of hot high-pressure heavy water was spilled into the reactor vault where it suffered a slight downgrading in isotopic purity. It was upgraded and the reactor returned to operation by the end of the month. All safety devices operated correctly during the incident as did the provisions for containment of heavy water. (author) [French] La premiere centrale nucleaire du Canada, NPD, est une centrale de demonstration, qui doit servir a verifier les performances des reacteurs fonctionnant a l'uranium naturel et utilisant de l'eau lourde comme ralentisseur et comme fluide de refroidissement. Elle a atteint sa pleine puissance le 28 juin 1962 bien que concue pour etre exploitee comme centrale de base, elle fonctionnera au debut comme centrale d'appoint, ce qui permettra d

  5. 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)

  6. Bois-Noirs ore. Recovery of uranium of solutions from acid treatment. Results of industrial tests at the Gueugnon plant; Minerai des Bois-Noirs. Recuperation de l'uranium des solutions d'attaques acides. Resultats des essais industriels effectues a l'usine de Gueugnon

    Energy Technology Data Exchange (ETDEWEB)

    Le Bris, J

    1959-04-01

    Industrial-scale tests are reported of the efficiency of two recovery processes for the separation of uranium from sulfuric acid pickling solutions used on ore from Bois-Noirs, at the Gueugnon works. The final stage of each process is sodium uranate. The earlier part of the report deals with tests of the separation of uranium from foreign metals by fractional precipitation. The second part deals with the separation of uranium from these metals by carbonation of the solutions. (author) [French] Le present rapport concerne les essais industriels de deux procedes de recuperation de l'uranium de solutions d'attaque sulfurique du minerai des Bois-Noirs a l'usine de Gueugnon. Le stade final pour ces deux procedes etant l'uranate de sodium, une premiere partie est consacree aux essais de separation de l'uranium des metaux etrangers par precipitation fractionnee; une deuxieme partie est consacree aux essais de separation de l'uranium des metaux etrangers par carbonatation des solutions d'attaque du minerai. (auteur)

  7. 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

  8. 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

  9. 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)

  10. 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

  11. Preliminary studies of vanadium-base alloys intended for use in fabrication of cans for fast reactors; Etudes preliminaires sur les alliages a base de vanadium envisages pour la fabrication de gaines de reacteurs rapides

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-03-15

    Preliminary research has been carried out on a series of vanadium-based alloys: V, 0.5 per cent Si; V, 5 per cent Ca; V, 5 per cent Mo; V, 5 per cent Nb; V, 2 per cent Zr; V, 20 per cent Ti; V, 10 per cent Al; V, 10 per cent Sn and v, 10 per cent Ti liable to be used as canning material in fast reactors. The transformation by forging at about 1000 deg. C and rolling between 200 deg. C and room temperature is satisfactory for all types of alloys except V with 10 per cent Sn and V with 10 per cent Al. The mechanical properties deduced from tensile strength tests carried out on alloy samples annealed 1 hour at 1050 deg. C in a vacuum show that, generally speaking, the addition elements lead to an improvement in these properties as compared to those of pure vanadium. After undergoing corrosion tests in a liquid sodium loop purified by a cold trap, the alloys become brittle at room temperature. Only the vanadium containing 20 per cent Ti keeps its plastic properties. These alloys are covered by a layer of vanadium carbide VC. After undergoing treatment in a liquid sodium loop purified by a hot trap, all the alloys keep their good mechanical characteristics. The surface layer with which they are covered is composed of two vanadium carbides VC and {sub {gamma}}VC, and a vanadium sub-oxide VO{sub 0.9}. (author) [French] Des etudes preliminaires ont ete faites sur une serie d'alliages a base de vanadium: V-0,5 pour cent Si, V-5 pour cent Ca, V-5 pour cent Mo, V-5 pour cent Nb, V-2 pour cent Zr, V-20 pour cent Ti, V-10 pour cent Al, V-10 pour cent Sn et V-10 pour cent Ti susceptibles d'etre utilises comme materiau de gainage pour les reacteurs rapides. La transformation par forgeage a 1000 deg. C environ et laminage entre 200 deg. C et la temperature ambiante est satisfaisante pour toutes les nuances d'alliage sauf le V-10 pour cent Sn et le V-10 pour cent Al. Les proprietes mecaniques deduites des essais de traction realises sur des eprouvettes d'alliages recuits 1 heure a

  12. Effect of the plutonium isotopic composition on the performance of fast reactors; Effet de la composition isotopique du plutonium sur le rendement de reacteurs a neutrons rapides; Vliyanie izotopnogo sostava plutoniya na rabotu reaktorov na bystrykh nejtronakh; Efectos de la composicion isotopica del plutonio sobre el funcionamiento de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Yiftah, S [Israel Atomic Energy Commission (Israel)

    1962-03-15

    The isotopic composition of plutonium to be used as fuel for fast reactors will depend on the source of plutonium. In principle three different sources are possible: (a) production reactors; (6) thermal power reactors (using natural uranium or enriched uranium as fuel); (c) fast reactor blankets. In general, source (a) and to some extent source (c) will provide relatively 'clean' plutonium, that is mostly Pu{sup 239}, while plutonium from source (6) will be 'dirty' plutonium, that is plutonium rich in Pu{sup 240}, Pu{sup 241}, and Pu{sup 242}. The degree of 'dirtiness' will depend on the kind of reactor, amount of burn-up and in general on the irradiation history of the fuel. The question then arises, can one use as fuel for fast reactors any kind of plutonium? To investigate the effect of different isotopic composition of the plutonium fuel, in the metallic, oxide and carbide form, on the performance of fast reactors, a limited series of spherical geometry 16-group diffusion theory calculations were performed, using the 16-group cross-section set developed recently by Yiftah, Okrent and Moldauer and taking three different kinds of plutonium, starting with pure Pu{sup 239} and increasing the amount of higher isotopes. For the systems studied-800, 1500 and 2500-l core-volumes, which are typical for large fast power reactors-the result is, when one takes into account only the thermally fissionable isotopes Pu{sup 239} arid Pu{sup 241}, that the 'dirtier' the plutonium, the smaller the critical mass and the higher the breeding ratio. For the 1500-l reactor, taken as an example, it is further found that in the metallic, oxide and carbide plutonium fuels the reactivity change upon removal of 40% of the sodium initially present in the core is made more negative (or less positive) when the plutonium is richer in higher isotopes. (author) [French] La composition isotopique du plutonium qui doit etre utilise comme combustible dans des reacteurs a neutrons rapides depend de

  13. 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

  14. Etude d'un concept de coeur hybride refroidi a l'eau supercritique

    Science.gov (United States)

    Delattre, Baptiste

    Facing the current weather and energy global problem, Canada chose to develop a reactor cooled by water at supercritical conditions (SCWR). Inspired by the current CANDU-6 pressure tube technology, this concept should allow to save a substantial amount of efforts for developping a brand new kind of reactor by using the well-known pressure tube CANDU design. In fact, this type of reactor should be able to reach a better energy efficiency as well as other essential criteria about safety, security, non-proliferation... Nevertheless, there are still a lot of technology challenges to be dealt with to satisfy the differents obligations related to the use of supercritical water (SCW). Thus, materials to use remain undetermined because of a 25 MPa operating pressure and a 650.C temperature for the SCW coolant. Actually, materials in presence of SCW should be able to avoid too much corrosion and remain low neutrons absorbers. Additionnaly, the use of a light water coolant makes the neutronic absorption more important than in CANDU heavy-water cooled reactors. Additionally, a positive coolant void reactivity (CVR) and safety related problem remains among the challenges to overcome for developping a SCWR. Bringing about a solutions to all these problems remains very difficult and that's why some concessions on these criteria have to be made in order to achieve a viable reactor. This study presents some thougts and works in that direction. Originally developped in early studies about thermodynamic cycle optimization for a SCW power plant, a new hybrid reactor concept with two channels types has arise. To this purpose, we imagine a pressure tube core design but with two different types of channels: . Some channels have thermodynamic conditions where water goes through a supercritical state. . The other channels have "CANDU like" thermodynamic conditions allowing the flow of pressurized light water under sub-critical conditions. These two kinds of features should mitigate the drawbacks and constraints of a whole SCWR core while keeping some of the benefit of the SCW use (improved energy efficiency). The aim of this study is to determine whether or not this kind of hybrid core with two types of channels is viable. To this purpose we used neutronics as thermalhydraulic models. First of all, in this thesis, we describe the SCWR-development history and specific back- ground with bibliography commentaries and a global description of the hybrid reactor developped. Secondly, we provide a brief theorical review of the neutronics and thermalhydraulics concepts used in the DRAGON, DONJON and ARTHUR calculation codes. Afterwards we will be considering some cell and core models and their description. Among others, some coupled neutronics-thermalhydraulic calculations have been done. This is followed by a presentation of the results we obtained and a discussion of the viability of our reactor. Finallly, we conclude.

  15. 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

  16. The 'Reacteur Jules Horowitz': The preliminary design

    International Nuclear Information System (INIS)

    Ballagny, A.; Frachet, S.; Minguet, J.L.; Leydier, C.

    1999-01-01

    The 'Reactor Jules Horowitz' is a new research reactor project dedicated to materials and nuclear fuels testing, the location of which is foreseen at the CEA-Cadarache site, and the start-up in 2008. The launching of this project arises from a double finding: 1) the development of nuclear power plants aimed at satisfying the energy needs of the next century cannot be envisaged without the disposal of experimental reactors which are unrivalled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. 2) the present park of experimental reactors is 30 to 40 years old and it is advisable to examine henceforth the necessity and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a reflection on the mid and long term irradiations needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfil the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The selected reactor project, among several different concepts, is finally a light water open pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows many irradiations in the core and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the main items, and have permit to determine: the core and fuel designs, with added pressurisation; the different core surrounding structures in connection with the core studies; overall layout of the reactor/auxiliary pools and reactor building. (author)

  17. 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

  18. 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

  19. Neutron Tests at the Start-Up of EDF1; Les essais neutroniques au demarrage du reacteur EDF1; Nejtronnye izmereniya pri puske reaktora EDF1; Ensayos neutronicos efectuados durante la puesta en marcha del reactor EDF1

    Energy Technology Data Exchange (ETDEWEB)

    Teste du Bailler, A. [Centre d' Etudes Nucleaires de Saclay (France); Janin, R. [Electricite de France, Paris (France)

    1963-10-15

    A series of neutron measurements, for which the principal experimental methods perfected at the Marcoule reactors were used, was carried out at the start-up of EDF1. The measurements were designed mainly to determine the efficiency of the control rods at different depths of insertion. From them a rod-withdrawal configuration was derived which allowed full-power operation without infringing certain limitations on cladding and gas temperatures. At the same time flux measurements were made for different shim-rod positions and different absorber loadings in certain channels. These measurements based on preliminary two-dimensional calculations, were obtained by activation of point detectors,using the standard technique of air poisoning. At certain temperature plateaus (up to 140{sup o}C), measurements of temperature coefficients and control-rod efficiency were made. Spectrum index measurements were carried out at the same time by activation of appropriate detectors (U, Pu, Lu, Mn, In, Au). The oscillation technique was used to measure the efficiency of certain shim rods. Finally, fast-neutron measurements were made in connection with studies of shielding and graphite damage. (author) [French] Une serie de mesures neutroniques utilisant les principales methodes experimentales mises au point sur les reacteurs de Marcoule a ete effectuee au cours du demarrage d'EDF1. Les mesures portent essentiellement sur l 'efficacite des barres de controle a differents enfoncements. On en deduit une configuration de montee des barres permettant d'obtenir la pleine puissance en respectant certaines limitations sur les temperatures de gaines et de gaz. Parallelement des mesures de flux ont ete faites pour differentes positions des barres de compensation et pour divers chargements d'absorbants dans certains canaux, suivant des calculs previsionnels a deux dimensions. Ces mesures sont obtenues par activation de detecteurs ponctuels, au moyen de la technique classique par empoisonnement a l

  20. Advanced epithermal thorium reactor (AETR) physics; Physique d'un reacteur au thorium, a neutrons epithermiques, de type perfectionne (AETR); Fizika usovershenstvovannog o nadteplovogo torievogo reaktora; Fisica del reactor epitermico de tipo avanzado, alimentado con torio (AETR)

    Energy Technology Data Exchange (ETDEWEB)

    Campise, A. V. [Atomics International, Canoga Park, CA (United States)

    1962-03-15

    'etude de cet ensemble a mis en relief l'importance des donnees relatives aux sections efficaces et de l'interpretation theorique des resultats experimentaux pour l'etude d'un reacteur au thorium de type perfectionne. La precision des methodes analytiques employees a ete demontree lors de l'analyse des resultats experimentaux obtenus avec le ZPR-III. L'auteur compare trois configurations pour le transfert de chaleur, en utilisant le temps de doublement comme parametre d'optimisation. Les effets de la production de {sup 233}Pa et d'isotopes de l'uranium sur le bilan neutronique, les taux possibles de surgeneration et les caracteristiques de la combustion sont evalues en tenant compte de l'imprecision des sections efficaces nucleaires. (author) [Spanish] El autor estudia la concepcion del reactor AETR desde el punto de vista de la teoria actual de los parametros nucleares y del balance neutronico. En los sistemas moderados por grafito examina el efecto de la captura por resonancia en el torio para energias medias de absorcion del orden de 0,10 a 100 keV. Aplica formulas de resonancia angosta y de resonancia ancha para obtener la integral de resonancia efectiva en funcion de la temperatura, correspondiente a las barras de torio, y dicho parametro se expresa como secciones eficaces equivalentes de varios grupos. Se ha disenado y construido un conjunto critico para obtener datos nucleares indispensables en la gama de energias intermedias. En el diseno nuclear de dicho conjunto, se ha tenido particularmente en cuenta la importancia de los datos relativos a secciones eficaces y la interpretacion teorica de estos resultados experimentales, cosas ambas relacionadas con el diseno del reactor AETR. La precision de los metodos analiticos ha quedado demostrada por el estudio de los resultados experimentales obtenidos con el reactor ZPR-III. Se comparan tres sistemas de transmision de calor utilizando el tiempo de duplicacion como parametro optimo. Se estudia el efecto de la formacion

  1. L’instabilité organisationnelle des districts industriels. Dynamiques des transformations internes d’un village de métier au nord du Viêt Nam The Organizational Instability of Industrial Districts: Dynamic Internal Transformations of a Craft Village in Northern Vietnam

    Directory of Open Access Journals (Sweden)

    Nguyen Quy Nghi

    2012-11-01

    Full Text Available Le village de métier est un sujet d’étude attractif, mais rarement traité comme une forme d’agglomération industrielle ou abordé sous l’angle de la sociologie industrielle. Le développement actuel des villages de métier montre bien la présence intensive des entreprises de tailles variées et l’implication d’autres acteurs en dehors du milieu de production (par exemple les autorités locales, le milieu de l’éducation. L’introduction d’acteurs nouveaux implique le changement des relations existantes, contribuant ainsi à la mise en place d’une nouvelle configuration. L’objectif de cet article est de montrer que l’organisation des villages de métier, qui est conçue comme un type de district industriel, est marquée par l’instabilité ou le changement permanent de la configuration générale, mais aussi par une transformation interne des acteurs au sein du district. À travers une étude de cas à Bát Tràng – un centre de production céramique au nord du Viêt Nam – cet article a pour objectif de montrer comment des entreprises familiales se réorganisent en structure plus formalisée, comment elles mobilisent du capital social pour régler leurs difficultés et les stratégies d’innovation qu’elles développent.Craft village is an attractive research subject, however it is rarely treated as a form of industrial agglomeration or studied from industrial sociology perspective. The current development of craft villages demonstrates the intensive presence of entreprises of all sizes and involvement of non productive actors (e.g. local authorities, education sector. The introduction of new actors therefore implies a change of existing relationships, contributing to the formulation of new configuration. The objective of this paper is to demonstrate that the organization of craft villages, as a type of industrial district, is marked by instability or permanent change in the overall configuration, but also by

  2. 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

  3. 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)

  4. 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)

  5. 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.

  6. 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)

  7. Methanization of industrial liquid effluents; Methanisation des effluents industriels liquides

    Energy Technology Data Exchange (ETDEWEB)

    Frederic, S.; Lugardon, A. [Societe Naskeo Environnement, 92 - Levallois-Perret (France)

    2007-09-15

    In a first part, this work deals with the theoretical aspects of the methanization of the industrial effluents; the associated reactional processes are detailed. The second part presents the technological criteria for choosing the methanization process in terms of the characteristics of the effluent to be treated. Some of the methanization processes are presented with their respective advantages and disadvantages. At last, is described the implementation of an industrial methanization unit. The size and the main choices are detailed: the anaerobic reactor, the control, the valorization aspects of the biogas produced. Some examples of industrial developments illustrate the different used options. (O.M.)

  8. Wachstumspolitik in Industrieländern / Manfred O. E Hennies

    Index Scriptorium Estoniae

    Hennies, Manfred O. E

    2003-01-01

    Praktilise majanduspoliitika jaoks on majanduskasv vältimatu eeltingimus, et säilitada rahvusvahelist konkurentsivõimet, lahendada tööhõive ja jaotusprobleeme, tõsta kõikide sotsio-ökonoomiliste gruppide elustandardit, samuti reaalmajanduslikult finantseerida hädavajaliku arenguabi andmist puudustkannatavatele arengumaadele

  9. Impacts des effluents liquides industriels sur l'environnement urbain ...

    African Journals Online (AJOL)

    industrialisation l'un des maillons essentiels du développement, après le succès de l'agriculture. Mais la mise en oeuvre de cette politique pose parfois des problèmes d'environnement à Abidjan. La présente étude a pour objectif principal ...

  10. Epuration des effluents industriels par électroflottation Belkacem ...

    African Journals Online (AJOL)

    étudié la séparation de quelques métaux lourds tels que le fer, le nickel, ... the separation of some heavy metals such as iron, nickel, copper, zinc, lead and ... généralement basées sur l'adsorption (par ... Différents essais de concentration en.

  11. qualite des rejets industriels textiles et ses repercussions

    African Journals Online (AJOL)

    client

    plantules. Les plants de riz croissent normalement jusqu'à la maturation. ..... endogenous and exogenous abscisic acid». Revue ... [28] - A.J. JOSHI, H. HINGLAJIA, « Effects of chloride and sulphate on seed germination in Prosopis juliflora.

  12. Systèmes de recommandation dans des contextes industriels

    OpenAIRE

    Meyer , Frank

    2012-01-01

    This thesis deals with automatic recommendation systems. Automatic recommendation systems are systems that allow, through data mining techniques, to recommend automatically to users, based on their past consumption, items that may interest them. These systems allow for example to increase sales on e-commerce websites: the Amazon site has a marketing strategy based mainly on the recommendation. Amazon has popularized the use of automatic recommendation based on the recommendation function that...

  13. Les risques industriels et le prix des logements

    DEFF Research Database (Denmark)

    Grislain-Letrémy, Céline; Katossky, Arthur

    2013-01-01

    Le prix des logements peut diminuer du fait de la proximité d’installations industrielles. Cet effet dépend de la perception du risque par les riverains et est donc potentiellement modifié par des événements changeant la perception du risque, tels que les plans de prévention des risques technolog......Le prix des logements peut diminuer du fait de la proximité d’installations industrielles. Cet effet dépend de la perception du risque par les riverains et est donc potentiellement modifié par des événements changeant la perception du risque, tels que les plans de prévention des risques...... proximité d’installations industrielles dans les agglomérations françaises de Bordeaux, Dunkerque et Rouen. La méthode des prix hédoniques permet d’estimer l’effet de la proximité des usines sur les prix des logements. Les résultats suggèrent que les écarts de prix ne sont modifiés ni par les incidents...

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor; Effets Cavitaires dans la Deuxieme Charge du Reacteur a Eau Lourde Bouillante de Halden (HBWR); Ehffekty pustotnoj reaktivnosti vo vtoroj zag HBWR; Effectos de Cavitacion en la Segunda Carga del Reactor de Agua Pesada Hirviente de Halden (HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Lunde, J. E. [OECD Halden Reactor Project (Norway)

    1964-02-15

    calculation of void effects. Preliminary theoretical comparisons are made for these experiments. Two-group diffusion theory is applied, and the conclusion can be drawn that fair agreement is obtained between theory and experiment for the perturbations in the lattice parameters for a void fraction equal to one, both at low and high temperatures. For intermediate void fractions, however, somewhat less satisfactory agreement is found. (author) [French] L'auteur a mesure, aussi bien lors d'experiences avec vide simule a puissance nulle que dans les conditions normales de puissance, l'effet cavitaire, provoque par l'ebullition qui se produit a l'interieur des canaux du refroidisseur, dans la deuxieme charge de HBWR. Les experiences avec vide simule ont consiste a mesurer les effets que produit sur la reactivite le fait d'enfoncer a des profondeurs differentes des tubes plus ou moins vides a paroi mince. Les tubes ont ete places en plusieurs endroits entre les barres, dans une seule cartouche formee de sept barres en grappe et pratiquement identique aux cartouches de combustible de la deuxieme charge. Cette experience permet de determiner comment la reactivite varie en fonction du volume cavitaire relatif et de l'emplacement des bulles dans le canal du refroidisseur. L'experience a ete effectuee dans le reacteur NORA de puissance zero, avec un coeur compose de 36 cartouches de la deuxieme charge de HBWR et dans une geometrie de reseau identique a celle de ce reacteur. L'auteur a observe comment l'effet cavitaire variait avec la temperature dans un ensemble de puissance zero avec le cceur a 100 cartouches de HBWR. Dans une seule cartouche, il a abaisse le niveau de l'eau a l'interieur du canal de refroidissement a des niveaux differents et mesure l'effet de cette perturbation sur la reactivite a differentes temperatures comprises entre 50 et 220 Degree-Sign C. L'auteur a mesure l'effet cavitaire, a l'interieur de HBWR et dans les conditions de puissance, en fonction de la puissance

  20. 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

  1. Measurements with a Pulsed and Modulated Source in a Reactor; Mesures au Moyen d'une Source Pulsee et Modulee dans un Reacteur; Izmereniya v reaktore s pomoshch'yu impul'snogo i moduliruemogo is tochnika; Mediciones Efectuadas en Reactor con una Fuente Pulsada y Modulada

    Energy Technology Data Exchange (ETDEWEB)

    Rotter, W. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

    1965-10-15

    a digital computer [French] Un generateur dont le debit neutronique est variable selon une fonction du temps quelconque a ete mis au point par les Laboratoires de recherches Philips. Son utilite pratique dans le domaine de la physique des reacteurs a ete demontree par une serie de mesures effectuees dans le reacteur BRO2 a fotat sous-critique. Sa bonne stabilite, la possibilite de faire varier brusquement l'intensite neutronique, de puiser le debit ou de le moduler sinusoldalement, rend ce generateur tres souple. Il permet de determiner la reactivite ({rho} = {Delta}k/{beta}) et le temps de vie des neutrons ( Script-Small-L /{beta}) d'apres differentes methodes independantes. Une comparaison exacte de ces methodes est possible puisqu'elles peuvent etre employees sans modifier les conditions de mesure. On a determine: a) {rho} sur la base des neutrons retardes, par une reduction instantanee du debit de neutrons; b) {rho} sur la base des neutrons instantanes par des bouffees de neutrons; c) Script-Small-L /{beta} par combinaison de a) et b) pour 0,5 $ < {rho} <2 $; d) Script-Small-L /{beta} sur la base de la fonction de transfert du reacteur pour une source modulee. Les fonctions de transfert pour un oscillateur de reactivite et pour une source modulee sinusoldalement sont discutees. Il est montre que la mesure de Script-Small-L /{beta} est possible pour 0,1 $ < {rho} < 10 $ en utilisant une source modulee. La meme methode fournit aussi la reactivite a l'aide du rapport des neutrons instantanes aux neutrons retardes pour une frequence optimale, pratiquement independamment des donnees relatives aux neutrons retardes et de la valeur de f Script-Small-L /{beta}. Par accumulation d'un grand nombre de cycles dans l'analyseur multicanal, la statistique peut etre amelioree pour chaque methode. Le debit du generateur etant bien sinusoiedal, la reponse du reacteur peut etre integree sur chaque quart d'une periode, etant donne que la chafhe de mesure est pilotee par le

  2. 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

  3. 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

  4. The effective lifetime and temperature coefficient in a coupled fast-thermal reactor; Temps de vie effectif et coefficient de temperature dans un reacteur a couplage neutrons rapides-neutrons thermiques; Ehffektivnyj srok zhizni i temperaturnyj koehffitsient nejtronov v dvoyakom reaktore na bystrykh i teplovykh nejtronakh; Vida efectiva y coeficiente de temperatura en un reactor con acoplamiento rapido-termico

    Energy Technology Data Exchange (ETDEWEB)

    Haefele, W. [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    The theory of coupled systems was extensively developed by Avery and co-workers at the Argonne National Laboratory. One of the main points of interest in a coupled system is the larger effective lifetime of neutrons. The effect of the thermal component acts as a sort of neutron-delayer. As in the theory of delayed neutrons the delaying effect disappears if the reactivity worth is high enough to make the fast component critical by itself. In the study a coupled reactor is considered where the fast component suffers a sudden reactivity step {alpha}{sub 0}. Because of the increasing power-level the temperature rises and two temperature coefficients start to work: the temperature coefficient of the fast component and the temperature coefficient of the thermal component. The problem is considered with one group of delayed neutrons (in the ordinary meaning). A formalism is given to express the effective lifetime and temperature coefficient during the different stages of the excursion. Excursions for different {alpha}{sub 0} are given so that the limit of fast-reactor kinetics is reached. (author) [French] La theorie des systemes a couplage a ete mise au point par Avery et ses collaborateurs au Laboratoire national d'Argonne. L'une des caracteristique les plus interessantes d'un systeme a couplage est que le temps de vie effectif des neutrons est plus long. L'effet de la partie thermique contribue en quelque sorte a retarder les neutrons. Comme dans la theorie des neutrons retardes, l'effet de retardement disparait lorsque la reactivite a une valeur suffisamment elevee pour rendre la partie rapide critique par elle-meme. L'auteur du memoire considere un reacteur a couplage dont la partie rapide subit un saut instantane de reactivite, {alpha}{sub 0}. La temperature s'eleve a cause de l'augmentation de puissance et deux coefficients de temperature commencent a s'appliquer: le coefficient de temperature de la partie rapide et le coefficient de temperature de la partie

  5. 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

  6. 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

    The high thermal fluxes, which are usual in the latest materials testing reactors, impose suitable paths for uniform heat transfer and a reliable heat removal avoiding bulk-vapour formation. Furthermore, to control the over-all swelling and reactor fuel behaviour, water-gap measurements will also be performed in post-irradiation experiments on spent fuel elements. For that purpose, a probe for measuring the 3-mm water-gap of the BR-2 fuel element over a 1-m length, based on the principle of ultrasonics, has been developed. In the case of post-irradiation experiments, the measuring probe should operate in a fuel element by being immersed in a water pool at a depth of at least 6 m. The probe can withstand prolonged immersion in water and is not affected by normal gamma-irradiation doses. Although operating on the usual pulse-reflection method, the system allows emitted and reflected pulses to be separated by a 10-MHz ferro-electric crystal with high inherent energy dissipation. Oscilloscope read-out can be used, whereby the time is displayed on the horizontal axis, the scanning speed being adjusted to bear a direct relation to the velocity of wave propagation, i.e. the gap distance. This type of read-out Is satisfactory where the number of measurements is restricted, but chart recorder read-out is obviously desirable. In this case, emitted and reflected pulses are shaped and fed to a time-voltage converter using transistor logic techniques. The instrument allows continuous adjustment of output zero for any arbitrary gap distance between 2 and 4 mm thereby permitting zero-centre recording. Furthermore, any desired 100-{mu}m gap distance variation can give a stable 1-V output voltage to a recorder. An accuracy of 5-{mu}m gap-distance variation is easily obtained. Several fuel elements have been measured. The results and reproducibility were very satisfactory. (author) [French] Etant donne que dans les plus recents reacteurs d'essai de materiaux les flux thermiques sont

  7. 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

  8. The Development of Materials for Application to Control Rod Systems in Graphite moderated Reactors; Mise au Point de Materiaux pour les Dispositifs de Controle a Barres, Utilbes dans les Reacteurs Ralentis au Graphite; Razrabotka materialov , primenyaemykh v sistemakh upravlyayushchikh sterzhnej v reaktorakh s grafitovym zamedlitelem; Perfeccionamiento de Materiales Aplicables a las Barras de Control en los Reactores Moderados por Grafito

    Energy Technology Data Exchange (ETDEWEB)

    Wade, G. E.; Kempf, F. J. [Hanford Atomic Products Operation, General Electric Company, Richland, WA (United States)

    1964-06-15

    Material problems associated with the control- and safety-rod systems for graphite moderated, tube-type reactors can be divided into two categories: control materials and operating-channel liner materials. The control materials, such as boron or gadolinium, can be integral with the rod sheath, as in the boron stainless steel used for safety rods. Another approach is the enclosure of a boron-containing sintered compact, such as B{sub 4}C-graphite or B{sub 4}C-aluminium, in a metallic sheath. Rods of the latter type are adaptable for control purposes because of the increased percentages of boron that can be included. Test and fabrication experience indicate that a wide range of satisfactory rod designs is possible with any of these materials. The rod operating channels in the reactor often require liners to protect the surrounding graphite moderator from rod-insertion impact loads and wear and to help maintain channel alignment. Abrasion- and impact resistant, high-strength, low cross-section materials that will operate uncooled are required for these liners. Pyrolytic graphite, pyrolytic graphite composites, aluminium oxide and silicon carbide have been tested for such applications. Physical and irradiation damage data indicate that some of these materials are suitable for lining rod-operating channels. (author) [French] Les problemes de materiaux lies aux dispositifs de controle a barres de reglage et de securite pour les reacteurs tubulaires ralentis au graphite sont doubles et concernent les materiaux absorbants d'une part et les materiaux de garnissage des canaux d'autre part. Les materiaux absorbants tels que le bore ou le gadolinium peuvent former un tout avec le materiau de gainage comme dans le cas ou les barres de securite sont en acier inoxydable au bore. Une autre technique consiste a enfermer un melange presse et fritte contenant du bore, tel que B4C-graphite ou B4C-aluminium, dans une gaine metallique. Les barres de ce dernier type peuvent etre adaptees

  9. 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

  10. Fabrication of the 4. set of fuel elements for the experimental pile EL2; Fabrication du 4. jeu de barreaux de la pile d'essai EL2

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    The reactor EL2 is the second atomic reactor built in France. It is a laboratory reactor using heavy water and natural uranium. Its cooling circuit operates with compressed CO{sub 2} gas at 8 kg/cm{sup 2} pressure. The subject of this lecture is the manufacturing of the fourth set of rods. The principle of uranium-can connection is exposed: that is the principle of a pre-pressed bound can. The EL2 reactor has been a prototype with respect to this aspect of the question, and a prototype which has been quite satisfactory. The main steps of the fabrication are exposed: the {gamma} phase extension of uranium, the machining, the three canning (die canning, hydraulic canning, compressed air treatment), the automatic argon arc welding of cups and the different manufacturing controls. (author) [French] Le reacteur EL2 est le deuxieme reacteur construit en France. C'est un reacteur de recherches qui utilise de l'eau lourde et de l'uranium naturel. Il est refroidi par du gaz carbonique sous 8 kg/cm{sup 2} de pression. On etudie dans cet expose la fabrication du quatrieme jeu d'elements combustibles. Le principe de la liaison uranium-gaine est expose: c'est celui d'une gaine precontrainte. La pile EL2 a constitue un prototype a ce point de vue, prototype qui a donne entiere satisfaction. Les principales etapes de la fabrication sont ensuite expliquees: le filage {gamma} de l'uranium, l'usinage des barreaux, les trois operations de gainages (gainage par filiere, gainage hydraulique, gainage a chaud), la soudure automatique des bouchons a l'argon-arc et les differents controles de fabrication. (auteur)

  11. 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

  12. New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

  13. The 'Reacteur Jules Horowitz': a new experimental reactor project

    International Nuclear Information System (INIS)

    Frachet, S.; Ballagny, A.

    1999-01-01

    The Jules Horowitz Reactor (RJH) is a new research reactor project dedicated to materials and nuclear fuel testing, the location of which is foreseen at the CEA-CADARACHE site, and the start-up in 2006. The launching of this project originated from a double finding: The development of nuclear power plants aimed at satisfying the energy needs of the next century, cannot be envisaged without experimental reactors which are unrivaled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. The existing experimental reactors are 30 to 40 years old and it is advisable to examine henceforth the necessity for and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a study on the mid and long term irradiation needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfill the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The reactor project selected among several different concepts, is finally a light water pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows for many irradiations in and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and the French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the following most important items: neutronic and thermalhydraulic studies on alternative core designs, with or without added pressurization, assessment of different core surrounding structures in connection with the core studies, overall layout of the reactor/auxiliary pools and reactor building. (author)

  14. The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

  15. Present Status of Nitrogen Fixation by Reactor Radiation; Etat Actuel des Recherches sur l'oxydation directe de l'azote sous irradiation dans des reacteurs; Sovremennoe sostoyani opytov po okisleniyu azota izlucheniem iz reaktorov; Estado actual de las investigaciones sobre fijacion del nitrogeno por irradiacion en reactores

    Energy Technology Data Exchange (ETDEWEB)

    Harteck, P; Dondes, S [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1960-07-15

    'oxydation directe de l'azote sous irradiation, entreprises depuis plusieurs annees par le Rensselaer Polytechnic Institute et le Brookhaven National Laboratory, utilisent directement les particules de recul de fission comme rayonnements ionisants, au moyen de la dispersion d'uranium-235 dans des fibres de verre de cinq microns de diametre environ. Les auteurs ont determine les effets de la temperature, de la pression et du rapport azote/oxygene sur la valeur de G pour l'oxydation de l'azote et ont publie le compte rendu de leurs travaux. Ils en donnent un bref apercu. Les recherches en question ont ete effectuees avec des systemes statiques; plus recemment des systemes statiques et des systemes a circulation ont ete utilises a la fois. Avec les systemes statiques, les auteurs se sont surtout attaches a etudier l'effet de l'intensite des rayonnements, notamment sur la cinetique d'equilibre sous irradiation. Ils ont constate que dans des melanges ou le rapport azote/oxygene est de 4 a 1 et de 2 a 1 N0{sub 2} et N{sub 2}0 se forment jusqu'a epuisement de tout l'oxygene present. Un systeme a circulation continue (cycling) fonctionne maintenant dans une boucle a l'interieur du reacteur de Brookhaven. Les auteurs fournissent sur les effets de la temperature, de la pression, du rapport azote/oxygene et de l'intensite des rayonnements des donnees que l'on pourra utiliser pour etablir un projet de reacteur de chimie nucleaire. Le systeme actuel fonctionne sous 10 atmospheres et a 150{sup o}C. La temperature est fonction de l'energie de fission liberee dans les fibres de verre et de la resistance thermique du circuit. Une autre boucle, qui doit fonctionner sous 50 - 75 atmospheres et a 600{sup o} C, est en construction. Il est possible, grace a ces boucles, d'etudier les caracteristiques d'un systeme continu, y compris le comportement des produits de fission liberes dans le courant, gazeux. Les auteurs distinguent trois stades dans la cinetique complexe de l'oxydation de l'azote: reactions

  16. 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

  17. 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

  18. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J C [Electricite de France (EDF), 75 - Paris (France); Zaetta, A [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G [CEA/Saclay, DEN, 91 - Gif sur Yvette (France); and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  19. Handling and Separation of Short-Lived Radioisotopes from Research Reactors; Manipulation et Separation des Radioisotopes a Courte Periode Produits dans des Reacteurs de Recherche; ПОЛУЧЕНИЕ И ОТДЕЛЕНИЕ КОРОТКОЖИВУЩИХ ИЗОТОПОВ В ИССЛЕДОВАТЕЛЬСКИХ РЕАКТОРАХ; Manipulacion y Separacion de Radioisotopos de Periodo Corto Obtenidos en Reactores de Investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Meinke, W. W. [University of Michigan, Ann Arbor, MI (United States)

    1963-03-15

    distillation, selective reduction, etc., also add to the variety of separation possibilities to be explored. The local research reactor, whether it is in a university in the United States, or in a developing country, thus opens a whole new era of tracer possibilities. (author) [French] L'emploi des radioisotopes a souvent ete limite aux radioisotopes dont la periode est superieure a un jour, etant donne l'eloignement du reacteur qui les produit. Ceci explique un certain manque d'interet a l'egard du traitement et de l'utilisation de ces radioisotopes, et par suite une certaine reticence de la part du consommateur a envisager meme les possibilites d'emploi de nombreux radioisotopes a courte periode. Comme il existe maintenant de nombreux reacteurs de recherche dans le monde, les laboratoires ne dependent plus de producteurs de radioisotopes eloignes; en outre, les radioisotopes a courte periode couvrent de nombreux champs d'experimentation nouveaux. Il importe, cependant, a cette fin de considerer la production des radioindicateurs sous un angle nouveau. Depuis pres de cinq annees, le programme execute au moyen du reacteur de recherche de l'Universite du Michigan comporte la manipulation, le traitement et la mesure de radioisotopes a courte periode. Les chercheurs de l'Universite emploient couramment des radioisotopes dont les periodes ne depassent pas plusieurs heures, voire quelques minutes. Les traveaux entrepris jusqu'a present avaient trait principalement a l'analyse par activation, mais le material, les methodes et les techniques utilises.peuvent s'appliquer a de nombreux autres domaines. Pour utiliser les radioisotopes a courte periode, il n'est pas necessaire de prevoir un roulement de trois equipes pour le reacteur; il n'est pas lion plus indispensable de disposer de stocks importants de radioisotopes, ni d'installations de traitement perfectionnees.En fait, de simples pinces, utilisees de la maniere courante, donnent generalement de meilleurs resultats que de

  20. Dispersions of Oxides in Oxide Matrices as High-Temperature Reactor Fuels; Dispersions d'oxyde dans des matrices d'oxyde, utilisees comme combustibles dans des reacteurs a haute temperature; Dispersiya okisej v okislovykh matritsakh v kachestve topliva dlya vysokotemperaturnogo reaktora; Empleo de dispersiones de oxidos en matrices de oxidos, como combustibles para reactores de elevada temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Williams, J. [Atomic Energy Research Establishment, Harwell (United Kingdom)

    1963-11-15

    The potential usefulness of dispersions of PuO{sub 2}, UO{sub 2} and ThO{sub 2} in matrices of BeO, Al{sub 2}O{sub 3}, MgO and SiO{sub 2} is reviewed in terms of fuel integrity and fabrication. Dimensional stability and fission-product retentivity are the two features most important to fuel integrity. Compatibility of the constituents of the fuels with one another and with the coolant will influence dimensional stability, but oxide fuels are well favoured in these respects. Dimensional changes under irradiation will contain contributions from neutron and fission fragment damage to the matrix, from radiation damage to the fissile-fertile phase and from agglomerated fission-product gases. Thermal stresses are also capable of effecting changes in shape. However, information on mechanisms for stress relaxation is too limited to enable any reasonable theoretical assessment of behaviour to be made. Both light irradiation and high burn-up studies of fission-product release from the fissile-fertile oxides have concerned themselves mainly with the gaseous products, chiefly xenon. Data on the release of other fission products is very limited as is also information on the movement of fission products in general through the potential matrix materials. Studies of the permeability of sintered pure oxides indicate that densities of at least 95% theoretical density (maybe even 98%) will be needed to eliminate open porosity in such matrices. A variety of techniques are available for the preparation of fissile-fertile particles, for their coating and for their incorporation into high-density matrices. Work on laboratory-scale fabrication processes is well advanced. (author) [French] L'auteur examine la possibilite d'utiliser des combustibles disperses - PuO{sub 2}, UO{sub 2} et ThO{sub 2} et matrices de BeO, Al{sub 2}O{sub 3}, MgO et SiO{sub 2} - dans des reacteurs a haute temperature, au point de vue de l'integrite du combustible et de sa transformation. La stabilite dimensionnelle

  1. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l'alliage Mg-Zr se comporte nettement mieux que le magnesium pur et surtout que l'alliage Mg-Zr-Zn. (auteur)

  2. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l'alliage Mg-Zr se comporte nettement mieux que le magnesium pur et surtout que l'alliage Mg-Zr-Zn. (auteur)

  3. Metallurgy of platinoids. Studies of industrial cases; Metallurgie des platinoides. Etudes de cas industriels

    Energy Technology Data Exchange (ETDEWEB)

    Blazy, P. [Ecole Nationale Superieure de Geologie, 54 - Vandoeuvre les Nancy (France); Jdid, E. A. [Institut National Polytechnique de Lorraine (ENSG/INPL/CNRS UMR 7569), 54 Vandoeuvre les Nancy (France)

    2004-03-01

    The platinoids deposits, currently in exploitation, are often associated with dunitic and gabbroitic ultra-basic rocks containing copper, nickel and iron sulfides. The proportion of platinoids in these deposits is variable and requires generally an adaptation of the industrial processes for their recovery. Nevertheless, we can consider that the classical industrial way consists to treat the composite concentrates of copper-nickel-platinoids sulfides by melting/conversion operations following by hydrometallurgical operations of recovery of the basic metals Cu, Ni. These last ones concentrate the platinoids in leaching residues. In these residues, the platinoids are extracted and separated by hydrometallurgical ways including different techniques: dissolution, selective precipitation, distillation, solvent extraction, resins...The great variety of uses of platinoids makes their recycling difficult. The wastes are often too complex for being economically recycling. The environmental protection depends of the type of industry (metallurgical, chemical, mining). Some products present risks of fire and explosion. The toxicity often appears by superficial allergies and by serious troubles in case of ingestion of soluble salts. (O.M.)

  4. Computer aided analysis and design of industrial energy systems; Rechnergestuetzte Analyse und Konzeption industrieller Energiesysteme

    Energy Technology Data Exchange (ETDEWEB)

    Augenstein, Eckardt Marc Guenter

    2009-03-02

    In this dissertation the concept and implementation of a software system supporting the analysis and the design of industrial energy systems is presented. As a basis, a software framework was designed supplying a domain specific object model allowing the description of energy systems as well as the energy auditing projects performed with the software. Moreover, a set of graphical and textual editors needed to model the examined systems is part of the framework. On the other hand, the professional methods for analysis, assessment and optimization of energy systems are implemented in modules integrated into the system via a plug-in interface. The object model whose definition was based on a meta model approach allows the description of network like structures typical to energy systems. In order to keep track of the different work steps performed during an analysis project, these steps are reflected in the object model as ''method applications'' using a tree as the basic structure of a project. In order to allow the compatibility of information a set of conventions for the evaluation of energy flows and system balances was introduced. Moreover, all data elements used in modules or model components are derived from a central database guaranteeing a consistent usage of terms, descriptions, validity ranges and data types. The single professional modules like simulators or optimization methods access the object model via an appropriate software interface. Moreover, they make use of the framework's user interface engine by delivering a generic description of dialog screens and result reports. As all modules share the same set of objects modelling the components of the energy system surveyed, the flow of information from module to module can be designed virtually seamless. Compared to a number of stand-alone solutions, this integrated design approach has the advantage that by combining a set of specialized methods an overall solution for complex engineering tasks can be created. The modules implemented up to now form a solution for the computational support of industrial energy audits. Nevertheless, the framework could be used as a basis for other fields of engineering. As an example for a complex module, a simulator for the calculation of industrial energy supply systems is presented. This module allows modelling of supply systems with low effort in order to calculate the annual costs, system efficiency and emissions. Besides technical components for the conversion, storage and transport of energy, other decisive elements like energy tariffs can be modelled. As input for the simulation, time series of the different target energy demands are needed. As detailed design data of the components is usually not available, the model parameters are typically restricted to the data found in technical data sheets. Moreover, typical sample times of energy demand time series will be 15 minutes or higher, so that dynamic effects below this time interval are neglected. For the purpose of analysis it turns out to be advantageous to assess a supply system without the influences of the concrete system control strategy, which means to run the simulation under the regime of an optimal control strategy. Even in cases where the control strategy is to be taken into account, this approach allows a simpler modelling of the system control as aspects with little impact on the system efficiency can be left to the optimizer instead of formulating appropriate rules. In order to allow an operation optimization of machines whose efficiency depends on temperatures in addition to the part load state (e.g. chillers), an optimization method which allows for quadratic constraints was selected. In order to achieve a method most robust towards the large variety of systems structures to be handled, a combination of evolutionary algorithms and mixed integer linear programming was chosen. In order to create supply system models, component models can be chosen from a template library and combined by connecting their external interfaces with bonds representing an energy or medium flow of a certain type. Each model component may contain subsystems leading to a hierarchical model structure. To define the behaviour of the components, a model description language adapted to the mathematical model was developed. The application of the system is demonstrated in an example of use. (orig.)

  5. Evaluation of the industrial programs of NOVEM. Evaluatie industriele programma's NOVEM

    Energy Technology Data Exchange (ETDEWEB)

    Koster, S J; Rademaker, B; Braun, A R; Gerritse, G

    1989-01-01

    The title subject has been evaluated by INDIS (Dutch Bureau for Industrial and Professional Investigations) in order to analyse the demand of the target group and to collect data for the product/market strategical part of the long-range planning for 1989-1993 and 1994-1998. INDIS collected data from 20 interviews with policy-making managers from industrial companies, and 9 interviews with consultancy bureaus, engineering bureaus, contractors, business organizations and the government. These data were supplemented by a literature search. The effectiveness of the NOVEM (Netherlands Agency for Energy and the Environment) has been investigated with regard to the possibilities of developed technologies. This investigation was done by Braun Technoconsult. Conclusions and recommendations are given with regard to energy and environmental projects, the roles of the government, the umbrella organizations and the NOVEM and finally the effectivity of the applicated tools (information, counseling, demonstrations, financing, etc.).

  6. Biomonitoring of toxic compounds of airborne particulate matter in urban and industriel areas

    DEFF Research Database (Denmark)

    Klumpp, Andreas; Ro-Poulsen, Helge

    2010-01-01

    The toxicity and ecotoxicity of airborne particulate matter is determined by its physical features, but also by its chemical composition. The standardised exposure of accumulative bioindicator plants is suggested as an efficient and reliable tool to assess and monitor effects of particulate matter...

  7. 127 Impact des rejets urbains et industriels sur la qualité des eaux ...

    African Journals Online (AJOL)

    PR BOKO

    scour of surfaces and underground. Keywords :aquifer, hydrochimy, rejections, pollution, El Kantara, Biskra, Algeria. 1. Introduction. La qualité des ..... A case study: Suquıa River Basin Au vue des résultats d'analyses hydrochimique, les concentrations en sulfates dépassent la norme (200-400mg/L). (Cordoba-. Argentina).

  8. The government favours the negotiation with industrialists; Le gouvernement privilegie la negociation avec les industriels

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2002-02-01

    The French government has finally opted for agreements negotiated with industrialists to reduce their greenhouse gases emissions, opening thus the way of future emission credits markets. The 'TGAP' (General Taxes on the Pollutant Activities) will thus not be implemented. (O.M.)

  9. The industrial development of atomic energy; Le developpement industriel de l'energie atomique

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    Countries with large stock of fissile material and producing large quantity of nuclear pure {sup 235}U and {sup 239}Pu are able to allocate part of the stock to non military research. For countries with low stock of fissile material, all the stock is allocated to military research. An economical and technical solution has to be find to dedicate a part of fissile material to non military research and develop the atomic energy industry. It stated the industrial and economical problems and in particular the choice between the use of enriched fuel with high refining cost or depleted fuel with low production cost. It discusses of four possible utilizations of the natural resources: reactors functioning with pure fissile material ({sup 235}U or {sup 239}Pu) or concentrated material ({sup 235}U mixed with small quantities of {sup 238}U after an incomplete isotopic separation), breeder reactors functioning with enriched material mixed with {sup 238}U or Thorium placed in an appropriate spatial distribution to allow neutrons beam to activate {sup 238}U or Thorium with the regeneration of fissile material in {sup 239}Pu, reactors using natural uranium or low enriched uranium can also produce Plutonium with less efficiency than breeder reactors and the last solution being the use of natural uranium with the only scope of energy production and no production of secondary fissile material. The first class using pure fissile material has a low energy efficiency and is used only by large fissile material stock countries to accumulate energy in small size fuel for nuclear engines researches for submarines and warships. The advantage of the second class of reactors, breeder reactors, is that they produce energy and plutonium. Two type of breeder reactor are considered: breeder reactor using pure fissile material and {sup 238}U or breeder reactor using the promising mixture of pure fissile material and Thorium. Different projects are in phase of development in United States, England and Scotland. The third class of reactor using natural uranium as fuel are presented as a possibility for double-function reactor with the production of plutonium and energy, but the neutron balance is lower than with breeder reactor. One solution is to increase the temperature of functioning but it induces to change the structure materials and moderators. Different solutions are discussed about the utilization of graphite or heavy water as moderators. The last class of reactors using natural uranium and producing only energy is considered by countries with no uranium stock, the energy efficiency and balance, as well as the costs, are then of more importance. Finally, it presented conclusions about the different economic strategies about the industrial development of atomic energy in countries with and without fissile material resources. (M.P.)

  10. High temperature heat pumps for industrial cooling; Hoejtemperatur varmepumper til industriel koeling

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, Lars; Nielsen, Jacob [Advansor A/S, Aarhus (Denmark); Kronborg, H. [Cronborg, Holstebro (Denmark); Skouenborg, K. [Jensens Koekken, Struer (Denmark)

    2013-03-15

    This report deals with theoretical analysis of various types of integration of heat pumps in the industry, as well as a demonstration plant that serves the project's practical execution. The report describes the system integration between heat pumps and existing industrial cooling systems. Ammonia plants in industry are estimated to have an allocation of 85%, which is why only an analysis of this type of installation as surplus heat supplier is included in this report. In contrast, heat pumps with both CO{sub 2} and Isobutane as the refrigerant are analysed, since these are the interesting coolants for generating high temperature heat. It can be seen through the project that the combination of heat pump with existing cooling installations may produce favorable situations where the efficiency of the heat pump is extremely high while at the same time electricity and water consumption for the cooling system is reduced. The analysis reflects that CO{sub 2} is preferred over Isobutane in the cases where a high level of temperature boost is desired, whereas Isobutane is preferable at low level of temperature boost. In the demonstration project, the report shows that the heat pump alone has a COP of 4.1, while the achieved COP is 5.5 when by considering the system as a whole. In addition to increased performance the solution profits by having a reduction in CO{sub 2} emissions of 81 tons/year and a saving of 470,000 DKK/year. (LN)

  11. The management of industrial wastes in hydrology; La gestion des dechets industriels en hydrologie

    Energy Technology Data Exchange (ETDEWEB)

    Elbaz-Seboun, V.

    1998-07-08

    The industrial wastes are made of different kind of wastes: the inert wastes, the banal wastes (municipal wastes), the special wastes containing noxious elements with respect to human health and environment, and the radioactive wastes. Each industry generates its own effluents (sludges from water treatment plants and leachates from rubbish dumps). The main water pollutions are due to the fermentescible organic matters, nitrates and heavy metals from the industrial waste waters. The aim of the public water agencies is to better protect the environment and to give help to the industrialists in the management of their wastes: reduction at the source, selective collection, valorization, transportation and processing. Non-valorizable wastes must be processed: physico-chemical and biological processing (bio-filtering, coagulation-flocculation, membranes and industrial gases), incineration (organic wastes), disposal in class 1 technical burial centres after stabilization (ultimate wastes). Since July 2002, only the ultimate wastes will be disposed off and all class 2 and 3 dumps must have been rehabilitated. This work is divided into 2 parts: part 1 gives a presentation of the different types of industrial wastes and of their management (origin of wastes, effluents, heavy metals, environmental impact, legal aspects, wastes management, valorization). The second part describes the different processes for the treatment of industrial wastes (conventional processes, physico-chemical and biological processes, incineration, tipping, processing of radioactive wastes). (J.S.)

  12. Days on safety of industrial radiographic controls; Securite des controles radiographiques industriels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    This program is divided in three parts: the context and the regulations, the preparation and the implementation, the tools of prevention and the initiatives and the perspectives.In the first part devoted to the context and regulation are: the context by the Authority of nuclear safety (A.S.N.), the regulation referential, the transport of gamma-graphs; in the second part are the distribution of liabilities, materials and associated requirements, the feedback of incidents and exploitation of it, training and base requirements, works of S.F.R.P./C.O.F.R.E.N.D. and the A.S.N. position; the third part includes help to evaluation of risks at working places of industry radiologists, dosimetry study of a working place, guide to evaluate oneself; the fourth part devoted to the initiatives and the perspectives are: regional experiences charters of good practices in industry radiography, integration of works and deployment by the members of the C.O.F.R.E.N.D., perspectives in matter of prevention of occupational risks in the area of industry radiography. (N.C.)

  13. Psychological reactions to catastrophes: fear as a reaction to accidents and emergencies in industriel complexes

    International Nuclear Information System (INIS)

    Wernli, G.

    1989-01-01

    This paper addresses the subject 'stress' and its effects on the personality. Specific types of behavior, which the human develops in a fearful situation are demonstrated by means of the psychoanalytical personality model. In the conclusion possible methods of alleviating fear, shock and panic reactions are described. 7 figs., 2 tabs., 9 refs

  14. Programproduktion som Kritisk Teori – eller tv-teksten som industriel iscenesættelse

    Directory of Open Access Journals (Sweden)

    John Caldwell

    2003-09-01

    Full Text Available Denne artikel peger på nødvendigheden af en revurdering af den tre-delte model, som Fiske og Gripsrud har fremført, ved at vise, hvorledes ‘sekundære’ og ‘tertiære’ tv-tekster uafladeligt bevæger sig eller rejser hen imod en ‘primær’ tekstuel status i det amerikan- ske multikanals-»flow«. En detaljeret gennemgang af industriens tekstuelle praksis – programbegivenheder, network-branding, ka- nallogoer, filmene bag filmen, pressemateriale på video, salgsma- teriale og tilgrænsende digitale medier – viser hvorledes industrien teoretiserer sine egne vilkår direkte på skærmen, og dermed også hvordan publikum vejledes igennem en sådan offentlig cirkulation af »inside«-viden om tv-systemet. Artiklen er oversat af Henrik Bødker.

  15. Eco-imposition du capital, emploi et développement industriel durable

    Directory of Open Access Journals (Sweden)

    Bernard Dupont

    2010-09-01

    Full Text Available En remplaçant la part des cotisations sociales à la charge de l’employeur par un éco-impôt sur le capital polluant, l’Etat peut inciter les entreprises à investir dans les trajectoires technologiques de la durabilité. Cet article expose un modèle macroéconomique à trois facteurs qui montre comment la fixation conjointe et combinée du taux de cotisations sociales et du taux d'éco-impôt peut assurer la stabilité globale des prix, garantir la neutralité budgétaire et conserver en l’état la compétitivité prix. Contrôlée par le ministère du développement durable, la déformation de la fonction de production provoquée par une fiscalité directe favorable au capital non polluant et au travail augmente le volume de l’emploi et améliore la qualité environnementale sans détériorer la compétitivité ni déséquilibrer les comptes sociaux.Replacing the welfare cost by an eco-taxation based on polluting capital can incite companies to invest in the technological trajectories for sustainability. This paper proposes a macroeconomic three factors model in which the jointed and combined determination of welfare cost and eco-taxation rates can maintain the global price stability, guarantees the budget balance of fiscal transfers and keeps the state of the competitiveness price as it was before. Controlled by the ministry of sustainable development, the deformation of the function of production, caused by a direct tax system favorable to the non-polluting capital and to the labour, increases the employment and improves the environmental quality without negative effects on competitiveness and social accounts.

  16. Caractérisation de deux effluents industriels au Togo :étude d'impact ...

    African Journals Online (AJOL)

    Characterization of two industrial effluents in Togo : environment impact study. Environment pollution due to two industrial effluents has been investigated. Results how that effluent derive from the factory of the treatment of Kpémé posphate ore was loaded with settling suspended matter (> 90 % of total suspended solids).

  17. Control and prevention of industrial air pollution: Special issue; Bestrijding en preventie van industriele luchtverontreiniging: Thema

    Energy Technology Data Exchange (ETDEWEB)

    Waque, W.P.G.M. [Bureau Vergunningen en Bedrijven, DCMR Milieudienst Rijnmond, Schiedam (Netherlands); Zijlstra, W.M. [Bureau Milieu en Ruimtelijke Ordening, VNO/NCW, The Hague (Netherlands); Buisman, C.J.N.; Dijkman, H. [Paques, Balk (Netherlands); Prins, W.L.; Verbraak, P. [Biostar Development, Balk (Netherlands); Den Hartog, A.J. [Hoogovens Corporate Research Laboratorium, IJmuiden (Netherlands); Jol, A. [Sector Milieutechnologie, DHV Milieu en Infrastructuur, Amersfoort (Netherlands); Van Ham, J. [ed.

    1994-12-01

    In four articles in this special issue of the magazine attention is paid to new techniques by which emissions to the air from the industry can be controlled and/or prevented. In the first article an overview is given of sources of air pollution, caused by dust. In the second article intermediate results of the KWS 2000 program (aimed at 50% reduction of the emission of volatile organic matter for the year 2000) are outlined. In the third article a cooperative biological (flue) gas desulfurization pilot plant project is discussed. In the fourth and last article the most important possible techniques to reduce the emission of volatile organic matter are highlighted

  18. Untersuchung bergmaennischer und industrieller Rueckstandshalden auf eine moegliche Freisetzung natuerlicher radioaktiver Elemente

    International Nuclear Information System (INIS)

    Schmitz, J.

    1985-01-01

    More than 350 deposits from mining and processing in Baden-Wuerttemberg and Lower Saxony were visited, measured radiometrically, assessed by a unified scheme and partly sampled. The highest local dose rates, up to 100 mSv/a, were found in the waste from uranium exploration, as expected. The values of these planned wastes were followed by waste from silver-cobalt mining dumped above ground, with dose rates up to 20 mSv/a. Fly ash and slag deposits had a dose rate of about 2 mSv/a, and are comparable with many porphyry and granite rocks. The higher values of slag deposits from iron ore smelting were clearly due to high thorium content. The lowest doses were from stone waste from the mining of hydrothermal lead-zinc and iron ores. The analyses of uranium, radium 226 and lead 210 confirmed the radiometric measurements. The analysis of waste seepage water and tunnel outlets gave only a low number of values which exceeded the drinking water concentration according to the Radiation Protection Order. (orig./RB) [de

  19. Numerical and experimental investigation of industrial electrostatic precipitators; Etude numerique et experimentale d`electrofiltres industriels

    Energy Technology Data Exchange (ETDEWEB)

    Tochon, P.

    1997-10-17

    This work deals with electrostatic precipitators or ESP used for gas-solid particles separation. By means of a dust-controlled testing loop created and realised at the GRETh`s plate-form (Research Group on Heat Exchangers) and a numerical model developed during this work from TRIO software, the study of the performances of different ESP geometries has been carried out. Many electrical, hydraulic and particular parameters governing solid particles collection under ionised electric field have been identified, measured and modelled. The numerical model, ratified with experimental data obtained during this study and from literature, allows to describe local and global phenomena occurring in any geometries. Furthermore, parametric studies have been carried out in order to propose some optimised geometries. allowing to increase collection efficiencies. At least, on-site measurements with CETIAT (Centre Technique des Industries Aerauliques et Thermiques) allow to identify dust particles likely to be thrown out to the atmosphere, and troubles peculiar to large scales industrial plants. The numerical model has also been tested on these data. At the end of this study, an efficient dust-controlled experimental tool, PACIFIC loop, and a numerical simulation allowing ESP sizing are available. (author)

  20. High power quality for customers; De l'electricite de qualite chez les industriels

    Energy Technology Data Exchange (ETDEWEB)

    Calani, B. [Electricite de France (EDF), 93 - Saint-Denis (France); Tanneau, G. [Electricite de France (EDF), 92 - Clamart (France)

    2000-01-01

    The electrical power quality is a strong criterion for many manufacturers. The improvement of the electrical network and devices has allowed to reach a good power quality level all over the French territory which satisfies the majority of most customer's expectations. However, in some cases, only local desensitization can ne efficient enough to meet the customer's requirement. Emergent equipment based on power electronic devices are evaluated at the Research and Development Division. The aim is to propose the best service offer to our customers on both technical and economical aspects. (authors)

  1. Use of glasses as industrial dosimeters; Utilisation des verres comme dosimetres industriels

    Energy Technology Data Exchange (ETDEWEB)

    Balestic, F. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Le Clerc, P.; Bonnaud, M. [Centre de Recherches des Glaceries de Saint-Gobain (France)

    1959-07-01

    Glasses have the property of colouring under the action of ionizing radiations. We endeavoured to specify the conditions under which the intensity of coloration can be used as a measure of the quantity of radiation to which the glass has been submitted. In the case of a glass loaded with cobalt, a study of the optical density at different wavelengths enabled us to find the factors governing the formation of coloured centres and their conservation in the glass. We give a set of calibrating curves for different values of these parameters (irradiation rate, irradiation temperature; fading time and fading temperature), enabling determination of radiation doses in the range from 10 000 to 1 000 000 rep from measured optical density. (author) [French] Les verres ont la propriete de se colorer sous l'action des rayonnements ionisants. Nous avons cherche a preciser les conditions dans lesquelles l'intensite de la coloration peut servir de mesure de la quantite du rayonnement auquel le verre a ete soumis. Dans le cas d'un verre charge au cobalt, l'etude de la densite optique a differentes longueurs d'onde a mis en evidence divers facteurs dont depend la formation des centres colores et leur conservation dans le verre. En prenant comme parametres ces divers facteurs (temperature d'irradiation, intensite d'irradiation, temperature de conservation et duree de conservation) nous avons etabli des courus d'etalonnage permettant la determination de doses entre 10 000 et 1 000 000 rep d'apres la densite optique observee. (auteur)

  2. Use of glasses as industrial dosimeters; Utilisation des verres comme dosimetres industriels

    Energy Technology Data Exchange (ETDEWEB)

    Balestic, F [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Le Clerc, P; Bonnaud, M [Centre de Recherches des Glaceries de Saint-Gobain (France)

    1959-07-01

    Glasses have the property of colouring under the action of ionizing radiations. We endeavoured to specify the conditions under which the intensity of coloration can be used as a measure of the quantity of radiation to which the glass has been submitted. In the case of a glass loaded with cobalt, a study of the optical density at different wavelengths enabled us to find the factors governing the formation of coloured centres and their conservation in the glass. We give a set of calibrating curves for different values of these parameters (irradiation rate, irradiation temperature; fading time and fading temperature), enabling determination of radiation doses in the range from 10 000 to 1 000 000 rep from measured optical density. (author) [French] Les verres ont la propriete de se colorer sous l'action des rayonnements ionisants. Nous avons cherche a preciser les conditions dans lesquelles l'intensite de la coloration peut servir de mesure de la quantite du rayonnement auquel le verre a ete soumis. Dans le cas d'un verre charge au cobalt, l'etude de la densite optique a differentes longueurs d'onde a mis en evidence divers facteurs dont depend la formation des centres colores et leur conservation dans le verre. En prenant comme parametres ces divers facteurs (temperature d'irradiation, intensite d'irradiation, temperature de conservation et duree de conservation) nous avons etabli des courus d'etalonnage permettant la determination de doses entre 10 000 et 1 000 000 rep d'apres la densite optique observee. (auteur)

  3. Suitability of Cadmium Tantalate and Indium Tantalate as Control Materials for High-Temperature Reactors; Le Tantalate de Cadmium et le Tantalate d'Indium Comme Absorbants pour les Reacteurs a Haute Temperature; Vozmozhnosti ispol'zovaniya tantalatov kadmiya i indiya v kachestve kontrol'nogo materiala dlya vysokotemperaturnykh reaktorov; Empleo del Tantalato de Cadmio y del Tantalato de Indio Como Materiales de Control Para Reactores de Alta Temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Preisler, E.; Haessner, F.; Petzow, G. [Max-Planck-Institut fuer Metallforschung, Stuttgart, Federal Republic of Germany (Germany)

    1964-06-15

    oxygen in their most stable valency states than the parent elements can. Therefore, the reduction of Cd{sup ++}to metal can be expected while indium tantalate should.be stable. This has been confirmed by experiments with SnO- and WO{sub 2}-additions to cadmium tantalate. Addition of copper oxide to the compounds suppresses this effect. (author) [French] Quelles que soient les conditions particulieres requises dans chaque cas d'espece, les absorbants utilises dans la pratique pour des reacteurs a haute temperature devraient avoir les proprietes suivantes: a) section efficace d'absorption elevee pour les neutrons d'une gamme etendue d'energies; b) forte capacite d 'absorption des neutrons; c) faible sensibilite au point de vue des dommages radioinduits; d) bonne resistance thermique; e) reactivite faible avec le milieu environnant; f) cout eleve et approvisionnemeent facile. Si l 'on tient compte de ces considerations et que l 'on veuille eviter les inconvenients des reactions (n, {alpha}), on s'interessera surtout aux elements suivants: cadmium, tungstene, indium et tantale. Il faut combiner un absorbant de neutrons thermiques efficace avec un absorbant de neutrons epi thermiques; le materiau ainsi obtenu est stable a des temperatures elevees ( Greater-Than-Or-Equivalent-To 700 Degree-Sign C). Les oxydes doubles CdWO{sub 4}, Cd {sub 2}Ta{sub 2}O{sub 7} et CdIn{sub 2}O{sub 2} conviennent bien a cette fin. En outre, c'est le tantalate de cadmium qui a la plus forte resistance thermique. Le tantalate d'indium est un autre oxyde double qui, en combinaison avec le tantalate de cadmium, possede un spectre d'absorption des neutrons interessant. Il a egalement une bonne resistance thermique. Etant donne qu'il faut souvent faconner les materiaux ceramiques absorbants par deformation plastique, on les utilise habituellement sous forme de cermets. C'est la raison pour laquelle ils doivent etre compatibles avec des metaux. Le tantalate de cadmium est compatible avec l'argent et le

  4. 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

  5. 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

  6. Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

  7. The Non-Destructive Testing of Fuel Elements and Their Components for the United Kingdom Power-Reactor Development Programme; Controle Non Destructif des Elements Combustibles et de Leurs Parties Constitutives dans le Cadre du Programme de Developpement des Reacteurs de Puissance au Royaume-Uni; Nedestruktivnoe ispytanie teplovydelyayushchikh ehlementov i ikh komponentov dlya osushchestvleniya programmy soedinennogo korolevstva po razrabotke ehnergeticheskikh reaktorov; Ensayo No Destructivo de Elementos Combustibles y sus Componentes, en el Marco del Programa de Reactores de Potencia del Reino Unido

    Energy Technology Data Exchange (ETDEWEB)

    Mann, C. A.; Campsie, I. C. [U.K.A.E.A., Reactor Fuel Element Laboratories, Springfields, Salwick, Preston, Lancs. (United Kingdom)

    1965-10-15

    and the ends closed. In addition, the integrity of end closures is established, by radiography. Multiple exposures are commonly made to examine the whole of circumferential weld adequately. The disposition of the fuel can also be recorded accurately by using a panoramic technique. The use of colour radiography is also discussed. Pins are normally tested for leakage after filling with helium, using a mass-spectrometer leak detector. Pins not filled with helium may be tested using a ''back-pressurizing'' technique. Conventional ''probing'' and ''sniffing'' methods are used when it is desirable to locate the sites of leaks. The bubble test in liquids is also used, as a cheap and simple test. The use of krypton-85 as a tracer gas is discussed. (author) [French] Les auteurs decrivent les methodes d'essai que les laboratoires charges des elements combustibles ont elaborees dans le cadre du programme etabli par le reacteurs> en vue de mettre au point des aiguilles de combustible pour diverses filieres de reacteurs. Ces aiguilles sont contenues dans des gaines de 5 a 15 mm de diametre, les materiaux utilises etant des aciers inoxydables et des alliages de zirconium, a) Detection de defauts dans les gaines. Examen par ultrasons a l'aide de deux traducteurs immerges. Les tubes sont animes d'un mouvement helicoidal rapide dans un reservoir fixe. Chaque signal de defaut est verifie et enregistre. Pour regler le dispositif et verifier sa stabilite, on utilise comme temoins des fentes'pratiquees a l'arc a la surface des tubes. Dans certains cas, on a egalement recours au controle par courants de Foucault. Les auteurs decrivent deux procedes: l'un, a debit rapide, est fonde sur un systeme de bobines encerclant le tube; l'autre, a exploration heliccfldale, utilise une bobine se deplacant le long du tube. Les signaux fournis par un circuit a pont sont selectionnes selon la phase et filtres, pour des frequences de 30 a 60 kHz. b) Controle des dimensions de tubes et de

  8. Improved Techniques for Low-Flux Measurement of Prompt Neutron Lifetime, Conversion Ratio and Fast Spectra; Methodes Perfectionnees de Mesure de la Duree de Vie des Neutrons Instantanes, du Rapport de Conversion et des Spectres de Neutrons Rapides, dans un Reacteur a Bas Flux; Usovershenstvovannye metody izmereniya vremeni zhizni mgnovennykh nejtronov, koehffitsienta konversii i spektra bystrykh nejtronov pri slabykh potokakh nejtronov; Tecnicas Perfeccionadas para la Determinacion del Periodo de los Neutrones Inmediatos, la Razon de Conversion y los Espectros de Neutrones Rapidos, con Flujos Reducidos

    Energy Technology Data Exchange (ETDEWEB)

    Armani, R. J.; Bennett, E. F.; Brenner, M. W.; Bretscher, M. M.; Cohn, C. E.; Huber, R. J.; Kaufmann, S. G.; Redman, W. C. [Argonne National Laboratory, Argonne, IL (United States)

    1964-02-15

    been concentrated on the use of pulse shape analysis to reject gamma-ray initiated events in hydrogen recoil proportional counters and the introduction of collimation in Li{sup 6}F solid-state detector ''sandwiches'' to improve the resolution obtained. A number of such instruments have been built and their response to mono-kinetic and reactor neutrons has been investigated. Use of the gamma-ray rejection technique was equivalent to a several hundred-fold effective reduction in gamma-ray sensitivity of the recoil counter and extends the usable range down to at least 30 keV. For the Li{sup 6} solid-state devices, resolutions as low as 70 keV full-width at half maximum (1.5%) have been observed for the sum pulse in thermal neutron irradiation. (author) [French] Dans le programme des reacteurs de puissance zero, on a utilise diverses methodes statistiques pour mesurer le rapport duree de vie des neutrons instantanes/duree de vie des neutrons differes. Les auteurs ont mis au point une methode nouvelle, qui consiste a analyser le bruit du reacteur a l'aide d'un filtre passe-bande, et ont perfectionne d'autres methodes telles que la mesure, a l'aide d'un compteur a impulsions, de la frequence des coincidences retardees en fonction du temps de retard et celle de la variance relative des flux de neutrons integres en fonction du temps d'integration. Ils ont etudie les domaines dans lesquels les differentes methodes peuvent etre utilisees avec le plus d'interet. II se sont aussi preoccupes de l'interpretation des resultats de ces mesures, et montrent que l'interpretation fondee sur un modele cinetique simple peut s'appliquer dans la pratique a une grande diversite de cas. Les auteurs decrivent plusieurs perfectionnements de leur methode d'activation pour la determination du rapport de conversion: application de techniques chimiques tres sensibles pour confirmer les resultats obtenus; correction pour les coups parasites en utilisant, dans la determination de la capture, des

  9. Industrial lightning. Towards a global cost approach of an industrial lightning installation; Eclairage industriel. Pour une approche en cout global d'une installation d'eclairage industriel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    This document proposes an approach of the global cost of an industrial lightning installation and provides information on the benefits and its associated performance of a good industrial lightning, examples of applications, the available products and the regulation texts of the domain. (A.L.B.)

  10. 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 a celle utilisee en simulation. Les parametres experimentaux ayant ete choisis en accord avec l'optimisation effectuee lors des simulations et apres une premiere optimisation du procede experimental. Finalement les resultats experimentaux confirment les resultats obtenus en simulation, et demontrent le potentiel de la methode developpee.

  11. Complete automation of nuclear reactors control; Automatisation complete de la conduite des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

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

    1955-07-01

    The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P{sub max} and R{sub max} (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P{sub max} is reached. (M.P.)

  12. Liquid distribution in trickle-bed reactor; Distribution du liquide en reacteur a lit ruisselant

    Energy Technology Data Exchange (ETDEWEB)

    Marcandelli, C.; Wild, G. [Centre National de la Recherche Scientifique (CNRS-ENSIC), Lab. des Sciences du Genie Chimique, 54 - Nancy (France); Lamine, A.S. [CNRS-Universite de Paris-Nord, Lab. d' Ingenierie des Materiaux et des Hautes Pressions, 93 - Villetaneuse (France); Bernard, J.R. [Elf Antar France, Centre de Recherche Elf de Solaize, 69 - Solaize (France)

    2000-07-01

    The aim of this study is to develop techniques to qualify the efficiency of liquid distribution in trickle-bed reactors, using cold mockups. The experimental setup consists mainly in a 0.3-m-ID packed-bed column with three different plates used to vary the quality of inlet liquid distribution. Liquid distribution has been qualified using several techniques: global pressure drop measurements, global RTD (Residence-Time Distribution) of the liquid, local heat transfer probes, capacitance tomography, collector at the bottom of the reactor with nine equal zones. The bed pressure drop and the overall external liquid saturation decrease when the maldistribution increases; quantitative information is however difficult to obtain this way. Global RTD of the liquid allows quantifying of the average liquid distribution in the bed. The local thermal sensors give an indication of local liquid velocity and indicate possible local maldistribution of the liquid (scale mm) even when global distribution is good. Concerning the results obtained with the collector, a maldistribution index is defined ranging from 0 (ideal distribution) to 1 (worst possible distribution), and the influence of the different operating parameters (gas and liquid velocities, particle shape) is discussed. (authors)

  13. The hydraulics of the pressurized water reactors; L'hydraulique des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Bouchter, J.C. [CEA Cadarache, SMET, 13 - Saint-Paul-lez-Durance (France); Barbier, D. [CEA/Grenoble, Dept. de Thermohydraulique et de Physique, DTP/SH2C, 38 (France); Caruso, A. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

    1999-07-02

    The SFEN organized, the 10 june 1999 at Paris, a meeting in the domain of the PWR hydraulics and in particular the hydraulic phenomena concerning the vessel and the vapor generators. The papers presented showed the importance of the industrial stakes with their associated phenomena: cores performance and safety with the more homogenous cooling system, the rods and the control rods wear, the temperature control, the fluid-structure interactions. A great part was also devoted to the progresses in the domain of the numerical simulation and the models and algorithms qualification. (A.L.B.)

  14. Fast neutron dosimetry in research reactors; Dosimetrie en neutrons rapides dans les reacteurs de recherche

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [French] On etudie principalement la mesure des flux de neutrons rapides a l'aide de detecteurs a seuil. On montre d'abord que les sections efficaces a utiliser pour les mesures par detecteurs a seuil, dependent grandement du spectre des neutrons, c'est-a-dire de l'emplacement ou s'effectue la mesure. La determination du spectre est effectuee par le calcul pour plusieurs emplacements des piles EL2 et EL3; on en deduit les sections efficaces a utiliser pour les mesures effectuees a ces emplacements. Dans la derniere partie du rapport, on indique quelles methodes sont possibles pour la determination experimentale du spectre. (auteur)

  15. Validation d'un nouveau calcul de reference en evolution pour les reacteurs thermiques

    Science.gov (United States)

    Canbakan, Axel

    Resonance self-shielding calculations are an essential component of a deterministic lattice code calculation. Even if their aim is to correct the cross sections deviation, they introduce a non negligible error in evaluated parameters such as the flux. Until now, French studies for light water reactors are based on effective reaction rates obtained using an equivalence in dilution technique. With the increase of computing capacities, this method starts to show its limits in precision and can be replaced by a subgroup method. Originally used for fast neutron reactor calculations, the subgroup method has many advantages such as using an exact slowing down equation. The aim of this thesis is to suggest a validation as precise as possible without burnup, and then with an isotopic depletion study for the subgroup method. In the end, users interested in implementing a subgroup method in their scheme for Pressurized Water Reactors can rely on this thesis to justify their modelization choices. Moreover, other parameters are validated to suggest a new reference scheme for fast execution and precise results. These new techniques are implemented in the French lattice scheme SHEM-MOC, composed of a Method Of Characteristics flux calculation and a SHEM-like 281-energy group mesh. First, the libraries processed by the CEA are compared. Then, this thesis suggests the most suitable energetic discretization for a subgroup method. Finally, other techniques such as the representation of the anisotropy of the scattering sources and the spatial representation of the source in the MOC calculation are studied. A DRAGON5 scheme is also validated as it shows interesting elements: the DRAGON5 subgroup method is run with a 295-eenergy group mesh (compared to 361 groups for APOLLO2). There are two reasons to use this code. The first involves offering a new reference lattice scheme for Pressurized Water Reactors to DRAGON5 users. The second is to study parameters that are not available in APOLLO2 such as self-shielding in a temperature gradient and using a flux calculation based on MOC in the self-shielding part of the simulation. This thesis concludes that: (1) The subgroup method is at least more precise than a technique based on effective reaction rates, only if we use a 361-energy group mesh; (2) MOC with a linear source in a geometrical region gives better results than a MOC with a constant model. A moderator discretization is compulsory; (3) A P3 choc law is satisfactory, ensuring a coherence with 2D full core calculations; (4) SHEM295 is viable with a Subgroup Projection Method for DRAGON5.

  16. Hybrid reactors: recent progress of a demonstration pilot; Reacteurs hybrides: avancees recentes pour un demonstrateur

    Energy Technology Data Exchange (ETDEWEB)

    Billebaud, Annick [Laboratoire de Physique Subatomique et de Cosmologie IN2P3-CNRS/UJF/INPG, 53 av. des Martyrs, 38026 Grenoble Cedex (France)

    2006-12-15

    Accelerator driven sub-critical reactors are subject of many research programmes since more than ten years, with the aim of testing the feasibility of the concept as well as their efficiency as a transmutation tool. Several key points like the accelerator, the spallation target, or neutronics in a subcritical medium were investigated extensively these last years, allowing for technological choices and the design of a low power European demonstration ADS (a few tens of MWth). Programmes dedicated to subcritical reactor piloting proposed a monitoring procedure to be validated in forthcoming experiments. Accelerator R and D provided the design of a LINAC for an ADS and research work on accelerator reliability is going on. A spallation target was operated at PSI and the design of a windowless target is in progress. All this research work converges to the design of a European demonstration ADS, the ETD/XT-ADS, which could be the Belgian MYRRHA project. (author)

  17. Calculus of a reactor VVER-1000 benchmark; Calcul d'un benchmark de reacteur VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Dourougie, C

    1998-07-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  18. Concurrence verticale industriels-distributeurs : stratégie des acteurs et évolution du droit

    OpenAIRE

    Chanut , Odile

    2015-01-01

    International audience; Price war on the shelves for national and international brands, pursuit of food retail groups concentration or supermarket procurement centres concentration, cartels in the industrial sector to cope with retailers purchasing power, French Competition Authority’s recent reports, ALUR Law (2014) and Macron Law (2015) including some policy on retail ... news shows that vertical competition between the food industry and the large retailers is exacerbated in a context of we...

  19. Volumes of common industrial wastes: a study report; Dechets industriels banals: quel tonnage? rapport d`etude

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The total common industrial waste volume production in France has been evaluated, taking into consideration all the industrial and commercial sectors and the following materials: glass, metals, plastics, rubber, textiles, papers, cardboard, wood, leather, organic matters, building wastes, mixtures. Results are presented for the various regions of France, as a function of enterprise size, waste type and destination; data are also given concerning packaging materials, and waste collection and processing. Comparisons are made with data from other information sources and calculations

  20. Utilization of efficiency potentials with coldness from industrial wast heat; Mit Kaelte aus industrieller Abwaerme Effizienz-Potenziale nutzen

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2013-04-15

    Increasing energy efficiency is one of the main challenges in the plastics processing industry. However, the structure and the parameters of the cooling sections of extrusion lines have been optimized only in a few cases with respect to maximum efficiency and minimal operating costs. Thus, an innovative cascaded cooling section enables not only an energy-efficient cooling, but also an improvement of the properties of the finished plastic products.

  1. Variation and management of the tertiary and industrial lightning; Variation et gestion de l'eclairage tertiaire et industriel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-02-01

    This document provides information on the systems of possibilities of power variation and their management for an industrial lightning in the ternary sector. These systems bring to the user more comfort and improve the global and environmental cost of the installations. (A.L.B.)

  2. Developments in nanocrystalline magnetic alloys for industry; Alliages magnetiques nanocristallins industriels. Etat de l'art et evolution

    Energy Technology Data Exchange (ETDEWEB)

    Waeckerle, T.; Cremer, P. [Imphy Ugine Precision, 92 - Paris la Defense (France); Gautard, D. [Mecagis, 45 - Amilly (France)

    2003-10-01

    The French industrial production of nanocrystalline precursor ribbon (Imphy Ugine Precision - IUP) and nanocrystalline wound cores (Mecagis) is now mature, promoting then one of the first worldwide provider in this market. Recent progress in ribbon elaboration will provide large increase of industrial efficiency, leading the cost of a nanocrystalline solution to be closed to the cost of a ferrite solution. The precise study and control of magnetoelastic energy allowed the production scattering to be reduced, the alloy to be weakly dependant on external stresses (production, packaging, thermal dilatation), further promoting the performances. Whatever the alloy is very brittle in the nanocrystalline state, some improvements are using or are going around this intrinsic behaviour, and are now developed: powder core for low dissipative filtering, cut core for storage and strong power transformation, wound cores from ribbon nano-crystallized with high stresses during annealing, for the storage and current metering. (authors)

  3. Hybrid adsorption compression for industrial applications. HYACINT. Public final report; Hybride Adsorptie Compressie voor Industriele Toepassingen. HYACINT. Openbare eindrapportage

    Energy Technology Data Exchange (ETDEWEB)

    Van der Pal, M. [ECN Biomass and Energy Efficiency, Petten (Netherlands)

    2013-06-15

    Heat driven heat transformers can upgrade heat from about 100-120 degrees Celsius to 180-200C. However, most of the waste heat is below 100C. The hybrid adsorption compression technology offers the possibility to upgrade by at least 50C. The hybrid concept combines a heat-driven heat transformer with a power-driven compression heat pump. As part of the HYACINT project it has been examined which components of the two heat pump technologies are the most suitable for application in a hybrid heat transformer. This is done through a literature survey, sociological research, model calculations and measurements of components [Dutch] Warmtegedreven warmtetransformatoren kunnen warmte vanaf ongeveer 100 a 120C opwaarderen tot warmte van 180 tot 200C. Het merendeel van de restwarmte bevindt zich echter onder 100C. De hybride adsorptie compressie technologie biedt de mogelijkheid om ook deze warmte met tenminste 50C te kunnen verhogen. Het hybride concept combineert een warmtegedreven warmtetransformator met een arbeid aangedreven compressie warmtepomp. Binnen het HYACINT project is onderzocht welke componenten van beide warmtepomp technologieen het meest geschikt zijn voor toepassing in een hybride warmtetransformator. Dit is gedaan door middel van literatuurstudie, sociaalwetenschappelijk onderzoek, toepassingspotentieelonderzoek, modelberekeningen en metingen aan componenten.

  4. Multi-domain comparison of safety standards; Comparaison de normes de securite-innocuite de plusieurs domaines industriels

    Energy Technology Data Exchange (ETDEWEB)

    Baufreton, Ph.; Derrien, J.C.; Ricque, B. [Sagem Defense Securite, 75 - Paris (France); Blanquart, J.P. [Astrium Satellites, France (France); Boulanger, J.L. [CERTIFER, 75 - Paris (France); Delseny, H. [Airbus, 31 - Toulouse (France); Gassino, J. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Ladier, G. [Airbus / Aerospace Valley, 31 - Toulouse (France); Ledinot, E. [Dassault Aviation, 92 - Saint Cloud (France); Leeman, M. [Valeo, 75 - Paris (France); Quere, Ph. [Renault, 75 - Paris (France)

    2011-07-01

    This paper presents an analysis of safety standards and their implementation in certification strategies from different domains such as aeronautics, automation, automotive, nuclear, railway and space. This work, performed in the context of the CG2E ('Club des Grandes Entreprises de l'Embarque'), aims at identifying the main similarities and dissimilarities, for potential cross-domain harmonization. We strive to find the most comprehensive 'trans-sectorial' approach, within a large number of industrial domains. Exhibiting the 'true goals' of their numerous applicable standards, related to the safety of system and software, is a first important step towards harmonization, sharing common approaches, methods and tools whenever possible. (authors)

  5. Addressing safety issues through a joint industry programme; Traiter des problemes de securite a travers un programme industriel commun

    Energy Technology Data Exchange (ETDEWEB)

    Pool, G.; Williams, T.P. [BG Technology (United Kingdom); Jones, A.M. [Health and Safety Executive (United Kingdom)

    2000-07-01

    In an increasingly fragmented gas market, the focus for national gas safety may not rest with one major utility or gas supplier but may be spread across many companies. There will also be many new organisations in a liberalized gas industry with varying views on the needs and benefits of safety related technology development but all agree there is a need to ensure that the good safety record of gas as a domestic fuel is maintained. The number of carbon monoxide (CO) incidents is not decreasing significantly despite an increased awareness of the problem. As a consequence, a two-year joint industry programme addressing issues related to carbon monoxide has been established, co-ordinated by BG Technology and supported by gas organisations, government agencies, manufacturers and suppliers across Europe and the World. The 2-year 2 pound million programme has been constructed as twelve separate projects addressing issues such as the reporting and analysis of domestic incidents, improved service or installation practice, CO alarm reliability and information dissemination. The paper gives results and achievements of the programme, through new techniques, standards, procedures or equipment and demonstrates how the gas industry can work together to meet common safety objectives. (authors)

  6. Space heating in buildings: thermal diagnosis of an industrial building; Chauffage des batiments: bilan thermique d`un batiment industriel

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, R.

    1996-12-31

    The various heat transfer equations used for calculations in thermal diagnosis of an industrial building are reviewed: calculation of the heat losses through walls as a function of building materials, calculation of the energy consumption for heating fresh air (as a function of the air pollution rate in the building), calculation of the total heat losses, the heating energy demand and the annual energy consumption. Data concerning building materials characteristics, insulation and heating loads in the various regions of France, are also presented

  7. Increased value creation by industrial refining of natural gas in Norway?; Oekt verdiskaping gjennom industriell foredling av naturgass i Norge?

    Energy Technology Data Exchange (ETDEWEB)

    Skjelvik, John Magne; Kjelland, Torunn; Bakke, Kathrine Stene; Pedersen, Kent Vincent; Roetnes, Rolf; Fjose, Sveinung

    2009-07-01

    The report assesses whether industrial exploitation of gas from possible new major discoveries outside Northern Norway could be profitable. Profits are uncertain and depend heavily on gas prices. Industrial exploitation depends on the development of large, capital-intensive plants. This increases the financial risk. In a situation where gas export through pipelines is not an alternative industrial exploitation could be an alternative to export via LNG. However, the pressing demand for natural gas increases the probability of the construction of a gas export pipeline as a realistic alternative. Carbon emissions from a land-based plant will be important, and the costs for trading of emission allowances will reduce profitability. The authorities should allow for industrial exploitation to be assessed on an equal basis with pipeline transport and LNG as alternatives for market solutions for major new discoveries. (EW)

  8. Electro-coagulation applied to the treatment of industrial effluents; Electrocoagulation appliquee en traitement des effluents industriels

    Energy Technology Data Exchange (ETDEWEB)

    Laplace, C.; Leboucher, G.; Coste, M. [Anjou Recherche, Vivendi Water, 78 - Maisons-Laffitte (France)

    2001-07-01

    The electro-coagulation is a water treatment technic in electrolysis cell with double anode. In substitution to the coagulant reagent often used in water de-pollution, it realizes also the coloring decomposition, the DCO abatement and sometimes improving the sludges processing. The technic presents meanwhile some limitations as its poor treatment capacity and the necessity of a high effluent conductivity. An example of application shows that this technic is economically competitive. (A.L.B.)

  9. Evaluation of an industrial gas-fired IR dryer; Utvaerdering av en industriell gaseldad IR-straalare

    Energy Technology Data Exchange (ETDEWEB)

    Stenstroem, S; Hermodsson, S

    1994-11-01

    The IR dryer is used in a paper making machine to dry the paper web after it has been coated with a surface layer. In part 1 of the project a mathematical model have been developed, capable of calculating the radiation intensity and other energy flows in the dryer. In part 2 of the project, measurements have been made on the IR radiator mounted in the paper making machine. The calculation model shows the efficiency of the radiator to 39% at full power and 35% at half power. The direct measurements were made at half power and gave an efficiency of 31% for new radiators and 28% for old ones. The conclusion is that the calculation model values corresponds very well compared with direct measurements.

  10. Radioprotection optimization in the electronuclear, industrial and medical domains; Optimisation de la radioprotection dans les domaines electronucleaire, industriel et medical

    Energy Technology Data Exchange (ETDEWEB)

    Schieber, C.; Abela, G.; Ammerich, M.; Balduyck, S.; Batalla, A.; Drouet, F.; Fracas, P.; Gauron, Ch.; Le Guen, B.; Lombard, J.; Mougnard, Ph.; Murith, Ch.; Rannou, A.; Rodde, S.; Selva, M.; Tranchant, Ph.; Schieber, C.; Solaire, T.; Le Tonqueze, Y.; Jolivet, P.; Chauveau, D.; Mathevet, L.; Juhel, T.; Mertz, L.; Bochud, F.O.; Desmaris, G.; Turquet de Beauregard, G.; Roy, C.; Delacroix, S.; Sevilla, A.; Rehel, J.L.; Bernhard, S.; Palut-Laurent, O.; Lochard, J.; Lebaron-Jacobs, L.; Wack, G.; Barange, K.; Delabre, H.

    2011-07-01

    This document gathers the slides of the available presentations given during these conference days. Thirty one presentations are assembled in the document and deal with: 1 - implementation of the ALARA principle in the nuclear, industrial and medical domains: status and challenges (C. Schieber); 2 - image quality and scanner irradiation: what ingredients to chose? (T. Solaire); 3 - radioprotection stakes and implementation of the ALARA approach during the IFMIF design (Y. Le Tonqueze); 4 - ALARA at the design stage of the EPR (P. Jolivet); 5 - alternative techniques to iridium 192 gamma-graphy for welds control: results and recommendations from the ALTER-X project (D. Chauveau); 6 - alternative techniques to ionizing radiations use in the medical domain: implementation of navigation strategies (L. Mathevet); 7 - justification of ionizing radiations use in non-medical imaging: overview of the French situation and perspectives status (S. Rodde); 8 - ISOE: task scheduling for radioprotection optimization in nuclear power plants (G. Abela); 9 - Practices and ALARA prospects among big nuclear operators (T. Juhel); 10 - experience feedback on the use of diagnostic reference levels (DRLs) in diagnostic imaging optimization (L. Mertz); 11 - DRLs: Swiss strategy and concept limits (F.O. Bochud); 12 - external dosimetry tools: the existing, the developing and the remaining problems (A. Rannou); 13 - is the optimization principle applicable to the aircraft personnel's exposure to cosmic radiation? (G. Desmaris); 14-15 - experience feedback of the ALARA approach concerning an operation with strong dosimetric stakes (P. Mougnard and N. Fontaine); 16 - optimization of reactor pool decontaminations ((P. Tranchant); 17 - radiopharmaceuticals transport - ALARA principle related stakes (G. Turquet de Beauregard); 18 - ALARA in vet radio-diagnosis activity: good practices guide (C. Roy); 19 - implementation of the ALARA approach at the Proton-therapy centre of Orsay's Curie Institute (S. Delacroix); 20 - optimisation of operational extremity dosimetry using video (A. Sevilla); 21 - optimization in interventional radiology: uterine embolization (J.L. Rehel); 22 - optimization of radon-linked radioprotection in dwellings (C. Murith); 23 - NORMS companies prospects about optimization implementation (S. Bernhard); 24 - management of sites and soils polluted by radioactive compounds: synthesis of the works made by the pluralist thinking group on the approach to implement for the rehabilitation of radio-contaminated sites (O. Palut-Laurent); 25 - new stakes of radioprotection optimization: CIPR's publications no. 101 and 103 (J. Lochard); 26 - place of optimization in the new basic safety standards (BSS) of the European Commission (J. Lebaron-Jacobs); 27 - ASN's point of view on ALARA's consideration (G. Wack); 28 - initiatives related to radioprotection culture (B. Le Guen); 29 - training as optimization tool (K. Barange); 30 - networks as spreading relays for radioprotection training and culture (A. Batalla); 31 - the radio-protectionists' club (H. Delabre and L. d'Ascenzo). (J.S.)

  11. Optimisation de géocomposites pour la filtration de boues dans le cadre d’un partenariat industriel

    Directory of Open Access Journals (Sweden)

    TOUZE-FOLTZ, Nathalie

    2016-03-01

    Full Text Available L’industrie minière produit d’énormes quantités de déchets sous forme de boue minérale, comme les résidus fins matures issus de l’exploitation des sables bitumineux. Ces boues sont difficiles à assécher de par leur forte argilosité et leur stockage reste complexe à gérer pour des raisons pratiques, économiques et environnementales. Des recherches sont donc initiées pour étudier la possibilité d’utiliser des produits géosynthétiques, et en particulier leur fonction de filtration, pour favoriser l’assèchement de ces boues minières.

  12. Vérification des propriétés temporisées des automates programmables industriels

    OpenAIRE

    Bel Mokadem , Houda

    2006-01-01

    Vérification of PLC (Programmable Logic Controller) programs is important when these programs are to control critical applications for reactive systems. This context has already been study for untimed programs. In this thesis, We are interested to timed properties and programs. More precisely, we give a formal semantics to (partial) Ladder diagrams and TON blocks, with timed automata. We also propose a timed logic in order to abstract transient events, where transient properties is parameteri...

  13. TOP-Energy - toolkit for optimization of industrial energy systems; TOP-Energy - Softwaregestuetzte Analyse und Optimierung industrieller Energieversorgungssysteme

    Energy Technology Data Exchange (ETDEWEB)

    Augenstein, E.; Kuperjans, I. [RWTH Aachen (Germany); Wrobel, G. [Gesellschaft zur Foerderung angewandter Informatik e.V. (GFal), Berlin (Germany); Gruezenich, D.

    2004-07-01

    The contribution presents the software package 'TOP-Energy' which supports energy consultants in their analysis and optimisation of industrial energy systems and is a tool for development and assessment of measures for reducing the energy cost and the consumption of energy resources. In particular, it supports data acquisition, evaluation, and presentation of results of routine work; it offers simulations of complel processes and systems as well as tools like integrated project management. TOP-Energy consists of several modules linked by a common framework. The framework is for data management, module integration and control, and offers a user interface in the form of adaptable editors, dialogues and menus. Power supply systems of industrial works can be modelled with all their components. The key module of Top-energy is a simulator for systems designed, with variable temporal load curves and other boundary conditions. (orig.)

  14. Industrial radioactive wastes: what are we talking about?; Les dechets radioactifs industriels: de quoi s'agit-il?

    Energy Technology Data Exchange (ETDEWEB)

    Le Bars, Y. [Agence Nationale pour la Gestion des Dechets Radioactifs ANDRA, 92 - Chatenay Malabry (France)

    2001-07-01

    The subject of radioactive wastes is developed through their origin, their classification, their scale of size. The different storage centers are given and the new channels of radioactive wastes management are tackled. The particular case of high level and long term radioactive wastes is detailed. (N.C.)

  15. Computersystemen ter ondersteuning en vervanging van industriele persoonlijke verkoop : een studie vanuit de marktkunde, informatica en operationele research

    NARCIS (Netherlands)

    Kerkhoven, J.

    1991-01-01

    This thesis concerns computer systems which can either support or replace Industrial Personal Selling (IPS)

    The objective of this study is to determine how IPS can be supported or replaced efficiently and effectively by computer systems. IPS is defined here as all activities aimed at face-to-face

  16. COMPORTEMENT D’UN COMPOSITE FINE MINERALE - CIMENT - BOIS ELABORE A L’AIDE DE DECHETS INDUSTRIELS SOLIDES

    Directory of Open Access Journals (Sweden)

    M.L BENMALEK

    2000-06-01

    Full Text Available Cinq fines minérales de différentes natures minéralogiques et de limites granulaires comparables, provenant de résidus de carrières, ont été traitées pour former le squelette de bétons légers d'isolation. L'allégement et le pouvoir isolant de ces bétons sont obtenus par incorporation de granulats de bois dans la matrice constituée de la fine minérale et de ciment CPA. Dans une première partie, sont présentés les principales caractéristiques de ces fines: minéralogie par diffraction des rayons X, granulométrie laser, conductivité thermique à la sonde TPS, densité sèche par  pycnomètrie à eau et morphologie au MEB. La seconde partie est consacrée à l'étude de l'effet des caractéristiques de ces fines sur le comportement hygrothermique et mécanique de ces bétons élaborés avec une même formulation et un même état de consistance au moulage. Les résultats de ces investigations sont comparés entre eux et  à ceux de matériaux usuels classés par la RILEM.

  17. Reform of the energy supply law seen from the point of view of industry. Reform des Energiewirtschaftsgesetzes aus industrieller Sicht

    Energy Technology Data Exchange (ETDEWEB)

    Boeke, E; Heller, W [Bundesverband der Deutschen Industrie e.V., Koeln (Germany). Abt. Energiepolitik

    1991-12-01

    From the point of view of the industry there are two central tasks for energy policy: 1. The German industry has to be supplied with energy at internationally competitive prices. Of the three aims of energy policies (secure, moderately-priced and low-pollutant) the price aim is in Germany in the international comparison today least achieved. 2. Policy always tends to solve problems in the energy sector with regulating interventions. An example from the recent past is the power supply law of 1990 with which the way to political electricity pricing was opened. But the precept for the Internal European Market has to be deregulation as only in this way the larger market can develop its efficiency to the full. Also for the reform of the economic laws concerning the energy sector these two central aims have to be taken into consideration. (orig./UA).

  18. Nouvelle methode d'integration energetique pour la retro-installation des procedes industriels et la transformation des usines papetieres

    Science.gov (United States)

    Bonhivers, Jean-Christophe

    The increase in production of goods over the last decades has led to the need for improving the management of natural resources management and the efficiency of processes. As a consequence, heat integration methods for industry have been developed. These have been successful for the design of new plants: the integration principles are largely employed, and energy intensity has dramatically decreased in many processes. Although progress has also been achieved in integration methods for retrofit, these methods still need further conceptual development. Furthermore, methodological difficulties increase when trying to retrofit heat exchange networks that are closely interrelated to water networks, such as the case of pulp and paper mills. The pulp and paper industry seeks to increase its profitability by reducing production costs and optimizing supply chains. Recent process developments in forestry biorefining give this industry the opportunity for diversification into bio-products, increasing potential profit margins, and at the same time modernizing its energy systems. Identification of energy strategies for a mill in a changing environment, including the possibility of adding a biorefinery process on the industrial site, requires better integration methods for retrofit situations. The objective of this thesis is to develop an energy integration method for the retrofit of industrial systems and the transformation of pulp and paper mills, ant to demonstrate the method in case studies. Energy is conserved and degraded in a process. Heat can be converted into electricity, stored as chemical energy, or rejected to the environment. A systematic analysis of successive degradations of energy between the hot utilities until the environment, through process operations and existing heat exchangers, is essential in order to reduce the heat consumption. In this thesis, the "Bridge Method" for energy integration by heat exchanger network retrofit has been developed. This method is the first that considers the analysis of these degradations. The fundamental mechanism to reduce the heat consumption in an existing network has been made explicit; it is the basis of the developed method. The Bridge Method includes the definition of "a bridge", which is a set of modifications leading to heat reduction in a heat exchanger network. It is proven that, for a given set of streams, only bridges can lead to heat savings. The Bridge Method also includes (1) a global procedure for heat exchanger network retrofit, (2) a procedure to enumerate systematically the bridges, (3) "a network table" to easily evaluate them, and (4) an "energy transfer diagram" showing the effect of the two first principles of thermodynamics of energy conservation and degradation in industrial processes in order to identify energy savings opportunities. The Bridge Method can be used for the analysis of networks including several types of heat transfer, and site-wide analysis. The Bridge Method has been applied in case studies for retrofitting networks composed of indirect-contact heat exchangers, including the network of a kraft pulp mill, and also networks of direct-contact heat exchangers, including the hot water production system of a pulp mill. The method has finally been applied for the evaluation of a biorefinery process, alone or hosted in a kraft pulp mill. Results show that the use of the method significantly reduces the search space and leads to identification of the relevant solutions. The necessity of a bridge to reduce the inputs and outputs of a process is a consequence of the two first thermodynamics principles of energy conservation and increase in entropy. The concept of bridge alone can also be used as a tool for process analysis, and in numerical optimization-based approaches for energy integration.

  19. Industrial solid and liquid waste treatment processes; Les procedes de traitement des dechets industriels solides et liquides

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1995-11-01

    This catalogue gives information on 68 chemical, mechanical, magnetic, electrical, thermal, etc. techniques for the processing of solid, viscous and liquid, common or special, industrial wastes. The various processes are presented as files, which are easily retrievable through keywords, waste type or industry codes, processing types, distributors. Technologies, performances and applications of each techniques are presented, together with references and company contacts

  20. Operational measurements in radioprotection in the industrial and medical environments; Mesures operationnelles en radioprotection dans les milieux industriel et medical

    Energy Technology Data Exchange (ETDEWEB)

    Rodde, S.; Vial, Th.; Truffert, H.; Kramar, R.; Batalla, A.; Roine, Ph.; Pin, A.; Lahaye, Th.; Rodde, S.; Bordy, J.M.; Paquet, F.; Veres, A.; Cadiou, A.; Branthonne, J.Y.; Noel, A.; Laloubere, L.; Moreau, St.; Gensdarmes, F.; Marques, S.; Lestang, M.; Valendru, N.; Tranchant, Ph.; Martel, P.; Bernhard, S.; Chareyre, P.; Gardin, I.; Casanova, Ph.; De Vita, A.; Tenailleau, L.; Masson, B.; Feret, B.; Guerin, M.; Guillot, L.; Gaultier, E.

    2009-07-01

    This document gathers the slides of the available presentations given during these conference days. Thirty presentations are assembled in the document and deal with: 1 - enforcement circular of the labor code dispositions relative to workers protection against ionizing radiation hazards (T. Lahaye); 2 - context and regulatory evolutions - public health code (S. Rodde); 3 - references and perspectives in external dosimetry (J.M. Bordy); 4 - CIPR's Committee 2 works (F. Paquet); 5 - from protection data to measurement data (A. Pin); 6 - dosimetric control in radiotherapy (A. Veres); 7 - calibration of irradiation measurement devices in industrial environment (A. Cadiou); 8 - calibration and verification of nuclear measurement devices (J.Y. Branthonne); 9 - calibration of measurement devices in medical environment (J.M. Bordy); 10 - quality control in radiotherapy (A. Batalla); 11 - in-vivo dosimetry in radiotherapy (A. Noel); 12 - calibration metrology of fixed post irradiation sensors (L. Laloubere); 13 - design requirements for the radiological zoning and the wastes cleanliness of Flamanville 3 EPR reactor (S. Moreau); 14 - efficiency of aerosol capture systems used in CNPE EDF (F. Gensdarmes); 15 - mobile surveillance means of the atmospheric contamination of CNPE EDF's reactor building (S. Marques and M. Lestang); 16 - experience feedback about the security gates at EDF's nuclear facilities (N. Valendru); 17 - metrology needs for radioprotection technical controls (P. Tranchant); 18 - technical evaluation of a flowmeter/dosemeter in the framework of the regulatory control of X-ray electric generators used in radio-diagnosis (P. Martel); 19 - reinforced natural radioactivity - the case of radon measurement (S. Bernhard); 20 - fires during radioactive materials transport (P. Chareyre); 21 - measurement in the framework of medical examinations: radiology service (A. Noel); 22 - operational measurements in nuclear medicine (I. Gardin); 23 - from the operational measurement to the committed dose (P. Casanova); 24 - technical control of environments in basic nuclear facilities (A. De Vita); 25 - measurements of the atmospheric tritium with low volume activity level in nuclear facilities (PREVAIR system) (L. Tenailleau); 26 - measurement of high-energy pulsed neutrons (B. Masson); 27 - dismantling: from in-situ measurement to final package characterization (B. Feret); 28 - new generation of detection and identification equipments (M. Guerin); 29 - DIRAD{sup TM} - radionuclides detection and identification - system of automatic detection and identification of moving radioactive objects (L. Guillot); 30 - ViewPoint - radioprotection supervision post (E. Gaultier). (J.S.)

  1. L'état de l'environnement industriel français est-il objectivement mesurable ?

    OpenAIRE

    Gotteland, David; Boulé, Jean-Marie

    2004-01-01

    Working paper serie RMT (WPS 04-07); the environmental state is a frequently modelized variable in marketing research. This paper proposes updated scales and panorama of the objective state of 58 industrial branches defined by INSEE based on three characteristic dimensions: dynamism, complexity and capacity.; l'état de l'environnement est une variable fréquemment modélisée dans la recherche en marketing. A partir de ce constat, cet article propose une échelle de mesure et un panorama actualis...

  2. Substitution af organiske opløsningsmidler med estre af vegetabilske olier ved industriel rengøring

    DEFF Research Database (Denmark)

    Jacobsen, Thomas

    1999-01-01

    Substitution af organiske opløsningsmidler, der belaster både arbejdsmiljø og det ydre miljø, er ofte mulig. På offset trykkerier kan rengøring i stor udstrækning ske med produkter, der hverken er flygtige eller giftige for mennesker og miljø. Tilsyneladende kan disse erfaringer også overføres ti...

  3. Globaliserede byrum

    DEFF Research Database (Denmark)

    Reeh, Henrik

    2006-01-01

    Arkitektur, globalisering, byteori, bykultur, Islands Brygge, samtidskunst, udstillingssteder, gentrification, industriel arkitektur......Arkitektur, globalisering, byteori, bykultur, Islands Brygge, samtidskunst, udstillingssteder, gentrification, industriel arkitektur...

  4. 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)

  5. 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)

  6. Investigations of the chemical states of carrier-free phosphorus-32 as extracted into water from pile-irradiated sulphur; Recherches sur les etats chimiques du phosphore-32 sans entraineur obtenu par extraction aqueuse a partir de soufre irradie dans un reacteur; Issledovanie khimicheskogo sostoyaniya svobodnogo ot nositelya fosfora-32 pri izvlechenii ego v vodu iz obluchennoj v yadernom reaktore sery; Estudio de los estados quimicos del fosforo-32 libre de portador que se obtiene por extraccion acuosa del azufre irradiado en un reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, J B; Birkelund, O R [Institutt for Atomenergi, Kjeller, Lillestrom (Norway)

    1962-01-15

    One of the methods of producing carrier free phosphorus-32 today is by extraction into water from pile-irradiated sulphur. The present work gives information concerning the chemical states of P{sup 32} in aqueous solutions at different steps of the routine production-process. The variation in the chemical state of P{sup 32} compounds in the final product has also been examined as a function of storage time. P{sup 32} bound as orthophosphate was found to be the main component. During the chemical processing, the amount of orthophosphate increased from about 70% at the beginning of the extraction to about 98 % in the final carrier-free P{sup 32} product. The residual amount consisted of a mixture of pyro-, tri-, tetra-, and other long-chain polyphosphates (number of P {>=} 5). No metaphosphates (ring-formed) were found in the solutions during production and storage. The results indicate that the polyphosphorus compounds were formed in the target material during irradiation. Special attention was paid to the adsorption of carrier-free P{sup 32} compounds to glassware under the existing experimental conditions. (author) [French] L'une des methodes employee a l'heure actuelle pour obtenir du phosphore-32 sans entraineur consiste a l'extraire dans l'eau a partir de soufre irradie dans un reacteur. Les auteurs donnent des indications sur l'etat chimique du phosphore-32 dans des solutions aqueuses, a differentes etapes du processus de preparation courant. Us examinent aussi les changements de l'etat chimique des composes du phosphore-32 dans le produit final en fonction de la duree de stockage. On a constate que le {sup 32}P combine sous forme d'orthophosphate etait le principal composant. Au cours du traitement chimique, la teneur en orthophosphate est passee d'environ 70% au debut de l'extraction a environ 98% lors de l'obtention du produit final sans entraineur. Le reste etait constitue d'un melange de pyro-, tri-, tetra- et autres polyphosphates a chaine longue (P

  7. Blowing loop in the EL-4 reactor: CO{sub 2} flow control analogue study; Boucle de soufflage de la centrale EL-4 - regulation du debit CO{sub 2} - etude analogique

    Energy Technology Data Exchange (ETDEWEB)

    Chazal, G; Merle, J P; Guillemard, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leroy, C; Robin, L; Jacquin, J C; Cornudet, A [Societe INDATOM, France (France)

    1966-07-01

    This report describes one study which contributed to the construction of the Monts d'Arree nuclear power station: EL-4. The reactor is cooled by a CO{sub 2} current provided by 3 turbo-blower groups. The priming vapour for the turbines is taken at the exit of the main CO{sub 2} - H{sub 2}O exchangers. The operation of EL 4 is based on a high degree of centralization of the controls which attributes an important role to the general regulation circuits. This general regulation includes in particular an internal blowing loop which controls the CO{sub 2} flow. The study of the control of this CO{sub 2} flow is made up of 3 parts: - analogue representation of the reactors cooling circuit and of the turbo blower unit. - first test campaign using the analogue computer describing the natural behaviour of the system in the absence of control. theoretical determination of the regulation factors; definition of the regulation using an analogue computer and second test campaign for recording the performances of the blowing loop. The 4. part of the report deals with the analogue study: analogue equations - development. (authors) [French] Ce rapport prend place parmi les etudes de realisation de la Centrale des Monts d'Arree EL-4. Le reacteur est refroidi par une circulation de CO{sub 2} assuree par 3 groupes turbosoufflantes. La vapeur d'entrainement des turbines est prelevee a la sortie des echangeurs principaux CO{sub 2} - H{sub 2}O. L'exploitation de EL-4 repose sur une centralisation poussee des moyens de controle-commande qui attribue un role essentiel aux circuits de regulation generale. Cette regulation generale comporte en particulier une boucle interne de soufflage qui realise un asservissement du debit de CO{sub 2}. L'etude de cette regulation du debit CO{sub 2} comprend 3 parties: - representation analogique du circuit de refroidissement du reacteur et de l'ensemble turbine-soufflante. - premiere campagne d'essais sur calculateur analogique decrivant le comportement

  8. Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [French] La France, qui a base son effort principal pour la production d'energie nucleaire sur la filiere des reacteurs a uranium naturel et graphite refroidis par gaz, et qui a un programme a plus

  9. Oxidation of steel heated in CO{sub 2} medium under pressure; Oxydation d'un acier ordinaire chauffe dans le gaz carbonique sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Darras, R.; Leclercq, D.; Bunard, C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Behaviour of low-alloyed steels heated in CO{sub 2} medium under pressure is reported. Tests are carried out in the range of erature reached in the CO{sub 2} cooled reactors (vessel, thermal shield, pipes). The observed weight increases are small, even after more than a thousand hours of heating at 350 deg. C, but oxidation curve looks like progressing linearly. Furthermore, the oxide formed under a pressure of 15 kg/cm{sup 2} is undoubtedly more compact and adherent than the one formed under a pressure of 1 kg/cm{sup 2}. Finally, for practical use, CO{sub 2} steel pipes surface has to be sand blast and pickled. A following phosphatizing protects it from atmospheric corrosion during assembling, but these treatments have no influence on the behaviour of these steels heated in CO{sub 2}. (author)Fren. [French] On etudie le comportement d'aciers au carbone faiblement allies, chauffes dans le gaz carbonique sous pression, aux temperatures atteintes dans les reacteurs refroidis par ce gaz (caisson, bouclier thermique, canalisations). Les augmentations de poids observees sont faibles, meme apres plus de 1000 heures de chauffage a 350 deg. C, mais l'oxydation semble se poursuivre lineairement. De plus, l'oxyde forme dans le gaz carbonique sous pression de 15 kg/cm{sup 2} est nettement plus compact et adherent que celui forme sous pression atmospherique de gaz carbonique. Enfin, dans la pratique, les surfaces d'acier du circuit de gaz carbonique sont necessairement sablees ou decapees; une phosphatation ulterieure le protege de la corrosion atmospherique pendant le montage. Ces traitements sont sans influence sur le comportement de ces aciers dans le gaz carbonique a chaud. (auteur)

  10. Heat exchange and pressure drop of herring-bone fin surfaces. Experimental cell results at constant wall temperature; Echange de chaleur et perte de charge de surfaces a ailettes en chevrons. Resultats experimentaux en cellule a temperature de paroi constante

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-07-01

    The increase in the specific power of nuclear reactors of the gas-graphite type has necessitated the use of high performance exchange surfaces for canning the fuel (natural uranium). For this, experiments were carried out on cans fitted with herring-bone fins, at constant wall temperature; a flow of water at 100 deg. C passes inside the can which is cooled externally by a flow of CO{sub 2} at 15 bars pressure. This experimental set-up makes it possible to compare the aero-thermal performances of the different cans with an accuracy of 5 per cent. This report presents the results obtained in the form of a friction coefficient f{sub 0} and mean Margoulis number m{sub 0} as a function of the Reynolds number Re{sub 0}, this latter varying from 3 x 10{sup 5} to 9 x 10{sup 5}. (authors) [French] L'augmentation de la puissance specifique des reacteurs nucleaires de la filiere graphite-gaz a necessite l'utilisation de surfaces d'echange a hautes performances pour gainer le combustible (uranium naturel). Dans cette optique, des gaines munies d'ailettes disposees en chevron ont ete experimentees a temperature de paroi constante: un courant d'eau a 100 deg. C circule a l'interieur de la gaine qui est refroidie exterieurement par un ecoulement de CO{sub 2} sous une pression de 15 bars. Cette methode experimentale permet de situer les performances aerothermiques des gaines les unes par rapport aux autres a 5 pour cent pres. Ce rapport presente les resultats obtenus sous la forme d'un coefficient de frottement f{sub 0} et d'un nombre de Margoulis moyen m{sub 0} en fonction du nombre de Reynolds Re{sub 0}, ce dernier pouvant varier de 3. 10{sup 5} a 9. 10{sup 5}. (auteurs)

  11. Oxidation of steel heated in CO{sub 2} medium under pressure; Oxydation d'un acier ordinaire chauffe dans le gaz carbonique sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Darras, R; Leclercq, D; Bunard, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Behaviour of low-alloyed steels heated in CO{sub 2} medium under pressure is reported. Tests are carried out in the range of erature reached in the CO{sub 2} cooled reactors (vessel, thermal shield, pipes). The observed weight increases are small, even after more than a thousand hours of heating at 350 deg. C, but oxidation curve looks like progressing linearly. Furthermore, the oxide formed under a pressure of 15 kg/cm{sup 2} is undoubtedly more compact and adherent than the one formed under a pressure of 1 kg/cm{sup 2}. Finally, for practical use, CO{sub 2} steel pipes surface has to be sand blast and pickled. A following phosphatizing protects it from atmospheric corrosion during assembling, but these treatments have no influence on the behaviour of these steels heated in CO{sub 2}. (author)Fren. [French] On etudie le comportement d'aciers au carbone faiblement allies, chauffes dans le gaz carbonique sous pression, aux temperatures atteintes dans les reacteurs refroidis par ce gaz (caisson, bouclier thermique, canalisations). Les augmentations de poids observees sont faibles, meme apres plus de 1000 heures de chauffage a 350 deg. C, mais l'oxydation semble se poursuivre lineairement. De plus, l'oxyde forme dans le gaz carbonique sous pression de 15 kg/cm{sup 2} est nettement plus compact et adherent que celui forme sous pression atmospherique de gaz carbonique. Enfin, dans la pratique, les surfaces d'acier du circuit de gaz carbonique sont necessairement sablees ou decapees; une phosphatation ulterieure le protege de la corrosion atmospherique pendant le montage. Ces traitements sont sans influence sur le comportement de ces aciers dans le gaz carbonique a chaud. (auteur)

  12. Study of heat transfer in superconducting cable electrical insulation of accelerator magnet cooled by superfluid helium; Etude des transferts de chaleur dans les isolations electriques de cables supraconducteurs d'aimant d'accelerateur refroidi par helium superfluide

    Energy Technology Data Exchange (ETDEWEB)

    Baudouy, B

    1996-10-04

    Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

  13. La détection infrarouge avec les plans focaux non refroidis : état de l'artUncooled focal plane infrared detectors: the state of the art

    Science.gov (United States)

    Tissot, Jean-Luc

    2003-12-01

    The emergence of uncooled detectors has opened new opportunities for IR detection for both military and commercial applications. Development of such devices involves a lot of trade-offs between the different parameters that define the technological stack. These trade-offs explain the number of different architectures that are under worldwide development. The key factor is to find a high sensitivity and low noise thermometer material compatible with silicon technology in order to achieve high thermal isolation in the smallest area as possible. Ferroelectric thermometer based hybrid technology and electrical resistive thermometer based (microbolometer) technology are under development. LETI and ULIS have chosen from the very beginning to develop first a monolithic microbolometer technology fully compatible with commercially available CMOS technology and secondly amorphous silicon based thermometer. This silicon approach has the greatest potential for reducing infrared detector manufacturing cost. After the development of the technology, the transfer to industrial facilities has been performed in a short period of time and the production is now ramping up with ULIS team in new facilities. LETI and ULIS are now working to facilitate the IRFPA integration into equipment in order to address a very large market. Achievement of this goal needs the development of smart sensors with on-chip advanced functions and the decrease of manufacturing cost of IRFPA by decreasing the pixel pitch and simplifying the vacuum package. We present in this paper the technology developed by CEA/LETI and its improvement for being able to designs 384×288 and 160×120 arrays with a pitch of 35 μm. Thermographic application needs high stability infrared detector with a precise determination of the amount of absorbed infrared flux. Hence, infrared detector with internal temperature stabilized shield has been developed and characterized. These results will be presented. To cite this article: J.-L. Tissot, C. R. Physique 4 (2003).

  14. Study of heat transfer in superconducting cable electrical insulation of accelerator magnet cooled by superfluid helium; Etude des transferts de chaleur dans les isolations electriques de cables supraconducteurs d'aimant d'accelerateur refroidi par helium superfluide

    Energy Technology Data Exchange (ETDEWEB)

    Baudouy, B

    1996-10-04

    Heat transfer studies of electrical cable insulation in superconducting winding are of major importance for stability studies in superconducting magnets. This work presents an experimental heat transfer study in superconducting cables of Large Hadron Collider dipoles cooled by superfluid helium and submitted to volume heat dissipation due to beam losses. For NbTi magnets cooled by superfluid helium the most severe heat barrier comes from the electrical insulation of the cables. Heat behaviour of a winding is approached through an experimental model in which insulation characteristics can be modified. Different tests on insulation patterns show that heat transfer is influenced by superfluid helium contained in insulation even for small volume of helium (2 % of cable volume). Electrical insulation can be considered as a composite material made of a solid matrix with a helium channels network which cannot be modelled easily. This network is characterised by another experimental apparatus which allows to study transverse and steady-state heat transfer through an elementary insulation pattern. Measurements in Landau regime ({delta}T{approx}10{sup -5} to 10{sup -3} K) and in Gorter-Mellink regime ({delta}T>10{sup -3} K) and using assumptions that helium thermal paths and conduction in the insulation are decoupled allow to determine an equivalent channel area (10{sup -6} m{sup 2}) and an equivalent channel diameter (25 {mu}). (author)

  15. A Survey of the Fuel Cycles Operated in the United Kingdom; Etude d'ensemble sur les cycles de combustible au Royaume-Uni; Obzor toplivnykh tsiklov, ispol'zuemykh v soedinennom korolevstve; Estudio de los ciclos de combustible utilizados en el Reino Unido

    Energy Technology Data Exchange (ETDEWEB)

    Allday, C. [United Kingdom Atomic Energy Authority, Risley, Warrington, Lancs (United Kingdom)

    1963-10-15

    son programme d'energie d'origine nucleaire sur la filie re des reacteurs a l'uranium naturel et au graphite, refroidis par un gaz. Les reacteurs de Calder Hall et de Chapelcross fonctionnent depuis sept ans; ceux de Berkeley et de Bradwell, qui dependent du Central Electricity Generating Board (CEGB), sont maintenant en service et sept autres reacteurs sont en construction ou en projet. Le combustible destine a ces reacteurs est etudie et fabrique dans l'usine de l'Atomic Energy Authority (AEA) a Springfields, puis transporte jusqu'au reacteur. Apres irradiation et dechargement, le combustible est transporte a l'usine de l'AEA, situee a Windscale, ou l'uranium et le plutonium sont separes des produits de fission. L'auteur decrit l'experience britannique en matiere d'etude et de fabrication du combustible, d'exploitation du reacteur, de transport et de traitement chimique du combustible irradie. Il examine brievement le comportement du combustible dans le reacteur et les differents programmes possibles de chargement et de dechargement, ce sujet etant developpe dans un autre memoire. b) Reacteurs utilisant les combustibles enrichis. Le Royaume-Uni met au point un reacteur perfectionne refroidi par un gaz (Advanced Gas Cooled Reactor = AGR), dont le prototype a ete mis en service en 1963. Le combustible est fabrique a partir d'oxyde d'uranium enrichi gaine d' acier inoxydable; apres irradiation, il sera traite dans une novelle installation qui sera ajoutee a l'usine de separation de Windscale ou est actu ellement traite le combustible gaine de magnox. L'uranium enrichi destine a AGR est produit par l'usine de diffusion situee a Capenhurst. L'oxyde d'uranium naturel enrichi au plutonium peut remplacer, comme combustible, l'oxyde d'uranium enrichi. L'auteur decrit l'experience acquise dans la transformation de l'oxyde destine a AGR et dans le fonctionnement du reacteur et indique comment on envisage de traiter le combustible irradie. Il examine le cas de l'utilisation d

  16. The effects of spots (or grains) and the mean work function of a polycrystalline emitter; Les effets des taches (ou grains) et le travail de sortie moyen d'un emetteur polycristallin

    Energy Technology Data Exchange (ETDEWEB)

    Devin, B; Phuc Nguyen, Xuan [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The direct conversion of thermal energy at high temperature to electrical energy using plasma diodes is studied in the Electronic Physics Service of the French Atomic Energy Commission. Research concerns the adaptation of these diodes to nuclear reactors with a view especially to providing power for operating the instruments of space vehicles. In parallel with the semi-industrial realizations and tests, an important part of the activity of the service is directed towards fundamental research into physical phenomena convected with thermionic emission with a view to improving present performances. (authors) [French] La conversion directe de l'energie thermique a haute temperature en energie electrique par diodes a plasma est etudiee au Service d'Electronique Physique du Commissariat a l'Energie Atomique. On etudie l'adaptation de ces diodes aux reacteurs nucleaires, notamment en vue de fournir l'energie de servitude dans les vehicules spatiaux. Parallelement aux realisations et essais semi-industriels, une part importante de l'activite du Service est orientee vers l'etude fondamentale des phenomenes physiques lies a la conversion thermoionique dans le but d'ameliorer les performances actuelles. (auteurs)

  17. Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor; Etude par simulation numerique des ecoulements turbulents reactifs dans les reacteurs d'oxydation hydrothermale: application a un reacteur agite double enveloppe

    Energy Technology Data Exchange (ETDEWEB)

    Moussiere, S

    2006-12-15

    Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

  18. Compression-absorption (resorption) refrigerating machinery. Modeling of reactors; Machine frigorifique a compression-absorption (resorption). Modelisation des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lottin, O; Feidt, M; Benelmir, R [LEMTA-UHP Nancy-1, 54 - Vandoeuvre-les-Nancy (France)

    1998-12-31

    This paper is a series of transparencies presenting a comparative study of the thermal performances of different types of refrigerating machineries: di-thermal with vapor compression, tri-thermal with moto-compressor, with ejector, with free piston, adsorption-type, resorption-type, absorption-type, compression-absorption-type. A prototype of ammonia-water compression-absorption heat pump is presented and modeled. (J.S.)

  19. Corrosion by cooling gases in nuclear reactors; la corrosion par les gaz caloporteurs dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Darras, R. [Commissariat a l' energie atomique et aux energies alternatives - CEA, Centre de Saclay, Section d' etude de la corrosion par gaz et metaux liquides (France)

    1960-07-01

    This article begins with a review of the various materials which can be used and the cooling gases in which they may be heated, emphasis being placed on the importance of reaching temperatures as high as possible. This is followed by a few general remarks on the dry oxidation of metals and alloys, particularly with regard to diffusion phenomena and their various possible mechanisms, and also the methods of investigation employed. Finally, the behaviour of the chief nuclear materials heated in the various gases is studied successively. Materials used for fuel (metallic uranium, uranium oxide, carbides and silicides), canning materials (magnesium, aluminium, zirconium, beryllium, stainless and refractory steels), structural materials (ordinary or slightly alloyed steels), and finally moderators (graphite, beryllium oxide) are deal with in this way. This account is backed up both by the results obtained at the CEA and by work published outside or abroad up to the present day. In conclusion, every effort has been made to direct future research on the basis of the foregoing. Reprint of a paper published in Industries Atomiques - no. 9/10, 1959, p. 3-23 [French] Dans cet article, on passe tout d'abord en revue les divers materiaux utilisables et les gaz de refroidissement dans lesquels ils peuvent etre chauffes, en insistant sur l'interet d'atteindre des temperatures aussi elevees que possible. On rappelle ensuite quelques generalites sur l'oxydation seche des metaux et alliages, notamment en ce qui concerne les phenomenes de diffusion et leurs divers mecanismes possibles ainsi que les methodes d'etude. Enfin, le comportement des principaux materiaux nucleaires chauffes dans les divers gaz est etudie successivement. On traita ainsi des materiaux combustibles (uranium metallique, oxyde, carbures et siliciures d'uranium), des materiaux de gainage (magnesium, aluminium, zirconium, beryllium, aciers inoxydables et refractaires), des materiaux de structure (aciers ordinaires ou faiblement allies), et enfin des moderateurs (graphite, oxyde de beryllium). Au cours de l'expose, on s'appuie a la fois sur les resultats obtenus au CEA et sur les travaux exterieurs ou etrangers publies a ce jour. En conclusion, on s'efforce de degager une orientation pour les recherches ulterieures. Reproduction d'un article publie dans Industries Atomiques - no. 9/10, 1959, p. 3-23.

  20. Compression-absorption (resorption) refrigerating machinery. Modeling of reactors; Machine frigorifique a compression-absorption (resorption). Modelisation des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lottin, O.; Feidt, M.; Benelmir, R. [LEMTA-UHP Nancy-1, 54 - Vandoeuvre-les-Nancy (France)

    1997-12-31

    This paper is a series of transparencies presenting a comparative study of the thermal performances of different types of refrigerating machineries: di-thermal with vapor compression, tri-thermal with moto-compressor, with ejector, with free piston, adsorption-type, resorption-type, absorption-type, compression-absorption-type. A prototype of ammonia-water compression-absorption heat pump is presented and modeled. (J.S.)

  1. Modelling of heterogenous neutron leakages in a nuclear reactor; Modelisation des fuites heterogenes de neutrons dans un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Wohleber, X

    1997-11-17

    The TIBERE Model is a neutron leakage method based on B{sub 1} heterogeneous transport equation resolution. In this work, we have studied the influence of the reflection mode at the boundary of the assembly. In particular the White boundary condition has been implemented in the APOLLO2 neutron transport code. We have compared the two TIBERE kinds of boundary conditions (specular and white) with the classical B{sub 1} homogeneous leakage method in the modelling of some reactors. We have remarked the better capability of the TIBERE Model to compute voided assemblies. The white boundary condition is also able to compute a completely voided assembly and, besides, wins a factor 10 in CPU time in comparison with the specular boundary condition. These two heterogenous leakage formalisms have been tested on a partially voided experiment and have shown that the TIBERE Model can compute this kind of situation with a greater precision than the classical B{sub 1} homogeneous leakage method, and with a shorter computational time. (author)

  2. New competition in the world market of nuclear reactors; La nouvelle concurrence sur le marche mondial des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Finon, D. [Centre National de la Recherche Scientifique (CNRS), CIRED (EHESS et CNRS), 75 - Paris (France)

    2005-06-01

    As nuclear orders are picking up a little, there are strengths competing against one another in the world industry of reactors, an industry that has been deeply affected for twenty years, by the smallness of the market and the reorganization of the electromechanical industry. Competition remains particularly difficult, even though, in terms of exports, national markets in industrialized countries such as the American market and European market are now open to foreign newcomers. One of the reasons of the difficulty is the increased commercial competition based on advanced reactor techniques untested due to strong faith in technology leading to forget the learning difficulties of older reactor types. On a narrow market, demanding and with very specific political interference, the reasoning is not like on an ordinary capital equipment market. Each builder tries to sell by relying on the assets it has in addition to the offered price and related services: industrial reputation and experience that play confusedly when untested advanced reactors are competing with one another, credit terms offered by the State and the government's influence on the market of emerging economies, the backing o the State's financial insurance in the event of risks taken in the sale of turnkey untested reactors. In the competition of the five manufacturers in the export market, American builders do not seem to have the best place, though even the leading position of Framatome ANP shows some limits. (author)

  3. The under-critical reactors physics for the hybrid systems; La physique des reacteurs sous-critiques des systemes hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Schapira, J P [Institut de Physique Nucleaire, IN2P3/CNRS 91 - Orsay (France); Vergnes, J [Electricite de France, EDF, Direction des Etudes et Recherches, 75 - Paris (France); Zaetta, A [CEA/Saclay, Direction des Reacteurs Nucleaires, DRN, 91 - Gif-sur-Yvette (France); and others

    1998-03-12

    This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

  4. Strategy for nuclear wastes incineration in hybrid reactors; Strategies pour l'incineration de dechets nucleaires dans des reacteurs hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Lelievre, F

    1998-12-11

    The transmutation of nuclear wastes in accelerator-driven nuclear reactorsoffers undeniable advantages. But before going into the detailed study of a particular project, we should (i) examine the possible applications of such systems and (ii) compare the different configurations, in order to guide technological decisions. We propose an approach, answering both concerns, based on the complete description of hybrid reactors. It is possible, with only the transmutation objective and a few technological constraints chosen a posteriori, to determine precisely the essential parameters of such reactors: number of reactors, beam current, size of the core, sub-criticality... The approach also clearly pinpoints the strategic decisions, for which the scientist or engineer is not competent. This global scheme is applied to three distinct nuclear cycles: incineration of solid fuel without recycling, incineration of liquid fuel without recycling and incineration of liquid fuel with on-line recycling; and for two spectra, either thermal or fast. We show that the radiotoxicity reduction with a solid fuel is significant only with a fast spectrum, but the incineration times range from 20 to 30 years. The liquid fuel is appropriate only with on-line recycling, at equilibrium. The gain on the radiotoxicity can be considerable and we describe a number of such systems. The potential of ADS for the transmutation of nuclear wastes is confirmed, but we should continue the description of specific systems obtained through this approach. (author)

  5. Presence de Carbone-13 dans les elements combustibles de type (U,Pu)O 2 irradies en reacteur rapide

    Science.gov (United States)

    Kryger, Bernard; Hagemann, Robert

    1982-06-01

    Du carbone-13 produit par la réaction de capture neutronique 168O + 10n → 136C + 42He se forme dans les combustibles de type oxyde irradiés en neutrons rapides. Cette réaction, dont le seuil d'énergie se situe à 2.35 MeV, conduit à la formation d'une quantité de carbone-13 qui peut varier notablement suivant le spectre neutronique du réacteur (entre 20 et 40 × 10 -6g 13C/g (U,Pu)O 2 pour une fluence de 2 × 10 23 n/cm 2). DES mesures effectuées sur le combustible et la gaine par spectrométrie de masse après irradiation montrent qu'une fraction égale ou supérieure à la moitié du carbone-13 produit dans l'oxyde peut être transférée dans la gaine. Un tel comportement nous fait considérer le carbone-13 comme un véritable marqueur du carbone plus généralement contenu dans l'oxyde et, à ce titre, la détection de cet isotope devrait contribuer à élucider tout particulièrement les mécanismes de carburation de la gaine par les combustibles (U,Pu)O 2 des réacteurs surgénérateurs.

  6. 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

  7. Capture and geological storage of CO{sub 2}. Innovation, industrial stakes and realizations; Captage et stockage geologique du CO{sub 2}. Innovation, enjeux industriels et realisations

    Energy Technology Data Exchange (ETDEWEB)

    Lavergne, R.; Podkanski, J.; Rohner, H.; Otter, N.; Swift, J.; Dance, T.; Vesseron, Ph.; Reich, J.P.; Reynen, B.; Wright, L.; Marliave, L. de; Stromberg, L.; Aimard, N.; Wendel, H.; Erdol, E.; Dino, R.; Renzenbrink, W.; Birat, J.P.; Czernichowski-Lauriol, I.; Christensen, N.P.; Le Thiez, P.; Paelinck, Ph.; David, M.; Pappalardo, M.; Moisan, F.; Marston, Ph.; Law, M.; Zakkour, P.; Singer, St.; Philippe, Th.; Philippe, Th

    2007-07-01

    The awareness of the international community and the convergence of scientific data about the global warming confirm the urgency of implementing greenhouse gases abatement technologies at the world scale. The growth of world energy demand will not allow to rapidly get rid of the use of fossil fuels which are the main sources of greenhouse gases. Therefore, the capture and disposal of CO{sub 2} is a promising way to conciliate the use of fossil fuels and the abatement of pollutants responsible for the global warming. The economical and industrial stakes of this technique are enormous. In front of the success of a first international colloquium on this topic held in Paris in 2005, the IFP, the BRGM and the Ademe have jointly organized a second colloquium in October 2007, in particular to present the first experience feedbacks of several pilot experiments all over the world. This document gathers the transparencies of 27 presentations given at this colloquium and dealing with: the 4. IPCC report on the stakes of CO{sub 2} capture and storage; the factor 4: how to organize the French economy transition from now to 2050; the technology perspectives, scenarios and strategies up to 2050; the European technological platform on 'zero-emission thermal plants'; the CO{sub 2} capture and storage road-map in the USA; research, development and implementation of CO{sub 2} capture and storage in Australia; the Canadian experience; ten years of CO{sub 2} capture and storage in Norway; the In Salah operations (Algeria); CO{sub 2} capture and storage: from vision to realisation; the oxi-combustion and storage pilot unit of Lacq (France); the Altmark gas field (Germany): analysis of CO{sub 2} capture and storage potentialities in the framework of a gas assisted recovery project; oil assisted recovery and CO{sub 2} related storage activities in Brazil: the Buracica and Miranga fields experience; carbon capture and storage, an option for coal power generation; steel-making industries and their CO{sub 2} capture and storage needs: the ULCOS program; CO{sub 2} capture technologies: road-maps and potential cost abatement; membranes: oxygen production and hydrogen separation; CO2GeoNet: integration of European research for the establishment of confidence in CO{sub 2} geologic storage; CO2SINK, CO{sub 2} geologic storage test at the European pilot site of Ketzin (Germany); storage in aquifers for European industrial projects: AQUA CO2; the US approach: US standards for the qualification of a CO{sub 2} storage in agreement with federal and state regulations; legal and regulatory aspects; societal acceptation; CO{sub 2} capture, geologic storage and carbon market; economic aspects of CO{sub 2} capture and storage; an experience of implementation of 'clean development mechanisms' in an industrial strategy; closing talk. (J.S.)

  8. Aspects of industrial production of solid electrolyte fuel cells (SOFC) by thermal spraying technology; Aspekte industrieller Fertigung von Festelektrolyt-Brennstoffzellen (SOFC) mittels thermischer Beschichtungsverfahren

    Energy Technology Data Exchange (ETDEWEB)

    Weckmann, Hannes

    2010-07-01

    The present thesis deals with measures to optimize the large-volume production of Solid Oxide Fuel Cells (SOFC) based on thermal spraying technology. Based on the well-established Vacuum Plasma Spraying (VPS) at DLR the potential of alternative thermal spraying techniques as well as alternative base materials was investigated in order to deposit SOFC-anode, electrolyte and insulating layers. Production costs, reproducibility and long-term stability of the production process as well as the fuel cell performance were major target criteria. Depending on the parameter set applied when using the cost efficient Atmospheric Plasma Spraying (APS) in combination with Nickel-Graphite as base material a significant improvement of gas permeability and electrical conductivity was achieved in comparison to the VPS sprayed reference anode. The power density of a fuel cell with an APS-Nickel-Graphite anode (184 mW/cm{sup 2}) was slightly better than the performance with a VPS reference anode (159 mW/cm{sup 2}). In comparison to the VPS process, ceramic electrolyte layers of fully stabilized Zirconia (YSZ) with significantly higher gas tightness could be demonstrated when high energy processes such as Low Pressure Plasma Spraying (LPPS). Thin-film Low Pressure Plasma Spraying (LPPS-Thin-film) and High Velocity Oxy Fuel Spraying (HVOF) were applied. The power density of a fuel cell equipped with an HVOF electrolyte was significantly improved to 234 mW/cm{sup 2} as compared to 187 mW/cm{sup 2} with the VPS sprayed reference cell. Further improvement of the power density was achieved with an LPPS-electrolyte (273 mW/cm{sup 2}). HVOF and VPS sprayed layers of pure Spinel in composite with metallic active braze (equivalent to the sealing between individual layers in the fuel cell stack) could exceed the demanded charge transfer resistance of >1 k{omega}cm{sup 2} at 800 C operating temperature only in few cases. When blended base powder of Spinel and Magnesia in combination with the VPS was applied significantly higher charge transfer resistance of up to 1,7 M{omega}cm{sup 2} could be demonstrated at layer thicknesses between 20 {mu}m and 117 {mu}m. In the further course of the thesis an online controller was developed for the VPS process based on insitu process diagnostic systems in order to reproducibly produce thermal sprayed functional layers. In pretests the process diagnostic system Accuraspray appeared to be cable of identifying changes in the process by detecting particle properties in the spray torch. In order to serve as a knowledge base for the controller the plasma spraying process was successfully modelled as correlation between machine parameters, spray torch- und layer properties by using artificial neural networks (ANN). The ability for the generalization as well as the accuracy of the model was improved further by measures such as active learning (additional experiments at poorly represented regions) and the combination of several models with different network architectures as an ensemble (mixture of experts). Based on the artificial neural network model, an online process controller was developed, which manipulates major process parameters in case of a deviation in order to get back to the initial state. The ability of the prototype controller could be demonstrated successfully in most cases. (orig.)

  9. Ascouf: an industrial software for the engineering of nuclear facilities in operation; Ascouf: un outil numerique industriel pour l`ingenierie nucleaire du parc en exploitation

    Energy Technology Data Exchange (ETDEWEB)

    Genette, P.; Martelet, B. [Electricite de France (EDF), 69 - Villeurbanne (France). Div. Mecanique des Structures du Septen; Debost-Eymart, I. [Electricite de France (EDF), 92 - Clamart (France). Dept. Mecanique et Modeles Numeriques

    1998-10-01

    The ASCOUF software has been developed par EDF to facilitate the quick analysis of defects contained in pipes or elbows of the primary loop. This preprocessing tool of Code Aster, the structural analysis finite element code of EDF, has been used to carry out, with an increase of productivity a series of numerical studies proving the mechanical strength of these components. Its validation, taking into account the feed-back from previous studies, leads us to rely on the results. ASCOUF has afterwards been extended to solve the problems of lack of thickness of pipes of the secondary loop. (authors)

  10. Modélisation de réacteurs et procédés de polymérisation de propylène industriels

    OpenAIRE

    Feng , Lian-Fang

    2006-01-01

    In this thesis, rigorous steady and dynamic models have been developed for an industrial propylene polymerization process. They have then been used to develop new processes based on the existing one. For that purpose, this thesis has: ? reviewed the state of the art of the olefin Ziegler-Natta catalysts; ? analyzed classical propylene polymerization processes and their characteristics; ? analyzed the characteristics of industrial slurry and gas phase propylene polymerization reactors using re...

  11. Les pouvoirs du danger. Zone indsutrielle de Fos-sur-Mer. Anthropologie politique des risques industriels et du conflit de l'incinérateur.

    OpenAIRE

    Girard , Tobias

    2012-01-01

    This thesis is interested in the political construction of environmental conflicts related to waste, industrial hazards and pollution management. The research is wondering about the political uses of danger, this is to say, about the devices the powerful, politicians, multinational companies, state representatives or NGOs leaders, used to fight against the threats, also produce and create them, make the threats growing bigger by lack of decision, or mask and exploit them. In order to find som...

  12. Cleansing of industrial sites: the Charbonnages de France example; Depollution des sites industriels: l'exemple de Charbonnages de France

    Energy Technology Data Exchange (ETDEWEB)

    Lagarde, R.; Guise, Y.; Gobillot, R. [Charbonnages de France, 92 - Rueil Malmaison (France); Bonin, H. [GRS Valtech, 78 - Carrieres sur Seine (France)

    2004-09-01

    Charbonnages de France, the French national coal board, has planned to stop its mining activities. Todays, its main goal concerns the remediation of its polluted mining and industrial sites. This article presents the cleansing techniques used at the Auby coking plant site for the removal of the polycyclic aromatic hydrocarbons (PAH) from the soil: protection of the aquifer, thermal desorption of the polluted earth, oxidation of the evaporated pollutants, valorization of the processed earth. (J.S.)

  13. Preservation of food by cold chains. Part 4. Brief history of the industrial food preservation; Conservering van voedingsmiddelen door koudeketens. Deel 4. Korte historie van de industriele voedingsmiddelenconservering

    Energy Technology Data Exchange (ETDEWEB)

    Van den Berg, C.; Berends, E.

    2011-10-01

    This is the fourth article in the series on cold chain food preservation, which describes the use of refrigeration in the food industry. Milestones were the building of cold chains for food distribution and new processes such as margarine and lager beer, and the scaling up of existing processes (e.g. for dairy, meat and fish products). By quick freezing, introduced by Clarence Birdseye in 1925, refrigeration can also be used for long term preservation of food products, as a powerful alternative for canning food. [Dutch] Dit vierde artikel in de reeks 'voedingsmiddelenconservering door koudeketens' completeert de beknopte historie van de voedingsmiddelenindustrie met een overzicht van het in gebruik nemen van de koelmachine. Mijlpalen zijn het ontstaan van koudeketens voor voedseldistributie en de nu mogelijk geworden nieuwe processen, de bereiding van pilsbier en margarine, de opschaling van bereidingsprocessen en het lange-afstandtransport van bederfelijke producten zoals zuivelproducten, vlees en vis. Door diepvriezen als methode aan te wenden, uitgevonden door Clarence Birdseye in 1925, kan koude ook worden gebruikt om voedingsmiddelen lang houdbaar te maken, een superieur alternatief voor inblikken van voedsel.

  14. Cartographie des risques industriels du dépot pétrolier à Hussein Dey, Alger (Algérie

    Directory of Open Access Journals (Sweden)

    FATIMA ZOHRA MOHAMED-CHÉRIF

    2015-12-01

    Full Text Available Hazard Mapping of Industrial Oil Storage in Hussein Dey, Algiers ( Algeria. The industrial hazardsare a danger to environment. It is the case of the fuel depot of Caroubier in the municpality of Hussein Dey in the outskirts of the capital Algiers. It is a potential danger in this urban area and a real challenge for the local authorities which do not have app ropriate tools for risk management. The municipality possesses single maps with colored items corresponding to the identification of spatial risks. The objective of this article is to map the risks generated by fuel depot. This mapping will help us localiz e the spatial vulnerability to suggest land planning measures.

  15. Capture and geological storage of CO{sub 2}. Innovation, industrial stakes and realizations; Captage et stockage geologique du CO{sub 2}. Innovation, enjeux industriels et realisations

    Energy Technology Data Exchange (ETDEWEB)

    Lavergne, R; Podkanski, J; Rohner, H; Otter, N; Swift, J; Dance, T; Vesseron, Ph; Reich, J P; Reynen, B; Wright, L; Marliave, L de; Stromberg, L; Aimard, N; Wendel, H; Erdol, E; Dino, R; Renzenbrink, W; Birat, J P; Czernichowski-Lauriol, I; Christensen, N P; Le Thiez, P; Paelinck, Ph; David, M; Pappalardo, M; Moisan, F; Marston, Ph; Law, M; Zakkour, P; Singer, St; Philippe, Th; Philippe, Th

    2007-07-01

    The awareness of the international community and the convergence of scientific data about the global warming confirm the urgency of implementing greenhouse gases abatement technologies at the world scale. The growth of world energy demand will not allow to rapidly get rid of the use of fossil fuels which are the main sources of greenhouse gases. Therefore, the capture and disposal of CO{sub 2} is a promising way to conciliate the use of fossil fuels and the abatement of pollutants responsible for the global warming. The economical and industrial stakes of this technique are enormous. In front of the success of a first international colloquium on this topic held in Paris in 2005, the IFP, the BRGM and the Ademe have jointly organized a second colloquium in October 2007, in particular to present the first experience feedbacks of several pilot experiments all over the world. This document gathers the transparencies of 27 presentations given at this colloquium and dealing with: the 4. IPCC report on the stakes of CO{sub 2} capture and storage; the factor 4: how to organize the French economy transition from now to 2050; the technology perspectives, scenarios and strategies up to 2050; the European technological platform on 'zero-emission thermal plants'; the CO{sub 2} capture and storage road-map in the USA; research, development and implementation of CO{sub 2} capture and storage in Australia; the Canadian experience; ten years of CO{sub 2} capture and storage in Norway; the In Salah operations (Algeria); CO{sub 2} capture and storage: from vision to realisation; the oxi-combustion and storage pilot unit of Lacq (France); the Altmark gas field (Germany): analysis of CO{sub 2} capture and storage potentialities in the framework of a gas assisted recovery project; oil assisted recovery and CO{sub 2} related storage activities in Brazil: the Buracica and Miranga fields experience; carbon capture and storage, an option for coal power generation; steel-making industries and their CO{sub 2} capture and storage needs: the ULCOS program; CO{sub 2} capture technologies: road-maps and potential cost abatement; membranes: oxygen production and hydrogen separation; CO2GeoNet: integration of European research for the establishment of confidence in CO{sub 2} geologic storage; CO2SINK, CO{sub 2} geologic storage test at the European pilot site of Ketzin (Germany); storage in aquifers for European industrial projects: AQUA CO2; the US approach: US standards for the qualification of a CO{sub 2} storage in agreement with federal and state regulations; legal and regulatory aspects; societal acceptation; CO{sub 2} capture, geologic storage and carbon market; economic aspects of CO{sub 2} capture and storage; an experience of implementation of 'clean development mechanisms' in an industrial strategy; closing talk. (J.S.)

  16. French and International experience on the dialogue around industrial sites; Experience francaise et internationales sur la concertation autour des sites industriels

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Th [Centre d' Etude sur l' Evaluation de la Protection dans le Domaine Nucleaire, CEPN, 92 - Fontenay aux Roses (France); Heriard Dubreuil, G; Gadbois, S [Mutadis, 94 - Vitry (France); Oudiz, A [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Remond Gouilloud, M [Paris-6 Univ. Sorbonne, 75 (France)

    2002-12-15

    This report presents the results of a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial installations'. It used the experience of the North Cotentin radioecology group where expertise has been implemented in order to evaluate the impact on health of the releases of the Cogema La Hague plant. This report is the fruit of an interdisciplinary group ( experts of activities with risks, radiation protection, regulation in environment). (N.C.)

  17. Exposure conditions, lung function and airway symptoms in industrial production of wood pellets. A pilot project; Exponeringsfoerhaallanden, lungfunktion och luftvaegsbesaer vid industriell produktion av traepellets. Ett pilotprojekt

    Energy Technology Data Exchange (ETDEWEB)

    Edman, Katja; Loefstedt, Haakan; Berg, Peter; Bryngelsson, I.L.; Fedeli, Cecilia; Selden, Anders [Oerebro Univ. Hospital (Sweden). Yrkes- och miljoemedicinska kliniken; Eriksson, Kaare [Umeaa Univ. Hospital (Sweden); Holmstroem, Mats; Rask- Andersen, Anna [Uppsala Univ. Hospital (Sweden)

    2002-02-01

    The production of wood pellets is a relatively new branch of the Swedish wood industry and has increased during the last years. A pilot study was performed to investigate the prevalence of airway symptoms, lung function and exposure among all 39 men employed in industrial production of wood pellets at six companies. The study included a questionnaire, medical examination, registration of nasal-PEF (peak expiratory flow) during a week, allergy screening (Phadiatop) and lung function (spirometry) before and after work shift. The results were compared with different reference data from other Swedish studies. Exposure measurements of monoterpenes and wood dust on filter and with a data logger (DataRAM) were also performed. The study group reported a higher frequency of cough without phlegm, awakening due to breathlessness and current asthma medication compared with reference data. For five of the six participants with physician-diagnosed asthma the disease debuted before the current employment and the results did not indicate an unusual asthma morbidity. Spirometry showed lower lung function before work shift than expected. However no difference over work shift was observed. A negative and non-significant correlation was seen between time with current work task and lung function. The study group reported a higher frequency of nasal symptoms mostly blockage, sneezing and dryness compared with reference data. The registrations of nasal-PEF did not show any differences between work and spare time. The prevalence of positive Phadiatop (23 %) did not differ from reference data. No association between exposure (wood dust and monoterpenes) and acute effects on lung function was observed. The wood dust exposure (0.16-19 mg/m{sup 3}) was high and 11 of 24 measurements exceeded the present Swedish occupational exposure limit of 2 mg/m{sup 3}. Peak exposures could be identified, e.g. at cleaning of engines with compressed air, with the DataRAM. The exposure to monoterpenes (0.64 and 24 mg/m{sup 3}) was low compared with the present Swedish limit of 150 mg/m{sup 3}. The monoterpene exposure does not seem to be a health or exposure problem in industrial production of wood pellets but wood dust exposure can effect the airways negatively. In this study the levels of wood dust were high and the study group reported more airway symptoms than expected. However the effect on lung function was small, but steps to reduce wood dust exposure should be done.

  18. Cooperative project for energy-oriented modernisation of industrial buildings - coordination results; Projektverbund zur energiegerechten Sanierung industriell errichteter Gebaeude - Ergebnisse der Koordination

    Energy Technology Data Exchange (ETDEWEB)

    Kerschberger, A. [Assmann Beraten und Planen GmbH, Stuttgart (Germany)

    1997-12-31

    The cooperative project concept of the federal ministry of science and technology is discussed. Coordination activities as well as pilot projects and comparative projects are described. Results obtained so far are reported in detail; heat energy conservation potentials as substantiated by means of measurements are shown in tabulated form. Further indicated in tabulated form are the redevelopment costs per residential unit, the payback period of the entire project, and a comparison of reduced heating costs with increased basic rent for apportioning the costs of modernization. (MSK) [Deutsch] Die Konzeption des Projektverbundes des BMBF wird naeher erlaeutert. Die Aktivitaeten der Koordination und die Versuchs-und Vergleichsbauvorhaben werden beschrieben. Die bisherigen Ergebnisse werden detailliert geschildert und die mess

  19. Energetic and ecological aspects of the modernisation of prefabricated large-panel buildings; Ergebnisse der Sanierung von industriell errichteten Gebaeuden der Blockbauart unter energetischen und oekologischen Aspekten

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, H. [Gebaeudewirtschaft Merseburg GmbH (Germany); Schramek, E.R.; Kaiser, M. [Gertec GmbH, Essen (Germany); Goerres, M. [Dortmund Univ. (Germany). Lehrstuhl Technische Gebaeudeausruestung

    1997-12-31

    In the new federal states, some four million residential units are still heated with individual stoves fired with briquetted brown coal. These stoves will be replaced by modern heating systems using natural gas or fuel oil or by district heat. Action taken within the framework of this research project is to permit making informative statements as to the amount of energy to be saved by modernization measures and the potential for cutting down carbon dioxide emissions. The data provided concern space heat energy consumption, end-use energy consumption, and primary energy consumption. Regarding the pollutant budget, information on climate-relevant and other pollutants is given. (MSK) [Deutsch] In den neuen Bundeslaendern werden noch ca. 4 Millionen Wohnungen mit Braunkolebriketteinzeloefen beheizt. Diese werden durch moderne mit Erdgas und Heizoel betriebene Heizungssysteme oder Fernwaerme ersetzt werden. Die in diesem Forschungsprojekt durchgefuehrten Massnahmen sollen aussagekraeftige Daten ueber die erreichbare Energieeinsparung durch Sanierungsmassnahmen und das Minderungspotential der Kohlendioxid-Emissionen liefern. Es sind Daten zum Heizwaermeverbrauch, zum Endenergieverbrauch, sowie zum Primaerenergieverbrauch enthalten. Fuer die Schadstoffbilanz werden Angaben zu klimarelevanten und zu sonstigen Schadstoffen gemacht. (MSK)

  20. A contribution from Gaz de France to the economic performance of industries; Contribution de Gaz de France a la perfomance economique des industriels

    Energy Technology Data Exchange (ETDEWEB)

    Depail, J.C. [Gaz de France (GDF), 75 - Paris (France)

    1996-12-31

    The aim of the policy of the French national gas utility, Gaz de France, towards industries, is to promote natural gas as a competitive fuel compared to fuels and electric power, with energy efficient solutions that are easy to implement and maintain: space heating, paint curing, surface cleaning, bath heating, vapour generation, waste treatment (especially for molding sand and volatile organic compounds, sludge drying). Gaz de France proposes also expertise schemes and audits

  1. Report from the commission about the industrial and financial project of EdF; Rapport de la commission sur le projet industriel et financier d'EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This report takes stock of the work carried out by the commission appointed by the French ministry of economy, finances and industry about the industrial and financial project of Electricite de France (EdF) in the framework of the liberalization of European energy markets. The report presents the conclusions of the commission about EdF's position in the new competition context, about the financial position of the group and about the foreseeable strategic options and their consequences in terms of equity fund needs. 5 appendixes present: the evolution of electricity prices, EdF and the energy policy, the electricity market and the competition in Europe, the EdF group: presentation and main adaptation stakes, the financial situation of EdF group. (J.S.)

  2. Assistance to the industrial process supervision: toward a methodology of conception; Aide a la supervision des processus industriels: vers une methodologie de conception

    Energy Technology Data Exchange (ETDEWEB)

    Benkhannouche, S

    1996-05-31

    This thesis presents a methodological approach to the design of computerized assistance for operators in control industrial processes. We are particularly interested in how to find the solutions which best suit their needs. Our preferred approach is focused on the operator: the main factors influencing his performance are reviewed and we make a synthesis which consists of a categorized list, or typology, of the extents of the operators` activities, tasks and errors. This typology is then used to classify the possible improvements as well as associated computer aids. The DIAPASON held system for fault diagnosis is integrated in this structure. This typology is our chosen basis for defining a specification method which enables the quality of the designed system to be guarantee. We propose a phased approach, the first phase of which involves analysing needs and thus identifying the objectives of the project. The second phase is the preparation of a performance specification which serves as a reference system for the project. In the third phase technical solutions are proposed to meet the requirements set out in the performance specification. The following phases involve studying the technical feasibility of the proposed solutions and the actual development of the system. Together with the feasibility study comes the step of making up a knowledge bank. The usual method of systems analysis are included in the typology of the aids. Furthermore, the SAGACE method uses a new approach to systems analysis based on its description which unites various points of view ; the evaluation of its possibilities forms a part of the construction of a reference system which gathers up the information needed to put the DIAPASON diagnosis system into action. (author).

  3. Bertrand Lavedrine, Jean-Paul Gandolfo, L’Autochrome Lumière. Secrets d’ateliers et défis industriels

    OpenAIRE

    Boulouch, Nathalie

    2009-01-01

    Bien que l’autochrome ait bénéficié de l’attention régulière des historiens et des éditeurs depuis une quinzaine d’années, il restait encore une part importante de connaissances à révéler à propos de l’histoire industrielle du premier procédé de photographie couleur commercialisé en juin 1907 par la Société Lumière et ses fils. C’est chose faite avec cet ouvrage qui s’appuie sur les recherches menées de longue date par Bertrand Lavedrine et Jean-Paul Gandolfo. Sa qualité réside dans le tressa...

  4. Methodological guide: management of industrial sites potentially contaminated by radioactive substances; Guide methodologique: gestion des sites industriels potentiellement contamines par des substances radioactives

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    At the request of the Ministries of Health and the Environment, IPSN is preparing and publishing the first version of the methodological guide devoted to managing industrial sites potentially contaminated by radioactive substances. This guide describes a procedure for defining and choosing strategies for rehabilitating such industrial sites. (author)

  5. Optimization of the workers radiation protection in the electro nuclear, industrial and medical fields; Optimisation de la radioprotection des travailleurs dans les domaines electronucleaire, industriel et medical

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    This conference is devoted to the radiation protection and the best way to optimize it. It reviews each area of the nuclear industry, and explores also the medical sector. Dosimetry, ALARA principle and new regulation are important points of this meeting. (N.C.)

  6. Management of ionizing radiation sources in university, medical and industrial environments; Gestion des sources ionisantes en milieux universitaire, medical et industriel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    This conference treats several subjects relative to the use of radioactive sources. The first session comprises three articles about ionizing sources and regulation. The second session, with three articles, tackles the question of radiation protection in the use of sources in industrial field. The third session, four articles, treats the same question but in the medicine and university media. The fourth session (three articles) is devoted to the organisation of radiation protection in the case of accidents. The fifth session concerns the management of spent sources (three articles). The sixth session studies the radiation protection of sources in Europe. The seventh and final session ends with the part and coordination of actors in radiation protection in the sources management (three articles). (N.C.)

  7. Quality of electricity service: Evaluation of nuisance index (IGI) of industrial customers; Qualite du service electrique:evaluation de l`indice de gene individuel des clients industriels

    Energy Technology Data Exchange (ETDEWEB)

    Naggar, R. [Hydro-Quebec, Montreal, PQ (Canada)

    1996-08-01

    The effects of power interruption on individual industrial customers by computing an individual nuisance index (IGI) is one of the tools planned by Hydro-Quebec to measure quality of service to its customers. When fully functioning, IGI will represent a combined value of loss of sales, overtime, lost materials and other direct costs, each IGI tailor-made for a particular company. Data for computing the index will be obtained from Hydro-Quebec`s own customer classification database, plus a commercial technical database (DTC) containing data required for the assessment of the nuisance, and a survey carried out by Hydro-Quebec involving some 1600 industrial customers. As of this date, the DTC is not yet available. A statistical analysis of survey responses was substituted to provide default values based on available parameters. Hydro-Quebec is confident that this new approach to evaluating service quality will open new horizons in quality assurance. 3 refs., 5 tabs.

  8. Valorisation des résidus carbonatés industriels pour le traitement de sulfure d'hydrogène dans les effluents gazeux

    OpenAIRE

    Galera Martinez , Marta

    2015-01-01

    The purpose of this study to valorize solid wastes from the Solvay process for the production of sodium carbonate as reagents for the treatment of H2S in air at concentrations typically found in wastewater treatment plants (tens to hundreds of ppmv of H2S). Firstly, the reactivity of two residues was evaluated in a gas-liquid-solid reactor at laboratory scale (250 ml). This reactor operates semi-continuously (continuous passage of gas through a fixed volume of slurry). The influence of operat...

  9. Biomass equipments. Dryers. Drying, crushing, agglomeration of agro-industrial products; Materiels pour la biomasse. Les secheurs, sechage, broyage, agglomeration de produits agro-industriels

    Energy Technology Data Exchange (ETDEWEB)

    Deur, O. [Promill (France)

    1997-12-31

    This paper describes the French Promill Company activity in the design and manufacturing of complete drying-crushing-agglomerating units for agro-industrial products (pulp of beet, lucerne, etc..). The paper focusses on the thermal and mechanical efficiency of the high temperature dryer and on the pulp granulating squeezer. (J.S.)

  10. Power frequency electric and magnetic fields: Questions and answers; Champs electriques et magnetiques produits par le courant industriel: Questions et reponses

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-10-01

    Power frequency fields are electric and magnetic fields produced by electric power. The first evidence to show that power frequency fields may have an effect on human health was noted in 1972 when it was reported that workers in high voltage switchyards suffered from a number of ailments. Fundamentals of electric and magnetic fields, and methods of their measurement, are explained. Some sources of power frequency fields are reviewed, and measured occupational exposures to these fields by electric utility workers are listed. Possible biological and health effects from power frequency fields are summarized from epidemiological, whole-animal, cellular, and observational studies reported in the literature. These include increased risk of cancer, decrease in melatonin, reproductive and developmental abnormalities, decreased heart rates, changes in calcium flow across cell membranes, and changes in neurotransmitters and corticosterone. Exposure standards for power frequency fields are given and methods for controlling exposure to those fields are noted. 53 refs., 1 fig., 2 tabs.

  11. Databases in the documentation management for big industrial projects; Les bases de donnees dans la gestion documentaire des grands projets industriels

    Energy Technology Data Exchange (ETDEWEB)

    Cauchet, A; Chevillard, F; Parisot, Y; Tirefort, C [Societe EURODOC, Saint Quentin en Yvenlines (France)

    1990-05-01

    The documentation management of a big industrial project involves a continuous update of information, both in the study and realization phase or in the operation phase. The organization of the technical documentation for big industrial projects requests complex information systems. In the first part of this paper are presented the methods appropriate for the analysis of documentation management procedures and in the second part are presented the tools by the combination of which a documentation system for the user is provided. The case of the documentation centres for the Hague reprocessing plant is described.

  12. Equipment for biomass. Dryers. Drying, crushing, aggregating of agro-industrial products; Materiels pour la biomasse, les secheurs, sechage, broyage, agglomeration de produits agro-industriels

    Energy Technology Data Exchange (ETDEWEB)

    Deur, O. [Promill, 28 - Serville (France)

    1997-12-31

    The French society Promill has developed complete units for the drying, crushing and aggregating of agro-industrial products (beet roots, agricultural wastes, lucerne, maize, etc.). Drying is conducted in a three-pass drum, using any type of fuel (fuel oil, gas, electric power, coal), and ensuring a thermal yield of 680 kCal/kg and ash emission rates complying with French and European legislation. Granulation is conducted with vapour addition, with a granulate flowrate reaching 15 T/h. Crushing is carried out in a hammer mill

  13. French and International experience on the dialogue around industrial sites; Experience francaise et internationales sur la concertation autour des sites industriels

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Th. [Centre d' Etude sur l' Evaluation de la Protection dans le Domaine Nucleaire, CEPN, 92 - Fontenay aux Roses (France); Heriard Dubreuil, G.; Gadbois, S. [Mutadis, 94 - Vitry (France); Oudiz, A. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Remond Gouilloud, M. [Paris-6 Univ. Sorbonne, 75 (France)

    2002-12-15

    This report presents the results of a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial installations'. It used the experience of the North Cotentin radioecology group where expertise has been implemented in order to evaluate the impact on health of the releases of the Cogema La Hague plant. This report is the fruit of an interdisciplinary group ( experts of activities with risks, radiation protection, regulation in environment). (N.C.)

  14. Analyse et modélisation de la précipitation de struvite : vers le traitement d'effluents aqueux industriels

    OpenAIRE

    Hanhoun, Mary

    2011-01-01

    La réduction des apports phosphorés des eaux usées régie par la Directive Européenne de 1991 (91/271/EEC) est considérée comme le facteur clé de la lutte contre la pollution des rivières et des lacs. Ces travaux concernent exclusivement l'étude de la formation maîtrisée de struvite (MgNH4PO4.6H2O) par précipitation comme alternative originale de récupération du phosphore et, par voie de conséquence, de l'ammonium à partir d'eaux usées. Un atout de ce procédé concerne la valorisation du précip...

  15. Methodical consultancy on optimized applications planning of industrial electrical process heat systems in competitive markets; Systematische Beratung zum optimierten Einsatz industrieller Elektroprozesswaermeanlagen im Wettbewerbsmarkt

    Energy Technology Data Exchange (ETDEWEB)

    Sonnenschein, P.

    1998-12-31

    In the competitive electricity market, industrial customers of electric utilities increasingly are in the position to demand from their power suppliers services tailored to their needs as well as excellent quality of products and services, and at competitive prices at that. The utilities therefore have to negotiate contracts with their customers for customization of services based on jointly performed analyses, with the relevant consequences for themselves in terms of utility and business management. Customer information and counselling on energy utilization and conservation has been crystallizing as an essential service demanded by the customers. The publication in hand is a tool of reference for utilities, presenting systematic guidance and strategies for achieving enhanced efficiency in processes such as customer care, soliciting of customers, and performance of services. The project examples given show that both customers and the utilities profit from the management approaches explained. It is recommended that in future, the marketing training of utility staff should be designed along the lines of education, training and on-the-job training schemes of sales engineers of other branches of industry in competitive markets. (orig./CB) [Deutsch] Die industriellen Kunden der Energieversorgungsunternehmen (EVU) konfrontieren ihre Lieferanten im neuen Wettbewerbsmarkt mit Forderungen nach individueller Betreuung sowie exzellenten Produkt- und Servicequalitaeten - und dies alles zu wettbewerbsfaehigen Preisen. Die Anforderungen an Energiedienstleistungen werden zunaechst aus Sicht der Kunden analysiert und die Konsequenzen fuer das Management von Geschaeftsbeziehungen dargelegt. Hierbei kristallisiert sich die Energieberatung als ein wesentliches Element der Energiedienstleistungen heraus. Mit der entwickelten Beratungssystematik sollen sowohl die Effektivitaet bei der Betreuung und der Akquisition der Kunden als auch die Effizienz bei der Erbringung der Dienstleistungen gesteigert werden. Die vorgestellten Projektbeispiele zeigen, dass sich hierbei Vorteile sowohl fuer die Kunden als auch fuer die EVU ergeben. Die Vertriebsausbildung der EVU-Mitarbeiter sollten sich an der Aus- und Weiterbildung fuer Vertriebsingenieure anderer kompetitiver Branchen orientieren. (orig.)

  16. Planning of decontamination and bleaching of textiles in an industrial cycle; Programmation des operations de decontamination et de blanchissage du linge dans un cycle industriel

    Energy Technology Data Exchange (ETDEWEB)

    Boutot, Pierre; Schipfer, Pierre [Commissariat a l' energie atomique et aux energies alternatives - CEA, Centre de Production de Plutonium de Marcoule, Service de Protection contre les Radiations (France)

    1964-10-15

    This note describes the operational planning for the decontamination and bleaching of textiles (clothes, protections, etc.) worn by personnel, in industrial-type washing machines. Various tests have been conducted with contaminated cotton samples using different cleaning products (and quantities) and various temperature cycles. The performance of the washing cycle (soaking, pre-washing, washing, rinsing) is discussed in terms of decontamination and washing efficiency, textile wear and resistance to shrinkage, whiteness, etc. The experimental washing machine is described [French] Cette etude programme les operations de decontamination et de blanchissage du linge au sein d'un cycle de traitement tel qu'il apparait dans les machines a laver industrielles a fort indice de production. Les echantillons de cotonnade, contamines au moyen de produits de fission, sont de meme nature que le tissu des vetements de protection. En matiere de decontamination les meilleurs resultats sont obtenus apres un trempage faiblement acide et un prelavage au moyen d'un sequestrant. Dans le cadre du blanchissage, seule une lessive industrielle employee dans la phase de lavage peut conferer aux tissus la luminance que requiert leur bonne presentation. Les taches persistantes sont effacees par blanchiment au cours du rincage tiede. Une analyse terminale permet de constater que l'usure des vetements est davantage liee aux conditions d'utilisation qu'aux operations de lavage et de decontamination. (auteurs)

  17. The economical accounting of the industrial wind power; Face au vent: si nous faisions le bilan economique de l'eolien industriel?

    Energy Technology Data Exchange (ETDEWEB)

    Poizat, F

    2006-11-15

    The aim of this document is the description of the implementing mechanism concerning the wind energy, of February 2000 for the principles and July 2006 for the investment program of the Government on the purchase obligation of the wind electricity by EDF. The laws and orders, the cost of the wind energy are detailed and discussed. (A.L.B.)

  18. Wobbe index control system in gas industry processes; Systeme de controle de l'index de Wobbe du gaz naturel dans les processus industriels

    Energy Technology Data Exchange (ETDEWEB)

    Cassibba, M.; Bertani, M. [SNAM, (Italy)

    2000-07-01

    Natural gas supplied to industry for process utilizations originates from different sources and that can cause fluctuations in gas composition. Changing gas composition may lead to production problems in industry with sensitive thermal processes (particularly glass industry and thermal metal treatments), such as efficiency and product quality. An equipment suitable to control and adjust such variations has been developed. Experimental tests in laboratory were carried out in order to investigate the control system accuracy and reliability. In particular five different settings were tested: at a preset thermal input by adjusting the natural gas flow rate in respect to Wobbe Index variations; at a set furnace temperature and stack oxygen level with variable thermal input by monitoring the Wobbe Index value; at constant Wobbe Index value by adding air to natural gas; at constant thermal input and prefixed Wobbe Index value by adding air to natural gas and varying the air and gas mixture flow rate; gross calorific value control by adding air or LPG to natural gas. All the tested settings gave good results. This report illustrates these results and the main features of the control system. The control and regulation system was installed in two glass factories for field tests. (authors)

  19. Pratiques enseignantes et réalité professionnelle : cas d’enseignants des lycées techniques et professionnels industriels au Gabon

    Directory of Open Access Journals (Sweden)

    EMMANUEL MOUDOUMA

    2011-07-01

    Full Text Available This article presents the outcomes of a study on the practises of Gabonese teachers who teach in the fields of industrial technical and vocational education. This study is interested in their daily activities and aims at characterising the inevitable variation which exists between the knowledge with work in the production activities industrial in company and those implemented in the institutions of vocational training. More generally, it is a question of appreciating the relation between training and employment. Thus, we are interested in the orientation given by the teacher to his daily practise to reduce or not this variation. The results of this study highlight the impact of epistemological and pragmatic dimensions on the organisation of the activities and thus on the practises. Obviously, the level of entry in the technical school course, the first job, the transport conditions, the appropriation of the reference frame of training and the development of the contents of courses are some of the items which impact this process of reduction of the difference between the two forms of knowledge.

  20. Définition et révision d'une stratégie de développement industriel

    OpenAIRE

    Choffray, Jean-Marie; Wagner, Philippe

    1983-01-01

    L'objet de cet article est de présenter une approche nouvelle de définition et de révision de la stratégie d'une entreprise, reposant sur l'utilisation de l'Analyse des Processus Hiérarchiques. Nous présentons le modèle permettant de mesurer les priorités à établir entre les différents objectifs et actions possibles à chaque niveau de la hiérarchie. Peer reviewed

  1. Héritage industriel et mémoire sensible : observations sur la constitution d'un "patrimoine sensoriel"

    OpenAIRE

    Simonnot, Nathalie; Siret, Daniel

    2014-01-01

    International audience; Through a variety of events and exhibitions, the concept of “sensory heritage” has appeared in several forms of expression of the cities’ industrial memory. It relies on the assumption that there would exist sensory aspects of a city related to its past or present activities, which would constitute its identity and therefore be subject to a particular conservation and diffusion. This paper will discuss the connection between these two apparently opposite notions. In a ...

  2. Innovation technologique et marketing industriel des entreprises productrices de circuits imprimés du district électronique des Marches

    OpenAIRE

    Maria Rosaria Marcone

    2005-01-01

    This paper deals with the first results of some research into the field of Business-to-Business, involving some processes of subcontracting implemented at an international level by SMEs operating in a high tech sector (PCBs). Particular attention has been placed on the prospects for development of the SME, which are located in the industrial district of Castelfidardo, and which operate in the global market. This research has the aim of studying the modifications that have effected the demand ...

  3. Strategic information for industrial policy-making in developing countries; Information strategique pour le policy-making industriel dans les pays en developpement

    Energy Technology Data Exchange (ETDEWEB)

    Gonod, P F

    1990-05-01

    The practice shows that many crucial decisions for industrialization in developing countries have been taken based on incomplete information. For strategic decisions an incomplete information may have catastrophic consequences. The function of policy-making is defined as the process by which the information generated/or used in a particular context is reevaluated in a different context in order to formulate/or execute a policy of alternative decisions. It follows that the industrial information must be presented in such a manner to allow a reevaluation and alternative decisions. 30 notes.

  4. An industrial perspective on environmental benefits of biotechnology on legal controls. Ein industrieller Ausblick: Nutzen der Biotechnologie fuer die Umwelt und gesetzliche Kontrollen

    Energy Technology Data Exchange (ETDEWEB)

    Brauer, D [Hoechst AG, Frankfurt am Main (Germany)

    1991-10-01

    Modern biotechnology was 'invented' in 1972/73. New technologies typically require a fifteen to twenty year incubation period of research and development before their commercial, industrial and other economic impacts begin to become significant. Biotechnology holds the key to many problems from environmental management to major savings for the consumer through improved performance of products and processes as well as from entirely new products and services. Simultaneously, environmental applications of biotechnology will produce enormous savings in the economic costs of pollution. Modern biotechnology, its current applications and its potential to reduce pressure on the environment, will be discussed in respect to biosafety, regulatory approaches, new processes, R and D, 'after processes' waste treatment and bioremediation. (orig.).

  5. Study of a method of detection for natural carbon-14 using a liquid scintillator, recent variations in the natural radio-activity due to artificial carbon-14 (1963); Etude d'une methode de detection du carrons 14 naturel, utilisant un scintillateur liquide - variations recentes de l'activite naturelle dues au carbone 14 artificiel (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-06-15

    carbone 14 rejete sous forme de {sup 14}CO{sub 2} par les reacteurs refroidis partiellement a l'air exterieur. (auteur)

  6. Study of a method of detection for natural carbon-14 using a liquid scintillator, recent variations in the natural radio-activity due to artificial carbon-14 (1963); Etude d'une methode de detection du carrons 14 naturel, utilisant un scintillateur liquide - variations recentes de l'activite naturelle dues au carbone 14 artificiel (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-06-15

    environs du site de Saclay, accroissement provoque par le carbone 14 rejete sous forme de {sup 14}CO{sub 2} par les reacteurs refroidis partiellement a l'air exterieur. (auteur)

  7. Research into zirconium alloys resistant to carbon dioxide under pressure at temperatures of up to 600 deg C (1963); Recherche d'alliages de zirconium compatibles avec le gaz carbonique sous pression jusqu'a 500 ou 600 deg C (1063)

    Energy Technology Data Exchange (ETDEWEB)

    Baque, P; Dominget, R; Bossard, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    Zirconium is a metal having a relatively low neutron capture cross-section and a high melting point; it is thus possible to consider its use in particular as a canning material for fuel elements in CO{sub 2}-cooled nuclear reactors. A preliminary study of several types of zirconium showed that the metal is already strongly oxidised in this gas at 500 deg C. The 'breakaway' phenomenon is generalised; the oxidation rate is then linear and depends on the carbon dioxide pressure. An attempt was therefore made to find binary and tertiary alloys in order to improve the metal behaviour. Several interesting compositions were found: 1, 1.6 and 2.5 per cent of copper, 2 per cent of vanadium, and 0.05 and 0.5 per cent of calcium. Tertiary copper-molybdenum and copper-phosphorus alloys are also less liable to oxidation and in particular do not exhibit the 'breakaway' phenomenon even after a prolonged treatment at 600 deg C. (authors) [French] Le zirconium se trouve parmi les metaux a section de capture neutronique relativement faible et possede une temperature de fusion elevee; aussi peut on songer a l'employer notamment comme materiau de gainage d'elements combustibles pour reacteurs nucleaires refroidis au gaz carbonique. Une etude prealable de plusieurs qualites de zirconium a montre que le metal est deja assez fortement oxyde dans ce gaz des 500 deg C. En effet, le phenomene de ''breakaway'' est general; la vitesse d'oxydation devient alors lineaire et depend de la pression du gaz carbonique. La recherche d'alliages binaires et ternaires a donc ete entreprise afin de tenter d'ameliorer le comportement du metal. Elle a permis d'aboutir a quelques compositions interessantes: cuivre 1, 1,6 et 2,5 pour cent, vanadium 2 pour cent, et calcium 0,05 et 0,5 pour cent. Des alliages ternaires au cuivre-molybdene et cuivre-phosphore sont egalement moins oxydables, et en particulier ne presentent pas le phenomene de ''breakaway'', meme apres une longue exposition a 600 deg C. (auteurs)

  8. The Grand Quevilly thermal test station - the SMW sodium circuit with a generator of superheated steam at 545 deg; Station d'essais thermiques de grand quevilly - circuit de sodium de 5 MW avec generateur de vapeur surchauffee a 545 deg

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    A 5 MW installation is described which is a reduced model of the heat exchange system of a sodium-cooled reactor. This plant, which is situated at Grand Quevilly (near Rouen), consists of: 1 - A primary sodium loop made up of a sodium re-heater running on heavy diesel oil, a mechanical pump and an intermediate exchanger made up of clusters of tubes fitted with baffles. 2 - A NaK(56 per cent of K) secondary loop consisting mainly of a mechanical pump and a double-wall steam generator with forced circulation and complete vaporization. 3 - A tertiary water loop consisting of the inside of the steam generator pipes, a pressure-reducing valve which cools down the super-heated fluid and acts as a turbine, a condenser, a charge-pump and a supply pump for the boiler. The heat is given finally to a water-source flowing into the Seine. Two important points of the installation are: - The water treatment unit - The control and regulation system Apart from the general satisfactory operation of the installation which it is hoped to obtain, a careful study will be made of the heat transmission coefficients of the important equipment such as the intermediate exchanger and the steam generator. The construction was finished on April 28, 1964. (author) [French] On decrit une installation de 5 MW figurant a echelle reduite un systeme de transfert de chaleur d'un reacteur refroidi au sodium. Cette installation, situee a Grand Quevilly (pres de Rouen) comprend: 1 - Une boucle de sodium primaire comportant un rechauffeur de sodium alimente en fuel lourd, une pompe mecanique et un echangeur intermediaire a faisceau tubulaire muni de chicanes, 2 - Une boucle de NaK (56% de K) secondaire dont les appareils principaux sont une pompe mecanique et un generateur de vapeur a double paroi, circulation forcee et vaporisation totale. 3 - une boucle tertiaire a eau comprenant l'interieur des tubes du generateur de vapeur, un detendeur-desurchauffeur simulant une turbine, un condenseur, une pompe de

  9. Du Pont de Nemours

    NARCIS (Netherlands)

    Ros JPM; LAE

    1994-01-01

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

  10. Produktie van pigmenten

    NARCIS (Netherlands)

    Etman EJ; Duesmann HB; Eijssen PHM; LAE

    1994-01-01

    Dit rapport over de produktie van pigmenten is gepubliceerd binnen het Samenwerkingsproject Procesbeschrijvingen Industrie Nederland (SPIN). In het kader van dit project is informatie verzameld over industriele bedrijven of industriele processen ter ondersteuning van het overheidsbeleid op het

  11. Farmaceutische industrie

    NARCIS (Netherlands)

    Ros JPM; van der Poel P; Etman EJ; Montfoort JA; LAE

    1995-01-01

    Dit rapport over de farmaceutische industrie is gepubliceerd binnen het Samenwerkingsproject Procesbeschrijvingen Industrie Nederland (SPIN). In het kader van dit project is informatie verzameld over industriele bedrijven of industriele processen ter ondersteuning van het overheidsbeleid op het

  12. Characteristics and construction problems of EL 4; Caracteristiques et problemes de construction d'EL4

    Energy Technology Data Exchange (ETDEWEB)

    Carle, R; Schulhof, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Sevin, Ph [Electricite de France (EDF), 75 - Paris (France); Buttin, J [Societe INDATOM (France)

    1964-07-01

    detail and the connections to be employed were tested. The whole circuit is made in fairly classical materials (slightly alloyed steels) whose behaviour in the carbon dioxyde at 500 deg. C was proved. The CO{sub 2}, and heavy water circuits construction will begin in October 1964. Aerodynamic tests were carried out for the helico-centrifugal blowers (of unit power 9 MW). The choice of a pressure as high as 60 kg/cm{sup 2} does not seem to induce new problems in connection with leaks on the machine shafts. Finally, the choice of a type of CO{sub 2} - steam heat exchanger with forced circulation led Electricite de France to test the operation and stability of a prototype exchanger in its test plant The reactor will be equipped with a de-pressurizing and de-superheating system which will allow the reactor to operate at 20 p. 100 of its nominal power whether the turbo-alternator is available or not. (authors) [French] EL 4 est le prototype d'une filiere originale de reacteurs moderes a l'eau lourde et refroidis au gaz carbonique. Son etude a ete menee dans la double optique de: - realiser un reacteur suffisamment important et complet pour y tester l'ensemble des problemes de construction et d'exploitation de la filiere; - menager dans l'installation les possibilites de tenir compte des ameliorations (materiaux nouveaux, elements combustibles ameliores) qui sont etudiees par ailleurs. Le premier objectif n'etait envisageable que sous reserve d'un volume d'etudes preliminaires important. A ce titre, ont ete realises et essayes de 1962 a 1964 plusieurs canaux prototypes, hors pile, mais dans les conditions reelles de temperature et de pression. Ces essais ont montre la bonne tenue des materiaux aux difficiles conditions mecaniques et chimiques du projet. Ces installations seront d'ailleurs disponibles pour eprouver, avant mise en pile, les modifications ulterieures. D'importants essais touchant la securite du reacteur en cas d'explosion du circuit de CO{sub 2}, ont ete realises

  13. 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

  14. 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

  15. On the phenomenon of the reversal of the cooling current in the hot pipes of a swimming-pool type pile cooled by forced convection; Sur un phenomene de renversement du courant de refrigeration dans les canaux chauds d'une pile piscine refroidie en convection forcee

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    It is shown, for a swimming-pool type pile cooled by forced convection (general flow downwards), that a permanently stable regime with downward flow in all the channels is not possible when the flow is below a critical value for a given power. In the hot channels the natural convection then becomes preponderant, the direction of the flow is reversed and a permanently stable regime exists for which the flow is upwards in the hot channels. Calculations are made, with simplifying hypotheses in the case of Melusine. (author) [French] Pour une pile piscine refrigeree en convection forcee (ecoulement global descendant), on montre qu'un regime permanent stable avec ecoulement descendant dans tous les canaux est impossible lorsque le debit est inferieur a une valeur critique pour une puissance donnee. Dans les canaux chauds, la convection naturelle l'emporte alors, le sens du courant s'inverse et un regime permanent stable existe, pour lequel le courant est ascendant dans les canaux chauds. On fait les calculs, avec des hypotheses simplificatrices, dans le cas de Melusine. (auteur)

  16. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  17. Very high temperature measurements: Applications to nuclear reactor safety tests; Mesures des tres hautes temperatures: Applications a des essais de surete des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Parga, Clemente-Jose

    2013-09-27

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  18. Fe Al40, a new canning material for reactors using refractory fuels; Le Fe Al40, un nouveau materiau de gainage pour les reacteurs a combustibles refractaires

    Energy Technology Data Exchange (ETDEWEB)

    Sainfort, G; Cabane, G; Salesse, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Fe Al40, owing to its high aluminium content, is more suitable than stainless steels for nuclear applications; it has two advantages: its nuclear cross section is half that of 18-10 stainless steels, and its compatibility with fuel elements and heat extracting fluids is exceptionally good. Ferrous alloys with more than 16 per cent by weight of aluminium are reputed to be brittle because of their ordered lattice. But actually, most of the brittleness of these alloys is due to the presence of intergranular precipitates. The vacuum casting of pure iron and aluminium, together with additions of scavenging elements, gives a very clean alloy with sufficiently reduced brittleness at high temperatures as to allow transformation with a very good yield. Studies of smelting and transformation have enabled the optimum composition and the best industrial fabrication conditions to be established. The mechanical properties of extruded or rolled products are dependent, on the ferritic ordered structure of the alloys prepared as follows: Extension at room temperature between 8 and 11 p.100, continuous increase of the elongation at rupture with increase of temperature so that at 800 C it exceeds 100 p.100; yield strength stable at 30 kg/mm{sup 2} from 20 to 550 C; progressive decrease of impact strength as the temperature increases Creep strength at 650-700 C in the region of 10 kg/mm{sup 2} with a very high elongation because of the appearance of a fine grain recrystallization during the test. Welding of Fe Al40 is facilitated by the fact that this alloy remains ferritic up to its melting point (1350 C); however it. is sensitive to hot short cracking and grain swelling. Several welding processes have been successfully applied to this alloy, under conditions where the superficial alumina layer was removed. This very impervious oxide skin gives Fe Al 40 a unique resistance to corrosion: in carbon dioxide at 700 C, 60 atm, the weight gain after 3000 h is about one third of that of a 18-10 niobium - stabilized stainless steel, in water, either de oxygenated or saturated with oxygen at 25 C, the weight gain is one hundred time smaller than that of mild steel, after a 3 month test, in water vapour at 500 C, also after 3 months, only the growth of impervious and very adherent oxide skins is observed, in sodium up to 1000 h at 700 C, the behaviour is at least as good as that of stainless steel. No diffusion reaction has been detected either with uranium dioxide up to 800 C, or with uranium carbide up to 700 C. Extension specimens and thin walled cans are subjected to high flux irradiations between 20 and 700 C; preliminary results will be given. (authors) [French] Le Fe Al40, grace a sa haute teneur en aluminium, presente de grands avantages sur les aciers inoxydables, pour les applications nucleaires; en particulier, sa section efficace est moitie de celle de l'acier a 18 p.100 Cr et 10 p.100 Ni, et sa compatibilite avec les elements combustibles et avec les fluides caloporteurs est exceptionnellement bonne. Les alliages ferreux, contenant plus de 16 p.100 en poids d'aluminium, sont reputes fragiles en raison de leur structure ordonnee. En fait, la plus grande partie de la fragilite de ces alliages est due a la presence de precipites intergranulaires. L'emploi de fer et d'aluminium purs, ainsi que des additions destinees a pieger les traces d'impuretes residuelles, permet d'obtenir, par coulee sous vide, un alliage tres propre dont la fragilite a chaud est suffisamment reduite pour permettre une transformation. avec un excellent rendement. Les etudes de fonderie et de transformation, qui ont defini la composition de l'alliage et les meilleures conditions industrielles de preparation, seront decrites. Les proprietes mecaniques des produits files ou lamines sont conditionnees par la structure ferritique ordonnee des alliages ainsi prepares: allongement par traction a temperature ambiante compris entre 8 et 11 p.100, augmentation progressive de l'allongement de rupture quand la temperature croit de sorte qu'a 800 C il depasse 100 p.100, limite elastique de 30 kg/mm{sup 2} de 20 jusqu'a 550 C, baisse progressive de la resilience quand la temperature croit, resistance au fluage a 650-700 C de l'ordre de 10 kg/mm{sup 2} avec un allongement tres important par suite de l'apparition d'une recristallisation, en grains tres fins en cours d'essai. La soudure du Fe Al 40 est facilitee par le fait que cet alliage reste ferritique jusqu'a son point de fusion (1350 C); toutefois, il est assez sensible aux contraintes thermiques et au grossissement du grain. Plusieurs procedes de soudure ont pu etre appliques, avec succes, a cet alliage, dans les conditions ou la couche d'alumine superficielle a pu etre enlevee. Cette couche d'oxyde, tres impermeable, confere au Fe Al40 une remarquable resistance a la corrosion: dans le gaz carbonique, a 700 C sous 60 atmospheres, gain de poids au bout de 3000 h environ, 1/3 de celui d'un acier inoxydable 18-10 stabilise au niobium; dans l'eau a 25 C, desoxygenee ou saturee en oxygene gain de poids 100 fois plus faible que celui d'un acier doux au bout de 3 mois; dans la vapeur d'eau a 500 C, en 3 mois aussi, formation de couches d'oxyde tres adherentes et impermeables; dans le sodium jusqu'a 1000 h a 700 C, comportement au moins aussi bon que celui de l'acier inoxydable. Aucune reaction de diffusion n'a ete decelee ni avec l'oxyde d'uranium jusqu'a 800 C, ni avec le carbure d'uranium jusqu'a 700 C. Des eprouvettes et des gaines a parois minces sont soumises a l'irradiation dans des flux neutroniques intenses entre 20 et 700 C; les resultats preliminaires seront communiques. (auteurs)

  19. Contribution to the study of the temperature reactivity coefficient for light water reactors; Contribution a l`etude du coefficient de temperature des reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Mounier, C.

    1994-05-01

    In this work, we looked for the error sources in the calculation of the isothermal temperature coefficient for light water lattices. We studied three fields implied: the nuclear data, the calculation methods and the temperature coefficient measurement. About the measurement, we pointed out the difficulties of he interpretation. So we used an indirect approach by the mean of critical states at various temperatures. In that way, we can say that if the errors in the effective multiplication factor are constants with temperature then the temperature coefficient is correctly calculated. We studied the neutronic influence of light water models which are used in the thermal scattering cross-section computation. This cross-section determines the thermalization process of neutrons. We showed that the actual model (JEF2) is satisfactory of the needs of the reactors physics. Concerning the majors isotopes ({sup 235}U, {sup 238}U, {sup 239}Pu), the uncertainties on the nuclear data do not seem as a preponderant cause of errors, without to be totally negligible. We also studied, with the neutron transport code Apollo-2, the influence of difference approximations for cell calculation . The new possibilities of the code has been used to represent the critical experiments, particularly the improvement of the resonance self-shielding formalism. The calculation scheme adopted permits to remove partially the fundamental mode approximation by the mean of a two-dimensional transport calculation with the SN method, the axial leakage being treated as an absorption in DB{sup 2}{sub Z}. The agreement between theory and experiment is good both for the reactivity and the temperature coefficient. (author). 114 refs., 40 figs., 163 tabs., 1 append.

  20. Behaviour of BF{sub 3} counters after in-pile irradiation; Comportement de compteurs a BF{sub 3} apres irradiation dans un reacteur

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-07-01

    The behaviour of a proportional BF{sub 3}LCT 14 NE 31 counter with aluminium cathode has been studied after irradiation from 10{sup 5} to 10{sup 6} n.cm{sup -2} in the swimming-pool reactor Triton. The pulse spectrum was only slightly modified by two successive irradiations at 10{sup 15} n.cm{sup -2}, and 1 hour after the end of the irradiation the decrease in sensitivity reached 10 to 20 per cent. One hour after irradiation at 10{sup 16} n.cm{sup -2} this decrease was 20 per cent but the pulse spectrum was affected. The advantages of an aluminium cathode with respect to a standard counter using a copper cathode and irradiated under the same conditions are given. Beforehand, it had been established that the limit for use of an aluminium-cathode counter in the presence of {gamma} radiation is about 200 R.h{sup -1}. (author) [French] Le comportement d'un compteur proportionnel a BF-3LCT 14 NE 31, a cathode d'aluminium, a ete etudie apres des irradiations de 10{sup 5} a 10{sup 6} n.cm{sup -2} dans la pile-piscine TRITON. Deux irradiations successives de 10{sup 15} n.cm{sup -2} ont peu modifie le spectre des impulsions et 1 heure apres la fin de l'exposition la perte de sensibilite etait de 10 a 20 pour cent. Une heure apres une irradiation de 10{sup 16} n.cm{sup -2}, cette perte etait de 20 pour cent mais le spectre d'impulsions etait perturbe. L'avantage d'une cathode en aluminium est mis en evidence par comparaison avec un compteur standard a cathode de cuivre irradie dans les memes conditions. En preliminaire, on avait etabli que la limite d'utilisation d'un compteur a cathode d'aluminium en presence de rayonnements {gamma} etait de l'ordre de 200 R.h{sup -1}. (auteur)

  1. Fuel and target programs for the transmutation at Phenix and other reactors; Programmes combustibles et cibles pour la transmutation dans Phenix et autres reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard-Groleas, G

    2002-07-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  2. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  3. Contributions to safety studies for new concepts of nuclear reactors; Contributions aux etudes de surete pour des filieres innovantes de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Perdu, F

    2003-12-01

    The complete study of molten salt reactors, designed for a massive and durable nuclear energy production, must include neutronics, hydraulics and thermal effects. This coupled study, using the MCNP and Trio{sub U} codes, is undertaken in the case of the MSRE (molten salt reactor experiment) prototype. The obtained results fit very well the experiment. Their extrapolation suggests ways of improving the safety coefficients of power molten salt reactors. A second part is devoted to accelerator driven subcritical reactors, developed to incinerate radioactive waste.We propose a method to measure the prompt reactivity from the decay following a neutron pulse. It relies only on the distribution of times between generations, which is a characteristic of the reactor. This method is implemented on the results of the MUSE 4 experiment, and the obtained reactivity is accurate within 5%. (author)

  4. Synthesis of a catalytic reactor membrane for synthesis gas production; Elaboration d'une membrane de reacteur catalytique pour la production de gaz de synthese

    Energy Technology Data Exchange (ETDEWEB)

    Juste, E.; Julian, A.; Chartier, T. [Limoges Univ., Lab. Science des Procedes Ceramiques et de Traitements de Surface (SPCTS, UMR 6638 CNRS), 87 (France); Juste, E.; Julian, A.; Del Gallo, P.; Richet, N. [Centre de Recherche Claude-Delorme, Air Liquide, 78 - Jouy en Josas (France)

    2007-07-01

    The conversion of natural gas to synthesis gas (mixture of H{sub 2} and CO) is a main challenge for the hydrogen and clean fuels production. Mixed (ionic O{sup 2-} and electronic) conducing ceramics membrane reactors seem particularly promising. The design considered for the membrane is a tri-layer system integrating a reforming catalyst and a dense membrane laying on a porous support. Among the materials considered for the dense membrane, perovskites La{sub 1-x}Sr{sub x}Fe{sub 1-y}Ga{sub y}O{sub 3-{delta}} seem to be interesting for their performances and stability. The oxygen flux through the membrane is measured in terms of temperature under different oxygen partial pressure gradients. In the industrial experimental conditions, the membrane is submitted to a strong oxygen (air/methane) partial pressure gradient of about 900 C which induces mechanical stresses, on account of the material expansion difference, in terms of p{sub O2}. In this framework, the evolutions of the performances and of the expansion coefficient have been followed in terms of the substitutions rates in La{sub (1-x)}Sr{sub x}Fe{sub (1-y)}Ga{sub y}O{sub 3-{delta}} with x{<=}0.5 and y{<=}0.5. (O.M.)

  5. Measurement of neutrinos released in nuclear reactors through the Borexino experiment; Mesure des neutrinos de reacteurs nucleaires dans l'experience Borexino

    Energy Technology Data Exchange (ETDEWEB)

    Dadoun, O

    2003-06-01

    The main goal of the Borexino experiment is to measure in real time the solar neutrino flux from the beryllium (Be{sup 7}) line at 862 keV. Beyond this pioneer low energy neutrino detection, Borexino will be able to measure solar neutrinos above the MeV, (B{sup 8} neutrinos and pep neutrinos), nuclear reactor neutrinos (with an average energy of 3 MeV) and the supernova neutrinos (their spectrum goes up to some ten MeV). In this work I mainly focus on the study of the nuclear reactors neutrinos. This field has recently been enriched by the results of the KamLAND experiment, which have greatly improved the determination of the neutrino oscillation parameters. In order to measure these events which are above the MeV, the Borexino collaboration entrusted the PCC group at College de France, with the tasks of developing a fast digit system running at 400 MHz: the FADC cards. The PCC group designed the FADC cards and completed them at the beginning of 2002. The first cards which were introduced in the main electronic acquisition unit allowed us to control their functioning and that of the acquisition software. FADC cards were also installed in the Borexino prototype, CTF. The data are analysed in order to determine a limit to the expected background noise of Borexino in measuring the nuclear reactor neutrinos. (author)

  6. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  7. Civacuve analysis software for mis machine examination of pressurized water reactor vessels; Civacuve logiciel d'analyse des controles mis des cuves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, Ph.; Gagnor, A. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    The product software CIVACUVE is used by INTERCONTROLE for the analysis of UT examinations, for detection, performed by the In-Service Inspection Machine (MIS) of the vessels of nuclear power plants. This software is based on an adaptation of an algorithm of SEGMENTATION (CEA CEREM), which is applied prior to any analysis. It is equipped with tools adapted to industrial use. It allows to: - perform image analysis thanks to advanced graphic tools (Zooms, True Bscan, 'contour' selection...), - backup of all data in a database (complete and transparent backup of all informations used and obtained during the different analysis operations), - connect PC to the Database (export of Reports and even of segmented points), - issue Examination Reports, Operating Condition Sheets, Sizing curves... - and last, perform a graphic and numerical comparison between different inspections of the same vessel. Used in Belgium and France on different kind of reactor vessels, CIVACUVE has allowed to show that the principle of SEGMENTATION can be adapted to detection exams. The use of CIVACUVE generates a important time gain as well as the betterment of quality in analysis. Wide data opening toward PC's allows a real flexibility with regard to client's requirements and preoccupations.

  8. Accident at the zero power reactor which happened on October 15 1958; Sur l'accident avec le reacteur de puissance zero du 15 octobre 1958

    Energy Technology Data Exchange (ETDEWEB)

    Savic, P [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine.

  9. Fast ultrasonic visualisation under sodium. Application to the fast neutron reactors; Visualisation ultrasonore rapide sous sodium. application aux reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Imbert, Ch

    1997-05-30

    The fast ultrasonic visualization under sodium is in the programme of research and development on the inspection inside the fast neutron reactors. This work is about the development of a such system of fast ultrasonic imaging under sodium, in order to improve the existing visualization systems. This system is based on the principle of orthogonal imaging, it uses two linear antennas with an important dephasing having 128 piezo-composite elements of central frequency equal to 1.6 MHz. (N.C.)

  10. Turbulent precipitation of uranium oxalate in a vortex reactor - experimental study and modelling; Precipitation turbulente d'oxalate d'uranium en reacteur vortex - etude experimentale et modelisation

    Energy Technology Data Exchange (ETDEWEB)

    Sommer de Gelicourt, Y

    2004-03-15

    Industrial oxalic precipitation processed in an un-baffled magnetically stirred tank, the Vortex Reactor, has been studied with uranium simulating plutonium. Modelling precipitation requires a mixing model for the continuous liquid phase and the solution of population balance for the dispersed solid phase. Being chemical reaction influenced by the degree of mixing at molecular scale, that commercial CFD code does not resolve, a sub-grid scale model has been introduced: the finite mode probability density functions, and coupled with a model for the liquid energy spectrum. Evolution of the dispersed phase has been resolved by the quadrature method of moments, first used here with experimental nucleation and growth kinetics, and an aggregation kernel based on local shear rate. The promising abilities of this local approach, without any fitting constant, are strengthened by the similarity between experimental results and simulations. (author)

  11. Systemic model for the aid for operating of the reactor Siloe; Modelisation systeme pour l`aide a l`exploitation du reacteur de recherche Siloe

    Energy Technology Data Exchange (ETDEWEB)

    Royer, J.C.; Moulin, V.; Monge, F. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires; Baradel, C. [ITMI APTOR, 38 - Meylan (France)

    1995-12-31

    The Service of the Reactor Siloe (CEA/DRN/DRE/SRS), fully aware of the abilities and knowledge of his teams in the field of research reactor operating, has undertaken a project of knowledge engineering in this domain. The following aims have been defined: knowledge capitalization for the installation in order to insure its perenniality and valorization, elaboration of a project for the aid of the reactor operators. This article deals with the different actions by the SRS to reach the aims: realization of a technical model for the operation of the Siloe reactor, development of a knowledge-based system for the aid for operating. These actions based on a knowledge engineering methodology, SAGACE, and using industrial tools will lead to an amelioration of the security and the operating of the Siloe reactor. (authors). 13 refs., 7 figs.

  12. Simulation des fuites neutroniques a l'aide d'un modele B1 heterogene pour des reacteurs a neutrons rapides et a eau legere

    Science.gov (United States)

    Faure, Bastien

    The neutronic calculation of a reactor's core is usually done in two steps. After solving the neutron transport equation over an elementary domain of the core, a set of parameters, namely macroscopic cross sections and potentially diffusion coefficients, are defined in order to perform a full core calculation. In the first step, the cell or assembly is calculated using the "fundamental mode theory", the pattern being inserted in an infinite lattice of periodic structures. This simple representation allows a precise modeling for the geometry and the energy variable and can be treated within transport theory with minimalist approximations. However, it supposes that the reactor's core can be treated as a periodic lattice of elementary domains, which is already a big hypothesis, and cannot, at first sight, take into account neutron leakage between two different zones and out of the core. The leakage models propose to correct the transport equation with an additional leakage term in order to represent this phenomenon. For historical reasons, numerical methods for solving the transport equation being limited by computer's features (processor speeds and memory sizes), the leakage term is, in most cases, modeled by a homogeneous and isotropic probability within a "homogeneous leakage model". Driven by technological innovation in the computer science field, "heterogeneous leakage models" have been developed and implemented in several neutron transport calculation codes. This work focuses on a study of some of those models, including the TIBERE model from the DRAGON-3 code developed at Ecole Polytechnique de Montreal, as well as the heterogeneous model from the APOLLO-3 code developed at Commissariat a l'Energie Atomique et aux energies alternatives. The research based on sodium cooled fast reactors and light water reactors has allowed us to demonstrate the interest of those models compared to a homogeneous leakage model. In particular, it has been shown that a heterogeneous model has a significant impact on the calculation of the out of core leakage rate that permits a better estimation of the transport equation eigenvalue Keff . The neutron streaming between two zones of different compositions was also proven to be better calculated.

  13. Relative measurement of the fluxes of thermal, resonant and rapid neutrons in reactor G1; Mesures relatives des flux thermique, resonnant et rapide dans le reacteur G1

    Energy Technology Data Exchange (ETDEWEB)

    Carle, R.; Mazancourt, T. de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    We sought to determine the behavior of the thermal, resonant and rapid neutron fluxes in the multiplier-reflector transition region, in the two principal directions of the system. We have also measured the variation of these different fluxes in the body of the multiplier medium in a canal filled with graphite and in an empty canal. The results are given in the form of curves representing: - the variation of the ratio of the thermal flux to the rapid flux in axial and radial transitions - the behavior of the thermal and resonant fluxes and the variation of their ratio in the same regions. (author) [French] Nous avons cherche a determiner le comportement des differents flux, thermique, resonnant et rapide a la transition milieu multiplicateur-reflecteur dans les deux directions principales du reseau. Nous avons egalement mesure la variation de ces differents flux au sein du milieu multiplicateur dans un canal rempli de graphite et dans un canal vide. Les resultats sont donnes sous forme de courbe representant: - La variation du rapport du flux thermique au flux rapide aux transitions axiale et radiale - L'allure des flux thermique et resonnant et la variation de leur rapport dans les memes regions. (auteur)

  14. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  15. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry; Experience ``Benchmark beton`` pour la dosimetrie hors cuve dans les reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, H.; D`Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

  16. Physicochemical state of the spent fuel leaving the reactors; Le combustible nucleaire et son etat physico-chimique a la sortie des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Dehaut, Ph

    2000-07-01

    This report focuses on the current knowledge, updated at the end of 1999, about the physicochemical state of the fuels leaving light water reactors, and particularly pressurized water reactors. Lessons are withdrawn from it making it possible to determine the points which require a necessary deepening of the data and coherence of interpretations. Lastly, evolution of the sailed fuel rod as well as the potential availability of gases and volatile fission products, during a secular storage or of a multi-millennium disposal, are the subject of an attempt at forecast. Accessible data in the scientific literature, or those acquired at the CEA, are particularly numerous. Their analysis and their synthesis are joined together to constitute a collection of references intended to the specialists in nuclear fuel and for all those which contribute to the reflexion on the storage or final disposal of the irradiated fuel. This memory is structured in ten chapters. The last chapter makes it possible to retain on some pages, the essential lessons of this study. Chapter I: Introduction; Chapter II: Characteristics of assemblies and fuels before irradiation; Chapter III: Transformations in reactor; Chapter IV: State of rods leaving the reactor; Chapter V: State of pellets; Chapter VI: Chemical and structural composition of the fuel; Chapter VII: Fuel fragmentation and density; Chapter VIII: Phenomena at the pellet periphery. Formation, characteristics and structure of the rim.Chemical interaction between pellet and cladding; Chapter IX: Location of fission gases and volatile fission products; Chapter X: Review, lessons and predictions. (authors)

  17. Strategy for nuclear wastes incineration in hybrid reactors; Strategies pour l'incineration de dechets nucleaires dans des reacteurs hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Lelievre, F

    1998-12-11

    The transmutation of nuclear wastes in accelerator-driven nuclear reactorsoffers undeniable advantages. But before going into the detailed study of a particular project, we should (i) examine the possible applications of such systems and (ii) compare the different configurations, in order to guide technological decisions. We propose an approach, answering both concerns, based on the complete description of hybrid reactors. It is possible, with only the transmutation objective and a few technological constraints chosen a posteriori, to determine precisely the essential parameters of such reactors: number of reactors, beam current, size of the core, sub-criticality... The approach also clearly pinpoints the strategic decisions, for which the scientist or engineer is not competent. This global scheme is applied to three distinct nuclear cycles: incineration of solid fuel without recycling, incineration of liquid fuel without recycling and incineration of liquid fuel with on-line recycling; and for two spectra, either thermal or fast. We show that the radiotoxicity reduction with a solid fuel is significant only with a fast spectrum, but the incineration times range from 20 to 30 years. The liquid fuel is appropriate only with on-line recycling, at equilibrium. The gain on the radiotoxicity can be considerable and we describe a number of such systems. The potential of ADS for the transmutation of nuclear wastes is confirmed, but we should continue the description of specific systems obtained through this approach. (author)

  18. Calculation of control rods in rectangular reactor, and applications (1960); Calcul des barres de conteole dans un reacteur rectangulaire et applications (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Goshen, S; Pazy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [French] Le but de ce rapport est de trouver une methode pour estimer l'antireactivite des barres de controle perpendiculaires a l'axe dans pile cylindrique. Le rapport se divise en deux parties. Dans la premiere nous donnons une methode de calcul des barres de controle dans une pile rectangulaire, analogue a la methode de Nordheim-Scalettar pour les piles cylindriques. A titre d'exemple, nous donnons les formules de theories a un et deux groupes de neutrons, la generalisation pour plusieurs groupes est evidente. Dans la deuxieme partie, nous trouvons, par une methode de variation, une formule qui permet d'estimer le laplacien d'une pile, qui peut etre divisee en parallelepipedes dont les laplaciens sont donnes. Nous utilisons enfin, cette formule pour calculer l'antireactivite des barres perpendiculaires a l'axe dans une pile cylindrique. (auteur)

  19. Oklo 2 Billion Years Before Fermi; Les reacteurs naturels d'Oklo (Gabon): 2 milliards d'annees avant Fermi

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    2005-02-15

    The author aims to present the little-known story of the Oklo natural reactors. He recalls the historical aspects of the Oklo reactors discovery by the CEA in 1972, he explains the scientific phenomenon and the interest, notably as a 'natural analogue' for the geological disposal of high level radioactive wastes. (A.L.B.)

  20. Detailed study of transmutation scenarios involving present day reactor technologies; Etude detaillee des scenarios de transmutation faisant appel aux technologies actuelles pour les reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    This document makes a detailed technical evaluation of three families of separation-transmutation scenarios for the management of radioactive wastes. These scenarios are based on 2 parks of reactors which recycle plutonium and minor actinides in an homogeneous way. A first scenario considers the multi-recycling of Pu and Np and the mono-recycling of Am and Cm using both PWRs and FBRs. A second scenario is based on PWRs only, while a third one considers FBRs only. The mixed PWR+FBR scenario requires innovative options and gathers more technical difficulties due to the americium and curium management in a minimum flux of materials. A particular attention has been given to the different steps of the fuel cycle (fuels and targets fabrication, burnup, spent fuel processing, targets management). The feasibility of scenarios of homogeneous actinides recycling in PWRs-only and in FBRs-only has been evaluated according to the results of the first scenario: fluxes of materials, spent fuel reprocessing by advanced separation, impact of the presence of actinides on PWRs and FBRs operation. The efficiency of the different scenarios on the abatement of wastes radio-toxicity is presented in conclusion. (J.S.)

  1. Contribution to the study of thermal-hydraulic problems in nuclear reactors; Contribution a l`etude de problemes de thermohydraulique dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G

    1998-07-07

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to participate in large international concerted actions is highlighted. (author) 64 refs.

  2. Contribution to the modelling of gas-solid reactions and reactors; Contribution a la modelisation des reactions et des reacteurs gaz-solide

    Energy Technology Data Exchange (ETDEWEB)

    Patisson, F

    2005-09-15

    Gas-solid reactions control a great number of major industrial processes involving matter transformation. This dissertation aims at showing that mathematical modelling is a useful tool for both understanding phenomena and optimising processes. First, the physical processes associated with a gas-solid reaction are presented in detail for a single particle, together with the corresponding available kinetic grain models. A second part is devoted to the modelling of multiparticle reactors. Different approaches, notably for coupling grain models and reactor models, are illustrated through various case studies: coal pyrolysis in a rotary kiln, production of uranium tetrafluoride in a moving bed furnace, on-grate incineration of municipal solid wastes, thermogravimetric apparatus, nuclear fuel making, steel-making electric arc furnace. (author)

  3. Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

    Energy Technology Data Exchange (ETDEWEB)

    Bonaccorsi, Th

    2007-09-15

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  4. 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)

  5. The sea water desalination by the nuclear reactors; Le dessalement de l'eau de mer par les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Nisan, S. [CEA Cadarache, Dir. du Developpement et de l' Innovation Nucleares DDIN, 13 - Saint-Paul-lez-Durance (France)

    2002-07-01

    This document underlines the importance of water shortage in many areas in the world in the future. The water sea desalination can be a efficient solution to this problem. The desalination methods are presented. In this context the desalination reactors appear as a competitive solution, facing the fossil energies systems not only for the simultaneous electric power and drinking water production, but also for the minimization of greenhouse gases. (A.L.B.)

  6. Relative measurement of the fluxes of thermal, resonant and rapid neutrons in reactor G1; Mesures relatives des flux thermique, resonnant et rapide dans le reacteur G1

    Energy Technology Data Exchange (ETDEWEB)

    Carle, R; Mazancourt, T de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    We sought to determine the behavior of the thermal, resonant and rapid neutron fluxes in the multiplier-reflector transition region, in the two principal directions of the system. We have also measured the variation of these different fluxes in the body of the multiplier medium in a canal filled with graphite and in an empty canal. The results are given in the form of curves representing: - the variation of the ratio of the thermal flux to the rapid flux in axial and radial transitions - the behavior of the thermal and resonant fluxes and the variation of their ratio in the same regions. (author) [French] Nous avons cherche a determiner le comportement des differents flux, thermique, resonnant et rapide a la transition milieu multiplicateur-reflecteur dans les deux directions principales du reseau. Nous avons egalement mesure la variation de ces differents flux au sein du milieu multiplicateur dans un canal rempli de graphite et dans un canal vide. Les resultats sont donnes sous forme de courbe representant: - La variation du rapport du flux thermique au flux rapide aux transitions axiale et radiale - L'allure des flux thermique et resonnant et la variation de leur rapport dans les memes regions. (auteur)

  7. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  8. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  9. 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)

  10. The development of fast reactors in France from March 1980 to March 1981; Le developpement des reacteurs a neutrons rapides en France de mars 1980 a mars 1981

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L. [Commissariat a l' Energie Atomique, CEN de Saclay, Gif-sur-Yvette (France)

    1981-05-15

    This paper describes general features concerning development in the field of fast reactors in France from March 1980 to March 1981. It concentrates mainly on: Rapsodie, Phenix NPP, prototype reactor Super Phenix 1, future fast reactor NPPs and current research and development programs in the field. The present situation is as follows. Rapsodie has restarted operation but at reduced power in July 1980 because of the problems in the primary circuit which have not yet been solved. Phenic operates in a very satisfactory manner. Construction of Super Phenix is continuing normally. Research activities are performed sometimes for the needs of Super Phenix and sometimes for the needs of future fast rector projects like Super Phenix 2. International cooperation is being continued.

  11. Research means to back the development of nuclear reactors; Les moyens de recherche en support a l'evolution des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    After 50 year long feedback experience on nuclear reactor operations it is legitimate to wonder whether experimental facilities used to support nuclear power programs are still necessary. The various participants of this conference said yes for mainly 4 reasons: -) to validate the extension of the service life of a reactor without putting at risk its high safety standard, -) to give the reactor more flexibility to cope with the power demand, -) to confront the results given by computerized simulations with experimental data, and -) to qualify the nuclear systems of tomorrow. (A.C.)

  12. Contribution to the optimization of the coupling of nuclear reactors to desalination processes; Contribution a l'optimisation du couplage des reacteurs nucleaires aux procedes de dessalement

    Energy Technology Data Exchange (ETDEWEB)

    Dardour, S

    2007-04-15

    This work deals with modelling, simulation and optimization of the coupling between nuclear reactors (PWR, modular high temperature reactors) and desalination processes (multiple effect distillation, reverse osmosis). The reactors considered in this study are PWR (Pressurized Water Reactor) and GTMHR (Gas Turbine Modular Helium Reactor). The desalination processes retained are MED (Multi Effect Distillation) and SWRO (Sea Water Reverse Osmosis). A software tool: EXCELEES of thermodynamic modelling of coupled systems, based on the Engineering Algebraic Equation Solver has been developed. Models of energy conversion systems and of membrane desalination processes and distillation have been developed. Based on the first and second principles of thermodynamics, these models have allowed to determine the optimal running point of the coupled systems. The thermodynamic analysis has been completed by a first economic evaluation. Based on the use of the DEEP software of the IAEA, this evaluation has confirmed the interest to use these types of reactors for desalination. A modelling tool of thermal processes of desalination in dynamic condition has been developed too. This tool has been applied to the study of the dynamics of an existing plant and has given satisfying results. A first safety checking has been at last carried out. The transients able to jeopardize the integrated system have been identified. Several measures aiming at consolidate the safety have been proposed. (O.M.)

  13. Calculation scheme for boiling water reactors cores; Methode de calcul des coeurs de reacteurs a eau bouillante par le systeme saphyr

    Energy Technology Data Exchange (ETDEWEB)

    Marsault, Ph [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SERSI), 13 - Saint-Paul-lez-Durance (France); Nicolas, A; Lenain, R; Richebois, E; Royer, E; Caruge, D [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France); Blaise, P [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPEX), 13 - Saint-Paul-lez-Durance (France); Gastaldi, B; Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPRC), 13 - Saint-Paul-lez-Durance (France)

    1999-07-01

    Boiling Water Reactors represent one third of the world's reactors. They are presently evolving towards greater simplification, allowing a reduction in the costs of operation, improved safety and a relative flexibility in their capacity to accommodate 100% MOX cores. The CEA, in a combined effort with its partners, the COGEMA and the EDF, would like to assess the interest of this reactor type, especially on this last point. A definition program and subsequent qualification of the calculation scheme have been undertaken. We are presenting here the specific features inherent in the calculation of these reactors, in comparison to PWRs, as well as the first results of the program. (authors)

  14. Integral validation of the effective beta parameter for the MOX reactors and incinerators; Validation integrale des estimations du parametre beta effectif pour les reacteurs Mox et incinerateurs

    Energy Technology Data Exchange (ETDEWEB)

    Zammit-Averlant, V

    1998-11-19

    {beta}{sub eff}, which represents the effective delayed neutron fraction, is an important parameter for the reactor nominal working as well as for studies of its behaviour in accidental situation. In order to improve the safety of nuclear reactors, we propose here to validate its calculation by using the ERANOS code with ERALIB1 library and by taking into account all the fission process physics through the {nu} energy dependence. To validate the quality of this calculation formalism, we calculated uncertainties as precisely as possible. The experimental values of {beta}{sub eff}, as well their uncertainties, have also been re-evaluated for consistency, because these `experimental` values actually contain a calculated component. We therefore obtained an entirely coherent set of calculated and measured {beta}{sub eff}. The comparative study of the calculated and measured values pointed out that the JEF2.2 {nu}{sub d} are already sufficient because the (E-C)/C are inferior to 3 % in average and in their uncertainly bars. The experimental uncertainties, even if lightly superior to those previously edited, remain inferior to the uncertainties of the calculated values. This allowed us to fit {nu}{sub d} with {beta}{sub eff}. This adjustment has brought an additional improvement on the recommendations of the {nu}{sub d} average values, for the classical scheme (thermal energy, fast energy) and for the new scheme which explains the {nu}{sub d} energy dependence. {beta}{sub eff}, for MOX or UOX fuel assemblies in thermal or fast configurations, can therefore be obtained with an uncertainty due to the nuclear data of about 2.0 %. (author) 110 refs.

  15. Cronos 2: a neutronic simulation software for reactor core calculations; Cronos 2: un logiciel de simulation neutronique des coeurs de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lautard, J J; Magnaud, C; Moreau, F; Baudron, A M [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)

    1999-07-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  16. Production and validation of nuclear data for reactor and fuel cycle applications; Production et validation des donnees nucleaires pour les applications reacteurs et cycle du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Trakas, C [Framatome ANP GmbH NBTT, Erlangen (Germany); Verwaerde, D [Electricite de France EDF, 75 - Paris (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); and others

    2002-07-01

    The aim of this technical meeting is the improvement of the existing nuclear data and the production of new data of interest for the upstream and downstream of the fuel cycle (enrichment, fabrication, management, storage, transport, reprocessing), for the industrial reactors, the research reactors and the new reactor concepts (criticality, dimensioning, exploitation), for the instrumentation systems (external and internal sensors), the radioprotection, the residual power, the structures (neutron bombardment effect on vessels, rods etc..), and for the activation of steel structures (Fr, Ni, Co). The expected result is the collection of more reliable and accurate data in a wider spectrum of energies and temperatures thanks to more precise computer codes and measurement techniques. This document brings together the communications presented at this meeting and dealing with: the process of production and validation of nuclear data; the measurement facilities and the big international programs; the users needs and the industrial priorities; the basic nuclear data (BND) needs at Cogema; the expression and evaluation of BND; the evaluation work: the efficient cross-sections; the processing of data and the creation of activation libraries; from the integral measurement to the qualification and the feedback on nuclear data. (J.S.)

  17. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  18. Study of trans-uranian incineration in molten salt reactor; Etude de l'incineration des transuraniens en reacteur a sel fondu

    Energy Technology Data Exchange (ETDEWEB)

    Valade, M

    2000-10-27

    With the emergence of new options for nuclear power, molten salt reactors are envisaged for waste management. The aim of this thesis is to show how molten salt reactors can help to solve the transuranics issue. Their high versatility regarding to isotopic vector allows to accommodate large fractions of minor actinides as compared to solid fuel system. In this thesis, a neutronics study of molten salt reactors, MSR, has been conducted. For this purpose, two reference systems were considered, TIER1 and AMSTER. In the case of TIER1, an optimisation was made to reach an equilibrium. The analysis of both systems showed the main characteristics of MSR: their link to chemistry and on line reprocessing. In this work, several methods to drive the system to a state of equilibrium have been implemented and compared. During this process the isotopic composition and neutron spectrum, thus the nuclear reaction cross sections, vary tremendously. It is essential to take these evolutions into account in order to accurately estimate the equilibrium state. This has been accomplished inside the multi-recycling procedure we set with ERANOS. A dedicated calculation schema has been realized to simulate superthermal systems with this computation code. These results were checked through a benchmark against other computer codes. Then, with multi-recycling method, several molten salt systems have been compared in order to define the optimal reactor for transuranics incineration. Nevertheless, a final choice can not only be done using only neutronics characteristics since chemistry and thermal-hydraulics constraints are really important for MSR. Moreover, a complete safety study would be required. (author)

  19. Analog study of the power control, temperature and flow of the reactor EL4; Etude analogique de la regulation de puissance, temperature et debit du reacteur EL4

    Energy Technology Data Exchange (ETDEWEB)

    Boulven, J; Chazal, G

    1966-07-01

    This study concerns the implementing and the exact determination of the gains of the plant control loops and the knowledge of the system response during possible solicitations under normal conditions. (A.L.B.)

  20. 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

  1. Critical experiments and nuclear calculations - LAMPRE-I; Experiences critiques et calculs nucleaires concernant le LAMPRE-I; Kriticheskie opyty i yadernye raschety - LAMPRE-I; Experimentos criticos u calculos nucleares relativos al LAMPRE-I

    Energy Technology Data Exchange (ETDEWEB)

    Battat, M E [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

    1962-03-15

    As part of a programme to develop plutonium fuels for fast-breeder reactors, the Los Alamos Scientific Laboratory has constructed and is operating a 1-MW sodium-cooled test reactor whose core contains a molten alloy of plutonium andiron (90 at. % Pu, 10 at. % Fe, m.p. 410 deg. C). Reactivity control is provided by the use of a stainless-steel reflector and four nickel control-rods located external to the core. Experiments have been performed at core temperatures (isothermal) of 80, 160 and 480 deg. C to determine critical mass and reflector worth at each of these temperatures. Control-rod worths, from period measurements, and temperature coefficient of reactivity were also measured. Calculations have been made, using the S{sub n} method for solving the neutron transport problem, to determine the basic nuclear parameters of the system. The comparison between calculated and measured values of parameters such as temperature coefficient, control-element worths, and critical mass is also of interest in evaluating the reliability of the design calculations. (author) [French] Un reacteur d'essais de 1 MW refroidi au sodium, dont le coeur contient un alliage fondu de plutonium et defer (90 at. % Pu, 10 at. % Fe, p. f. 410 deg. C), a ete construit et est en fonctionnement au Laboratoire scientifique de Los Alamos, dans le cadre d'un programme d'etudes sur les combustibles au plutonium pour reacteurs surgenerateurs a neutrons rapides. Le controle de la reactivite est assure au moyen d'un reflecteur en acier inoxydable et de quatre barres de controle en nickel, a l'exterieur du coeur. On a fait des experiences a des temperatures du coeur de 80, 160 et 480 deg. C afin de determiner la masse critique et la quantite de reflecteur qui correspond a chacune de ces temperatures. On a aussi mesure l'efficacite des barres de controle, a partir de mesures de periode, ainsi que le coefficient thermique de reactivite. Afin de determiner les parametres nucleaires de base du reacteur, on a

  2. Modeling of acoustic wave propagation and scattering for telemetry of complex structures; Modelisation de la propagation et de l'interaction d'une onde acoustique pour la telemetrie de structures complexes

    Energy Technology Data Exchange (ETDEWEB)

    LU, B.

    2011-11-07

    ) using a procedure similar to the physical theory of diffraction (PTD). The refined KA provides an improvement of the prediction in the near field of a rigid scatterer. The initial (non refined) KA model is then extended to deal with the scattering from a finite impedance target. The obtained model, the so-called 'general' KA model, is a satisfactory solution for the application to telemetry. Finally, the coupling of the stochastic propagation model and the general KA diffraction model has allowed us to build a complete simulation tool for the telemetry in an inhomogeneous medium. (author) [French] Cette etude s'inscrit dans le cadre du developpement d'outils de simulation de la telemetrie qui est une technique possible pour la surveillance et le controle periodique des reacteurs nucleaires a neutrons rapides refroidis par du sodium liquide (RNR-Na). De maniere generale, la telemetrie consiste a positionner au sein du reacteur un transducteur qui genere un faisceau ultrasonore. Ce faisceau se propage a travers un milieu inhomogene et aleatoire car le sodium liquide est le siege de fluctuations de temperature qui impliquent une variation de la celerite des ondes ultrasonores, ce qui modifie la propagation du faisceau. Ce dernier interagit ensuite avec une structure immergee dans le reacteur. La mesure du temps de vol de l'echo recu par le meme transducteur permet de determiner la position precise de la structure. La simulation complete de la telemetrie necessite donc la modelisation a la fois de la propagation d'une onde acoustique en milieu inhomogene aleatoire et de l'interaction de cette onde avec des cibles de formes variees; c'est l'objectif de ce travail. Un modele stochastique base sur un algorithme de type Monte-Carlo est tout d'abord developpe afin de simuler les perturbations aleatoires du champ de propagation. Le champ acoustique en milieu inhomogene est finalement modelise a partir du champ calcule dans un

  3. The Scottish Research Reactor Centre and its Facilities for the Production and Exploitation of Short-Lived Radioisotopes; Le Réacteur de Recherche Ecossais et ses Installations pour la Production et l'Exploitation des Radioisotopes a Courte Periode; ШОТЛАНДСКИЙ ИССЛЕДОВАТЕЛЬСКИЙ РЕАКТОРНЫЙ ЦЕНТР И ЕГО ТЕХНИЧЕСКИЕ СРЕДСТВА ДЛЯ ПРОИЗВОДСТВА И ИСПОЛЬЗОВАНИЯ КОРОТКОЯИВУШХ РАДИОИЗОТОПОВ; El Centro del Reactor de Investigacion de Escocia y sus Instalaciones para la Produccion y Empleo de Radioisotopos de Periodo Corto

    Energy Technology Data Exchange (ETDEWEB)

    Ward, A. [The Royal College of Science and Technology, Glasgow, Scotland (United Kingdom)

    1963-03-15

    The Scottish Research Reactor Centre is now under construction and will be completed and in operation by the summer of 1963. The reactor is 100 kW of the tank type with water cooling and water/graphite moderation using enriched U{sup 235} fuel. The experimental facilities include a large thermal column, a large shield experiment water tank and a radioisotope production facility with transfer rabbit tubes. There are effectively three through tubes in the central core; one through tube in the thermal column, several small central vertical stringers and one six-inch square vertical stringer penetrating to the centre of the core. Many horizontal stringers pass through the thermal column, the central one penetrating to within one inch of a fuel tank. The reactor facilities are supported by a wide variety of adjacent small laboratories. These include hot source handling and preparation facilities, changing rooms, electrical and mechanical workshops, darkrooms, microcurie laboratories, animal house, biological and chemical laboratories, low-background counting room, lecture theatre and a library. It is expected that the research will extend over many scientific and technological disciplines; a good proportion of the work will involve short-lived radioisotopes and typical projects are described. (author) [French] Le reacteur de recherche ecossais actuellement en construction sera termine et entrera en service pendant l'ete 1963. Il s'agit d'un reacteur-piscine de 100kW a uranium-235 enrichi, refroidi a l'eau eti ralenti a l'eau et au graphite. Les installations experimentales comprennent une grande colonne thermique, un grand reservoir d'eau blinde pour les experiences, et une installation de production de radioisotopes munie de tubes de transfert a fonctionnement pneumatique. Trois tubes traversent le coeur central, un autre traverse la colonne thermique. Il existe eniin plusieurs petites gouttieres verticales centrales et une gouttiere verticale a section carree de 15 cm

  4. Current market of industrial bio-products and biofuels, and predictable evolutions by 2015/2030. Synthesis; Marche actuel des bioproduits industriels et des biocarburants et evolutions previsibles a echeance 2015 / 2030. Synthese

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    The main objectives of this study were to describe the current status of the energetic and industrial bio-product markets (biofuels, bio-lubricants, biomaterials, papers, cosmetics, and so on), to identify and analyze the evolution perspectives of these new markets on a long and medium term, to define scenarios of evolution for different sectors (agro-industry, energy, organic chemistry), to identify the most promising new markets, and to select the priority agro-industrial sectors

  5. Simulation of the thermal behaviour of electric industrial resistance furnaces using the I-DEAS/TMG code; Simulation du comportement thermique des fours electriques industriels a resistances a l`aide du code I-DEAS/TMG

    Energy Technology Data Exchange (ETDEWEB)

    Plard, Ch; Branchu, K; Le Cloirec, B [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches

    1997-12-31

    In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

  6. Application of a radiant heat transfer model to complex industrial reactive flows: combustion chambers, electric arcs; Application d`un modele de transfert radiatif a des ecoulements reactifs industriels complexes: chambres de combustion, arcs electriques

    Energy Technology Data Exchange (ETDEWEB)

    Mechitoua, N; Dalsecco, S; Delalondre, C; Simonin, O [Electricite de France (EDF), 78 - Chatou (France). Lab. National d` Hydraulique

    1997-12-31

    The direction of studies and researches (DER) of Electricite de France (EdF) has been involved for several years in a research program on turbulent reactive flows. The objectives of this program concern: the reduction of pollutant emissions from existing fossil-fueled power plants, the study of new production means (fluidized beds), and the promotion of electric power applications in the industry. An important part of this program is devoted to the development and validation of 3-D softwares and to the modeling of physical phenomena. This paper presents some industrial applications (furnaces, boilers, electric arcs) for which radiant heat transfers play an important role and the radiation models used. (J.S.) 8 refs.

  7. Towards a better use of digestate. For agricultural and industrial entrepreneurs with a basic knowledge on co-digestion; Naar een betere toepassing van digestaat. Voor agrarische en industriele ondernemers met basiskennis over (co-)vergisting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-06-15

    Fermentation of various digestable materials not only produces biogas, but also electricity, heat, green gas, and digestates (residual matter). The digestates can be processed and applied in various ways. In this brochure digestion is discussed in general and attention is paid to fermentation of, for example, waste from vegetables, fruits and gardens. [Dutch] Bij vergisting van verschillende vergistbare stoffen ontstaat naast biogas, elektriciteit, warmte en groen gas, ook digestaat. Digestaat is het restproduct dat overblijft. Dit digestaat kan op verschillende manier worden bewerkt en toegepast. Deze brochure behandelt vergisting in het algemeen maar ook vergisting van bijvoorbeeld GFT (groente-, fruit- en tuinafval)

  8. New trends in processing and disposal of municipal and industrial sewage sludges. 8. Joint seminar `waste water technology` with exhibitor`s forum; Neue Trends bei der Behandlung und Entsorgung kommunaler und industrieller Klaerschlaemme. Achtes gemeinsames Seminar `Abwassertechnik` mit Ausstellerforum

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The present conference proceedings handle all aspects of sewage sludge disposal: agricultural use of sewage sludge, processing in anaerobic reactors, composting, drain, drying, gasification, combustion, economical aspects, ecological aspects, building, design and operation of modern sludge treatment plants. (SR) [Deutsch] Der vorliegende Tagungsband behandelt alle Aspekte der Klaerschlammentsorgung: landwirtschaftliche Nutzung von Klaerschlamm, Behandlung in Anaerobreaktoren, Kompostierung, Entwaesserung, Trocknung, Vergasung, Verbrennung, wirtschaftliche Aspekte, oekologische Aspekte, Bau, Ausruestung und Betrieb von modernen Schlammbehandlungsanlagen. (SR)

  9. A review on the Cigeo project, the industrial centre of geological storage of the most radioactive wastes; Le point sur le projet Cigeo, centre industriel de stockage geologique pour les dechets les plus radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-15

    This document briefly presents the Cigeo project which is designed for the underground geological storage of the most radioactive wastes. Requirements comprise safety after closure and without any human intervention, and a reversible operation during at least 100 years. The storage principle is briefly described. A brief history of this research project is reported

  10. The HILW-LL (high- and intermediate-level waste, long-lived) disposal project: working toward building the Cigeo Industrial Centre for Geological Disposal; Le projet HA-MAVL: vers la realisation du centre industriel de stockage geologique Cigeo

    Energy Technology Data Exchange (ETDEWEB)

    Labalette, Th. [Agence Nationale pour la Gestion des Dechets Radioactifs - ANDRA, Dir. des Projets, 92 - Chatenay Malabry (France)

    2011-02-15

    The French Act of 28 June 2006 identifies reversible disposal in deep geological facilities as the benchmark solution for long-term management of high-level waste (HLW) and for intermediate-level long-lived waste (ILW-LL). The Act tasks ANDRA (national agency for the management of radioactive wastes) with the pursuit of studies and research on the choice of a site and the design of the repository, with a view to examining the licence application in 2015 and, provided that the licence is granted, to make the facility operational by 2025. At the end of 2009, ANDRA submitted to the Government its proposals regarding the site and the design of the Industrial Centre for Geological Disposal, known as CIGEO. With the definition of a possible area for the construction of underground disposal facilities, one of the key stages in the project has been achieved. The choice of a surface site will be validated following the public consultation scheduled for the end of 2012. The project is now on the point of entering the definition stage (preliminary design). CIGEO will be a nuclear facility unlike any other. It will be built and operated for a period of over 100 years. For it to be successful, the project must meet certain requirements related to its integration in the local area, industrial planning, safety and reversibility, while also controlling costs. Reversibility is a very important concept that will be defined by law. It is ANDRA's responsibility to ensure that a reasonable balance is found between these different concerns. (author)

  11. Dialogue around industrial sites. Synthesis of a thinking method of I.R.S.N; Concertation autour des sites industriels. Synthese d'une demarche de reflexion de l'IRSN

    Energy Technology Data Exchange (ETDEWEB)

    Sugier, A.; Oudiz, A. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Heriard Dubreuil, G.; Gadbois, S. [Mutadis, 94 - Vitry (France); Schneider, Th. [Centre d' Etude sur l' Evaluation de la Protection dans le Domaine Nucleaire CEPN, 92 - Fontenay aux Roses (France)

    2003-12-15

    The present report gives an account of results on a research work about 'the stakes of the dialogue around the follow up of nuclear and non nuclear industrial facilities' and on conclusions of a seminar, on the same subject that stood at Ville D' Avray from the 21. to 22. of January 2003. This seminar has gathered different actors (administration, experts, associations, industrial operators) concerned by the dialogue around these installations. The work has been directed by I.R.S.N. and had for object to give the knowledge of the French and International experience in matter of dialogue around nuclear and non nuclear industrial sites. (N.C.)

  12. Cryo magnetic separation adaptation to environment technologies: application to industrial effluents; Adaptation de la separation cryomagnetique aux technologies de l`environnement: application a l`epuration d`effluents liquides industriels

    Energy Technology Data Exchange (ETDEWEB)

    Bureau, V

    1993-12-20

    Cryomagnetic separation adaptation to environment technologies application to industrial liquid effluents. The performance, obtained by superconducting high filed - high gradient magnetic separation, permitted to foresee the magnetic treatment of heavy metals in rinse waters, derived from the surface finishing industry. The paramagnetic ions, precipitated in basic media as hydroxides, present a very hydrated amorphous structure, which masks their subjacent magnetic properties. Coprecipitation of a `magnetic carrier`, jointly with the heavy metals, has been studied: ferric chloride forms in basic media, an hydrated iron oxide. Its structure is of the goethite type, and it stabilizes as hematite. The magnetic susceptibility of the obtained product is still weak and its crystalline structure is not enough affirmative to utilize magnetic filtration with efficiency. Mixture of ferrous sulphate and ferric chloride forms, in a basic media, an hydrated magnetite. Initial ideal ratio between divalent iron and trivalent iron, varies between 0,5 and 1,2. This mixture, coprecipitated with the heavy metals, permits to optimize the magnetic cleaning of the fluids in a high field - high gradient filter. (author)

  13. Detection, in real time, of metallic pollutants present in the industrial atmospheric effluents by inductively coupled plasma torch; Detection, en temps reel, d'elements metalliques presents dans les rejets atmospheriques industriels par torche a plasma a couplage inductif

    Energy Technology Data Exchange (ETDEWEB)

    Vacher, D.

    2001-12-15

    This work is devoted to the development of a process of detection in real time of metallic pollutants present in industrial atmospheric effluents. The method of measurement is the atomic spectrometry of emission coupled to an ICP torch (Inductively coupled Plasma). The technology of the fluidized beds is used as system of introduction of the metallic particles into the ICP torch, the interest of the principle of detection resting on the stamping from the usual procedure of calibration of the analytical system. The results are presented in two parts. The first relates to the diagnosis of plasmas formed with various mixtures of N{sub 2}/O{sub 2} which one corresponds to pure air, the second presents the setting process of detection in real time starting from the intensities ratios of the spectral lines of the metallic element with those of the plasma-producing element (argon or pure air) The study of the diagnosis of plasmas made up of mixtures N{sub 2}/O{sub 2} relates to the determination of the atomic excitation temperature from the spectral lines of the copper element and the evaluation of the thermal disequilibrium q Te/Th. This last is obtained by considering the mass enthalpy of various mixtures N{sub 2}/O{sub 2}. The existence of a small thermal disequilibrium is highlighted. The study of detection in real time by ICP torch, without calibration of the system, is based on three points: - spectroscopic data processing to determine the values of the intensities ratios of spectral lines; - the insertion of the intensities ratios and the characteristics of plasma (argon or pure air) into a calculation code of plasma composition; - the comparison of the mass flux values of the metallic pollutants, in real time, obtained by experiments with those resulting from the elutriation calculation, term which defines the phenomenon of entrainment of the particles out of the fluidized bed. The results made it possible to show the similarity of the analytical system response between the use of argon plasma and that of air. (author)

  14. Perceptions of industrial and nuclear risks. Stakes, negotiations and social development of levels of risk acceptance; Perceptions des risques industriels et nucleaires: enjeux, negociations et construction sociale des seuils d'acception des risques

    Energy Technology Data Exchange (ETDEWEB)

    Bernier, S.Ch

    2007-11-15

    In this thesis we will question the perceptions of industrial risks in the occidental world at the beginning of the 21. century. For this purpose we will try to understand how concepts such as sustainable development, precautionary principle, liability, or even zero-risk bias have progressively developed around a thought model based on the scientific rationality. This model is now undermined by its incapacity to fully address the issues it raises and completely avoid the potential risks. However, despite consistent weaknesses, it remains a reference value moulded by past accidents which have led to the making of laws aiming mainly at defining liability and protecting those who are held liable. Thus, public information becomes a requirement for democracy and the protection of this thought model. In this context, the protagonists at stake are security-conscious, economical and political lobbies that constantly redefine the limits of risk acceptance. We come to the realization that our lifestyle and value system remain unchallenged even though undergoing a crisis. The specificity of this research lies into the importance we give to the local approach, dealing with registered Seveso sites and nuclear plants located in Indre et Loire. We have polled five categories of respondents through interviews or questionnaires in order to understand their opinion regarding situations involving technological risks. The result of this survey helps us understand and set the levels of risk acceptance that they define with regard to the industrial risks and show the complexity of a situation involving political stakes, environmental pressures, a profit-driven economy and security constraints, in a vague and complex context. This work gives us a contrasted picture of today's perceptions of risks. (author)

  15. Speciation and evolution of cyanide compounds contained in industrial residues from coal pyrolysis; Speciation et evolution des composes cyanures contenus dans des residus industriels issus de la pyrolyse de la houille

    Energy Technology Data Exchange (ETDEWEB)

    Proffit, D.

    2002-10-15

    Occurrence of cyanide compounds, mainly [Fe(CN)6]2-/3-, in soils and groundwater is due to industrial waste deposits. The main polluting source are purifier wastes stored on former manufactured gas plant sites. In order to estimate the environmental risks associated with cyanide, a specific analytical procedure combining analyses on the solids (optical and scanning electron microscopies, infrared spectroscopy and x-ray diffraction) with analyses carried out on leachates obtained by aqueous extraction at various pHs was developed. For a given purifier waste, Prussian blue was evidenced as the major cyanide species but other minor compounds were also observed. Some of them are extremely soluble (e.g. potassium ferrocyanide) whereas others are very stable (e.g. potassium and zinc ferrocyanide). The study of other polluted solids allowed to predict the hazards linked with their storage. Measurements realized on percolation columns effluents revealed that cyanide species can migrate as colloidal species as well as under soluble forms, which confirmed some site observations where cyanide pollution was observed in soils located beneath a priori stable wastes. Such a combination of techniques can then be considered as a useful diagnosis and risk assessment tool. However, as the full procedure is rather time-consuming a partial combination of the techniques developed can even be used advantageously in some specific cases. We also developed a quantification and speciation method (spectrometric UV-Visible - mathematical code) to determine cyanide compounds concentrations in water extraction filtrates. Finally, irradiation tests of purifier wastes were carried out. They revealed that the influence of natural light could be considered as negligible in comparison to the other factors affecting cyanide compounds migration.

  16. Energetic retrofitting of industrial heat supply systems. Possibilities of enhancing the efficiency and energy conservation at large combustion engineering plants; Energetische Modernisierung industrieller Waermeversorgungssysteme. Moeglichkeiten der Effizienzsteigerung und der Energieeinsparung an grossen feuerungstechnischen Anlagen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-12-15

    In the contribution under consideration, the Deutsche Energie-Agentur GmbH (Berlin, Federal Republic of Germany) reports on an energetic modernization of industrial heat supply systems. Possibilities of an enhancement of the energetic efficiency and energy conservation at large combustion engineering plants are described. After an introduction to this theme, the author of this contribution provides an overview of the optimization of heat supply systems, and reports on the following aspects: Optimisation of the heat demand; energy efficient heat generation; heat recovery; energy efficient conversion technology and generation technology; associate partners for more energy efficiency in industry and commerce; best practice examples.

  17. Simulation of the thermal behaviour of electric industrial resistance furnaces using the I-DEAS/TMG code; Simulation du comportement thermique des fours electriques industriels a resistances a l`aide du code I-DEAS/TMG

    Energy Technology Data Exchange (ETDEWEB)

    Plard, Ch.; Branchu, K.; Le Cloirec, B. [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches

    1996-12-31

    In order to answer the modeling needs of manufacturers and users of electrical resistance furnaces, Electricite de France (EdF) has been appealed to search for a numerical simulation tool for the modeling of thermal phenomena and to carry out its qualification. The TNG software has been retained according to its modeling characteristics of radiant heat transfers and to the coupling with thermal conduction. After a description of the main characteristics of this code, two examples of application to electrical furnaces are presented. The first example illustrates how it can be possible to accurately reproduce the behaviour of a big industrial furnace. The second example is an illustration of numerical simulation possibilities for the optimization of processes performed with an electric furnace. (J.S.)

  18. Application of a radiant heat transfer model to complex industrial reactive flows: combustion chambers, electric arcs; Application d`un modele de transfert radiatif a des ecoulements reactifs industriels complexes: chambres de combustion, arcs electriques

    Energy Technology Data Exchange (ETDEWEB)

    Mechitoua, N.; Dalsecco, S.; Delalondre, C.; Simonin, O. [Electricite de France (EDF), 78 - Chatou (France). Lab. National d`Hydraulique

    1996-12-31

    The direction of studies and researches (DER) of Electricite de France (EdF) has been involved for several years in a research program on turbulent reactive flows. The objectives of this program concern: the reduction of pollutant emissions from existing fossil-fueled power plants, the study of new production means (fluidized beds), and the promotion of electric power applications in the industry. An important part of this program is devoted to the development and validation of 3-D softwares and to the modeling of physical phenomena. This paper presents some industrial applications (furnaces, boilers, electric arcs) for which radiant heat transfers play an important role and the radiation models used. (J.S.) 8 refs.

  19. Aactor GT - Development of an inverse gas-turbine that uses renewable energy sources and industrial waste heat - Phase 2; Aactor GT - Entwicklung einer Inversen Gasturbine 'Aactor' zur Nutzung erneuerbarer Energie und industrieller Abwaerme. Phase 2 - Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Schmid, M.

    2009-09-15

    This final report for the Swiss Federal Office of Energy (SFOE) presents the results of the second phase of the development of a small-scale gas turbine that can use lean gas. The aim of this project phase - the design of a project development unit (PDU) for a micro-turbine with a nominal grid feeding power of 2.4 kWe is discussed. Parallel to this work, peripheral components such as burner and recovery device shall also be designed, produced and tested in the laboratories of the Swiss Center for Appropriate Technology and Social Ecology. The burner is specially designed for the combustion of lean gases. The goals of the following project phase, including the production and field-testing of the unit are discussed. On the basis of this PDU, a prototype lean gas micro-turbine with 9 kWe electrical generation power is to be derived. Project goals, work done and results obtained are reviewed, as is further work to be done.

  20. The evolution of the energy demand in France in the industrial, residential and transportation sectors; L'evolution de la demande energetique en France dans les secteurs industriel, residentiel, et des transports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    This document provides information, from 1970 to 2005, on the evolution of the energy intensity (ratio between the primary energy consumption and the gross domestic product in volume) and the actions of energy control for the industrial, residential and transportation sectors. (A.L.B.)

  1. Identification de facteurs génétiques, environnementaux et technologiques associés à la variabilité de la valeur nutritionnelle du blé et des produits industriels dérivés

    OpenAIRE

    Nurit , Eric

    2015-01-01

    Wheat is the second largest crop cultivated around the world and constitutes a major part of the daily diet in Europe. During the course of improving the baking quality of wheat cultivar, most of the nutritional attributes have been underestimated. It is therefore unfortunate that most of wheat-based food products are mostly produced from refined white flour from which peripheral tissues (germ and envelopes) are removed. However, these tissues, which are eliminated and serve mainly for animal...

  2. New project in Europe of safety enhancement of an industrial Seveso 2 classified site; Projet inedit en Europe de securisation d'un site industriel classe seveso 2

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2006-03-15

    Butagaz company has developed a new technical solution to reinforce the safety of its propane storage site of Aumale. This solution is based on the construction of a cylindrical reinforced concrete envelope around the spherical gas tank in order to greatly reduce the risk of boiling liquid expanding vapor explosion (BLEVE). Short paper. (J.S.)

  3. Selective catalytic reduction of nitrogen oxides from industrial gases by hydrogen or methane; Reduction catalytique selective des oxydes d'azote (NO{sub x}) provenant d'effluents gazeux industriels par l'hydrogene ou le methane

    Energy Technology Data Exchange (ETDEWEB)

    Engelmann Pirez, M

    2004-12-15

    This work deals with the selective catalytic reduction of nitrogen oxides (NO{sub x}), contained in the effluents of industrial plants, by hydrogen or methane. The aim is to replace ammonia, used as reducing agent, in the conventional process. The use of others reducing agents such as hydrogen or methane is interesting for different reasons: practical, economical and ecological. The catalyst has to convert selectively NO into N{sub 2}, in presence of an excess of oxygen, steam and sulfur dioxide. The developed catalyst is constituted by a support such as perovskites, particularly LaCoO{sub 3}, on which are dispersed noble metals (palladium, platinum). The interaction between the noble metal and the support, generated during the activation of the catalyst, allows to minimize the water and sulfur dioxide inhibitor phenomena on the catalytic performances, particularly in the reduction of NO by hydrogen. (O.M.)

  4. Gassmaks. Study of requirement for national focus on research for increased value-added industrial process of natural gas. Final report; Gassmaks. Utredning av behov for nasjonal satsing paa forskning for oekt verdiskaping fra naturgass gjennom industriell foredling. Endelig rapport

    Energy Technology Data Exchange (ETDEWEB)

    2006-08-15

    Final report concludes the importance of establishing the Research and Development program called 'Gassmaks'. The target of this program is increased value added to the natural gas loop. Strengthened know-how, industrial development and international competition force shall contribute to higher value added to community through industrial refining of natural gas. Gassmaks will by research based foundation exploit Norwegian natural gas resources environmental friendly. Highly prioritised are converting and use of natural gas to plastic raw materials, synthesis gas, synthetical fuel, energy processes, carbon materials, metallurgical processes and nutrients as proteins and fat. (AG). 28 refs., 3 figs., 3 tabs

  5. Modeling and experimental study of heat transfer in innovate building components for industrial production; Modelisation et etude experimentale des echanges de chaleur dans des composants innovants de batiment industriels

    Energy Technology Data Exchange (ETDEWEB)

    Lacena-Neildez, A.

    2000-05-15

    The main objective of the thesis is the study of a new roof thermal behaviour with the thermal transfer by radiation and convection in order to propose an alternative to the mechanical air-conditioning in the overseas islands, and to determine the way of improving the energy performance. Two fundamental results were obtained. First, the coating has a great importance in relation with the radiative properties. Secondly, air channel geometries can complete insufficient coatings in their cooling purpose. In this thesis, both experimental and modelling analyses were carried out. A solar simulator was used for the experiment to carry out a comparative study of prototypes. A performance indicator was defined; it is the eliminated flux, because the temperature is not sufficient to describe the heat exchange. At the same time, the radiative properties of coatings were measured and a characterisation of paints allows to start a new research of innovative infrared reflective paints. A mono-dimensional model was developed describing thermal exchanges in an air channel component. A sensibility study was carried out which allows to determine the main parameters. The experimental study validated the model in rear flux, more realistic than stagnation. Finally, the feasibility of this new steel component was concluded by a technical-economical study. (author)

  6. Facteurs socio-économiques affectant l'utilisation des sous-produits agro-industriels pour l'embouche bovine à contre-saison dans l'Adamaoua, Cameroun

    Directory of Open Access Journals (Sweden)

    Deffo V.

    2009-01-01

    Full Text Available Socio-economic factors affecting the use of agro-industrial by-products for cattle fattening in the dry season in Adamawa, Cameroon. Cattle production is the major economic activity in the Adamawa. Feed deficiency that causes about 129 to 187 g weight loss per day is an important constraint during the dry season. A possible alternative to overcome this constraint is the use of agro-industrial by-products. However, the adoption and effective use of these potentials are still to be encouraged. This study, which objective was to find out the socio-economic factors that may affect the usage of these resources, permitted to show, through interviews and surveys, that more than 8,200 t of agro-industrial by-products (maize and wheat bran, soybean seed and maize seed cake useable for cattle feeding were produced annually in Adamawa and that only 16 % of this production were used by livestock farmers. Among the factors affecting the effective use of these by-products, prices were the most determinant. The herds size and the number of sedentary animals had also shown a significant positive effect. On the other hand, livestock farming experience and the farmer’s age had instead shown a strong negative correlation with the use of by-products; same was the level of farmer’s education which showed positive correlation only with respect to cotton seed cake use. The distance from the livestock farming sites to the by-products production/distribution centre and/or difficult access to the sites had strong positive correlations with the by-products’ prices. The problem of supplies as a result of long distance or of difficult access to cattle production sites, the high and unstable prices of agro-industrial by-products and poor awareness of the different types of by-products produced in the Adamawa, were noted as major constraints for their use as cattle feed. Based on the above results, an effective extension system and well organized farmers groupings could improve the level of utilization of agro-industrial by-products for cattle feeding.

  7. Les performances mécaniques et thermiques d’un béton léger à base de déchets industriels solides et de granulats de bois

    Directory of Open Access Journals (Sweden)

    Mohamed Larbi Benmalek

    2014-04-01

    Les résultats expérimentaux montrent qu’à densité égale, les matériaux élaborés sont mécaniquement et thermiquement compétitifs. Ils sont intéressants du point de vue confort et stockage thermiques.

  8. CO{sub 2} emissions abatement and geologic sequestration - industrial innovations and stakes - status of researches in progress; Reduction des emissions et stockage geologique du CO{sub 2} - innovation et enjeux industriels - le point des recherches en cours

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This colloquium was jointly organized by the French institute of petroleum (IFP), the French agency of environmental and energy mastery (Ademe) and the geological and mining research office (BRGM). This press kit makes a status of the advances made in CO{sub 2} emissions abatement and geological sequestration: technological advances of CO{sub 2} capture and sequestration, geological reservoir dimensioning with respect to the problem scale, duration of such an interim solution, CO{sub 2} emissions abatement potentialities of geological sequestration, regulatory, economical and financial implications, international stakes of greenhouse gas emissions. This press kit comprises a press release about the IFP-Ademe-BRGM colloquium, a slide presentation about CO{sub 2} abatement and sequestration, and four papers: a joint IFP-Ademe-BRGM press conference, IFP's answers to CO{sub 2} emissions abatement, Ademe's actions in CO{sub 2} abatement and sequestration, and BRGM's experience in CO{sub 2} sequestration and climatic change expertise. (J.S.)

  9. Interactions between industrial organic pollutants and rhizosphere components and documentation of material streams in plant-based wastewater treatment plants - laboratory experiments; Wechselwirkungen industrieller organischer Schadstoffe mit Rhizosphaerenkomponenten und Bilanzierung von Stoffstroemen in Pflanzenklaeranlagen - Laborversuche

    Energy Technology Data Exchange (ETDEWEB)

    Plugge, J.

    2001-07-01

    The purpose of the present study was to examine the suitability of plant/soil systems for cleaning organically polluted effluents and to assess the influence of plant growth and dissolved humic substances on processes leading to the elimination of organic pollutants. This involved an examination of sorption interactions between selected pollutants on the one hand and sand and root material on the other, use of vertically irrigated plant-bearing sand columns for simulating real plant-based wastewater treatment plants, assessment of the cleaning efficiency of these systems with respect to the employed model pollutants and determination of the contamination of the filter material and plants with pollutants. Radiotracer techniques were used to determine pollution paths of phenanthrene and its microbial degradation in the model system. [German] In der vorliegenden Arbeit wurde die Eignung von Pflanze/Boden-Systemen zur Reinigung carbochemisch belasteter Abwaesser untersucht und der Einfluss eines Pflanzenbewuchses sowie geloester Huminstoffe auf die Prozesse, die zur Entfernung organischer Schadstoffe fuehren, bewertet. Die Bearbeitung dieses Themas umfasste Untersuchungen zu Sorptionswechselwirkungen ausgewaehlter Schadstoffe mit Sand- und Wurzelmaterial, die Anwendung vertikal durchstroemter, bepflanzter Sandsaeulen zur Nachbildung realer Pflanzenklaeranlagen, die Erfassung der Reinigungseffizienz dieser Systeme fuer die Modellschadstoffe sowie die Bestimmung der Schadstoffkontamination des Filtermaterials und der Pflanzen. Unter Anwendung der Radiotracertechnik erfolgte darueber hinaus die Bestimmung der Schadstoffpfade von Phenanthren einschliesslich des mikrobiellen Abbaus im Modellsystem. (orig.)

  10. 4. S.F.R.P. days on the optimization of radiation protection in the electronuclear, industrial and medical areas; 4. journees SFRP sur l'optimisation de la radioprotection dans les domaines electronucleaire, industriel et medical

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    These days are dedicated to the implementation of the radiation protection optimization in the activities of the electronuclear sector, of the industrial sector, the medical sector, the laboratories and the centers of research and the university sector. All the aspects of the practical application of the radiation protection optimization of the workers, the public and the patients will be approached. The oral communications and posters concern the following subjects: foundations of the optimization principle, new statutory context, transmission of ALARA principle, operational dosimetry, conception, operating and maintenance of the installations, the construction sites of dismantling, industrial radiology, radioactive waste management. (N.C.)

  11. Reduction of emissions and geological storage of CO{sub 2}. Innovation an industrial stakes; Reduction des emissions et stockage geologique du CO{sub 2}. Innovation et enjeux industriels

    Energy Technology Data Exchange (ETDEWEB)

    Mandil, C.; Podkanski, J.; Socolow, R.; Dron, D.; Reiner, D.; Horrocks, P.; Fernandez Ruiz, P.; Dechamps, P.; Stromberg, L.; Wright, I.; Gazeau, J.C.; Wiederkehr, P.; Morcheoine, A.; Vesseron, P.; Feron, P.; Feraud, A.; Torp, N.T.; Christensen, N.P.; Le Thiez, P.; Czernichowski, I.; Hartman, J.; Roulet, C.; Roberts, J.; Zakkour, P.; Von Goerne, G.; Armand, R.; Allinson, G.; Segalen, L.; Gires, J.M.; Metz, B.; Brillet, B

    2005-07-01

    An international symposium on the reduction of emissions and geological storage of CO{sub 2} was held in Paris from 15 to 16 September 2005. The event, jointly organized by IFP, ADEME and BRGM, brought together over 400 people from more than 25 countries. It was an opportunity to review the international stakes related to global warming and also to debate ways of reducing CO{sub 2} emissions, taking examples from the energy and transport sectors. The last day was dedicated to technological advances in the capture and geological storage of CO{sub 2} and their regulatory and economic implications. This document gathers the available transparencies and talks presented during the colloquium: Opening address by F. Loos, French Minister-delegate for Industry; Session I - Greenhouse gas emissions: the international stakes. Outlook for global CO{sub 2} emissions. The global and regional scenarios: Alternative scenarios for energy use and CO{sub 2} emissions until 2050 by C. Mandil and J. Podkanski (IEA), The stabilization of CO{sub 2} emissions in the coming 50 years by R. Socolow (Princeton University). Evolution of the international context: the stakes and 'factor 4' issues: Costs of climate impacts and ways towards 'factor 4' by D. Dron (ENS Mines de Paris), CO{sub 2} emissions reduction policy: the situation in the United States by D. Reiner (MIT/Cambridge University), Post-Kyoto scenarios by P. Horrocks (European Commission), Possibilities for R and D in CO{sub 2} capture and storage in the future FP7 program by P. Fernandez Ruiz and P. Dechamps (European Commission). Session II - CO{sub 2} emission reductions in the energy and transport sectors. Reducing CO{sub 2} emissions during the production and conversion of fossil energies (fixed installations): Combined cycles using hydrogen by G. Haupt (Siemens), CO{sub 2} emission reductions in the oil and gas industry by I. Wright (BP). Reducing CO{sub 2} emissions in the transport sector: Sustainable transport systems by P. Wiederkehr (EST International), The prospects for reducing CO{sub 2} emissions in the transport sector (cars and aviation) by A. Morcheoine (ADEME), The contribution of biofuels and alternative fuels to reducing CO{sub 2} emissions in the transport sector by I. Drescher (Volkswagen AG). Session III - Technological progress in the capture and geological storage of CO{sub 2}: European projects on CO{sub 2} capture and storage by P. Dechamps (European Commission, Research Energy Conversion and Transport); Capture of CO{sub 2}: Innovative CO{sub 2} capture concepts by P. Feron (TNO), Capture of CO{sub 2} in pre- and oxy-combustion by A. Feraud, N. Otter (Alstom); Geological storage of CO{sub 2}: Geological storage capacity by N.P. Christensen (GEUS), Feedback from industrial CO{sub 2} storage projects by T. Torp (Statoil), The main avenues of current research by P. Le Thiez (IFP) and I. Czernichowski (BRGM), Long-term industrial experience with underground gas storage by J. Hartman (GDF). Session IV - Regulatory, economic and financial aspects. Legal and regulatory framework for capture and geological storage: UK's perspective on the regulatory framework for CO{sub 2} storage by J. Roberts (DEFRA-UK), Monitoring and reporting of CCS in the European Union Emission Trading Scheme by P. Zakkour (ERM), Social need and public questions and perceptions by G. von Goerne (Greenpeace); Economic and financial impact: The costs of CCS by G. Allinson (CO{sub 2}-CRC), The characteristics of CO{sub 2} markets: players, volumes exchanged, and term and spot transaction prices by L. Segalen (European Carbon Fund), CO{sub 2} management by J.M. Gires (Total), The forthcoming IPCC special report on carbon dioxide capture and storage by B. Metz (IPCC Working Group III). Closing Address by Bernard Brillet, Ministry for Higher Education and Research. (J.S.)

  12. Contribution à l'étude de l'environnement industriel des systèmes de distribution: Le cas de la presse imprimée dans les pays de l'Union Européenne

    OpenAIRE

    YALLES , Ahmed

    2000-01-01

    The present thesis deals with industrialization which takes place within press distribution systems in certain countries of European Union., in particular France, Great Britain, Germany, Italy and Spain. It stresses liberalization and regulation of realization operation of titles diffusion according to mechanisms and processes borrowed from industrial logistics and marketing distributor specific to the sector of the current goods distribution. After introduction constraints and specificities ...

  13. Poweo half-year 2006 earnings. Positive net income, implementation of the 1. steps of the industrial plan; Poweo resultats du 1. semestre 2006. Resultat net positif, mise en oeuvre des 1. etapes du plan industriel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-09-15

    POWEO, the leading independent energy operator in France, presents in this document its key financial data and highlights for the first half of 2006: - Half-year revenue amounts to euro 119.4 m, multiplied by 3.4 compared to the same period last year; - The Energy Management activity has achieved a net margin of euro 34.3 m; - EBIT amounts to euro 6.2 m, compared to euro -2.9 m in the first half of 2005; - Net income amounts to euro 8.9 m, compared to euro -2.9 m in the first half of 2005; - Completion of the preliminary steps to the building of a first thermal power plant (CCGT) is close at hand, two other projects launched; - Strengthening of internal structures in view of the residential market opening up; - Outlook for 2006: total sales expected to reach euro 220 m and positive EBITDA; - LNG terminal building project in Le Havre.

  14. Les aliments industriels (hors laits et céréales) destinés aux nourrissons et enfants en bas âge : un progrès diététique ?

    OpenAIRE

    Ghisolfi, J.; Bocquet, A.; Bresson, J. L.; Briend, André; Chouraqui, J. P.; Darmaun, D.; Dupont, C.; Frelut, M. L.; Girardet, J. P.; Goulet, O.; Hankard, R.; Rieu, D.; Simeoni, U.; Turck, D.; Vidailhet, M.

    2013-01-01

    Processed baby foods designed for infants (4-12 months) and toddlers (12-36 months) (excluding infant formula, follow-on formula, the so-called growing-up milks, and cereal-based foods for infants), which am referred to as baby foods, are specific products defined by a European regulation (Directive 2006/125/CE). According to this Directive, such foods have a composition adapted to the nutritional needs of children of this age and should comply with specifications related to food safety in te...

  15. Surveillance des systèmes à événements discrets commandés: Conception et implémentation en utilisant l'automate programmable industriel

    OpenAIRE

    Allahham , Adib

    2008-01-01

    This thesis presents an approach to monitoring the controlled discrete events systems. The study is limited to the interruptible faults : intermittent and permanent. To increase the availability of a system, it is crucial to reduce the unnecessary interruptions. For that, the acceptable behavior of these systems is introduced. This behavior presents a tolerance to the intermittent faults. An approach of the construction of monitoring system is presented. We model, firstly, the behavior of the...

  16. Mise en pratique du schéma 2BSvs dans le groupe industriel Sofiprotéol : étude de cas sur toute la filière (de l’agriculteur au pétrolier

    Directory of Open Access Journals (Sweden)

    Guizouarn Kristell

    2015-01-01

    Full Text Available Pour réduire sensiblement les émissions de CO liées aux transports routiers, l’Union européenne mise sur le développement des biocarburants, en respectant des conditions de durabilité. Afin de répondre à cette obligation et d’être en mesure de démontrer la conformité du biodiesel à ces critères, depuis l’agriculteur jusqu’au distributeur pétrolier, la filière des oléagineux a engagé la rédaction d’un schéma de vérification volontaire : 2BSvs, pour Biomasse Biocarburant Schéma volontaire sur la durabilité. Reconnu par la Commission Européenne le 19 juillet 2011, le schéma 2BSvs couvre l’ensemble de la chaîne de production des biocarburants, du producteur de biomasse à l’entrée dans un entrepôt sous douanes. À travers une étude de cas sur toute la filière, de l’agriculteur au pétrolier, cet article présente les implications pour les acteurs de la filière oléagineuse, les changements intervenus depuis avril 2013, et les défis futurs à relever.

  17. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels; RPV-1: un premier reacteur virtuel pour simuler les effets d'irradiation dans les aciers de cuve des reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Jumel, St

    2005-01-15

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  18. Et arkitektonisk anliggende

    DEFF Research Database (Denmark)

    Dahl, Torben

    2004-01-01

    En satsning på industriel innovation må tage udgangspunkt i at etablere samarbejde mellem byggeriets parter, og projekteringen bør medtænke de kvalitative tilskud, som en avanceret industriel produktion tilbyder. Det kræver forståelse og udnyttelse af de teknologiske potentialer, der ligger i...... industriel produktion fra udvikling af komponenter over åbne systemer og byggeriets processer til det endelige værk....

  19. Optimization of fuel cycles: marginal loss values; Optimisation des cycles de combustibles: valeurs marginales des pertes

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J [Commissariat a l' Energie Atomique, 75 - Paris (France); Lasteyrie, B de; Doumerc, J [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, 75 - Paris (France)

    1965-07-01

    comme definitivement perdue, alors que le reste pourrait etre recupere et recycle. Le cout eleve des pertes, recyclees ou non, d'autant plus eleve que l'uranium est plus enrichi, exige qu'il en soit tenu compte dans l'optimisation generale des cycles de combustible. Il importe donc de determiner leur niveau le plus souhaitable economiquement, aux diverses etapes d'elaboration du combustible nucleaire. Mais en France et dans d'autres pays, la production de matieres fissiles est geree par l'Etat, tandis que la fabrication de l'element combustible est effectuee par l'industrie privee. Les criteres d'optimisation et l'interet economique accorde aux pertes sont donc differents pour les deux parties de la chaine de fabrication. Pour tenter neanmoins d'atteindre un optimum conforme a l'interet collectif sans intervenir dans la politique de prix de l'entreprise, on peut utiliser la propriete des couts marginaux d'etre egaux entre eux a l'optimum, pour un volume de production donne. On peut donc ajuster le niveau des pertes pour realiser cette egalite des couts marginaux dont le calcul est plus facile a obtenir de la firme que la justification des prix eux-memes. On s'apercoit d'ailleurs que, bien qu'axee essentiellement sur les pertes, cette analyse globale peut conduire a une meilleure utilisation d'autres facteurs de production. On donne un expose theorique et des exemples pratiques de cette methode d'optimisation economique dans le cadre de la fabrication d'elements combustibles destines a des reacteurs du type: uranium naturel, moderes au graphite et refroidis par le gaz carbonique. (auteurs)

  20. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

  1. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal d