Sample records for reacteurs nucleaires vver

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

    Hennion, A


    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)

  2. Nuclear reactor (1960); Reacteurs nucleaires (1960)

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


    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)

  3. Cross-correlation of two detectors in a nuclear reactor; Intercorrelation de deux detecteurs dans un reacteur nucleaire

    Dalfes, A.; Beliard, L.; Cazemajou, J.; Froelicher, B. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires


    Auto and cross-correlation functions of signals given by neutron detectors situated in a subcritical nuclear reactor are determined by a numerical method. Values of the prompt neutrons decay constant obtained by means of the autocorrelation function of each detector and the cross-correlation function of the two detectors are compared to the reference value given by a classical pulsed neutrons measurement. Agreement between results seems to be satisfactory. (authors) [French] Les fonctions d'autocorrelation et d'intercorrelation des signaux issus de deux detecteurs de neutrons places dans un reacteur nucleaire sous critique sont determinees par une methode numerique. On compare les valeurs de la constante de decroissance des neutrons prompts donnees par les fonctions d'autocorrelation de chaque detecteur et la fonction d'intercorrelation des deux detecteurs au resultat de reference fourni par une manipulation dite de 'neutrons pulses'. L'accord entre les resultats parait satisfaisant. (auteurs)

  4. 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

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


    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)

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

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


    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

  6. 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

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


    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.

  7. Improvement of Sodium Neutronic Nuclear Data for the Computation of Generation IV Reactors; Contribution a l'amelioration des donnees nucleaires neutroniques du sodium pour le calcul des reacteurs de generation IV

    Archier, P.


    The safety criteria to be met for Generation IV sodium fast reactors (SFR) require reduced and mastered uncertainties on neutronic quantities of interest. Part of these uncertainties come from nuclear data and, in the particular case of SFR, from sodium nuclear data, which show significant differences between available international libraries (JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0). The objective of this work is to improve the knowledge on sodium nuclear data for a better calculation of SFR neutronic parameters and reliable associated uncertainties. After an overview of existing {sup 23}Na data, the impact of the differences is quantified, particularly on sodium void reactivity effects, with both deterministic and stochastic neutronic codes. Results show that it is necessary to completely re-evaluate sodium nuclear data. Several developments have been made in the evaluation code Conrad, to integrate new nuclear reactions models and their associated parameters and to perform adjustments with integral measurements. Following these developments, the analysis of differential data and the experimental uncertainties propagation have been performed with Conrad. The resolved resonances range has been extended up to 2 MeV and the continuum range begins directly beyond this energy. A new {sup 23}Na evaluation and the associated multigroup covariances matrices were generated for future uncertainties calculations. The last part of this work focuses on the sodium void integral data feedback, using methods of integral data assimilation to reduce the uncertainties on sodium cross sections. This work ends with uncertainty calculations for industrial-like SFR, which show an improved prediction of their neutronic parameters with the new evaluation. (author) [French] Les criteres de surete exiges pour les reacteurs rapides au sodium de Generation IV (RNR-Na) se traduisent par la necessite d'incertitudes reduites et maitrisees sur les grandeurs neutroniques d'interet. Une part

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

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


    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

  9. The Pegase reactor loops; Les boucles du reacteur Pegase



    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)

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

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


    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.

  11. Physics of nuclear reactors; La physique des reacteurs nucleaires

    Marguet, S. [Ecole Nationale Superieure de Risques Industriels de Bourges, 18 (France); Institut de Transfert de Technologie d' EDF, 92 - Clamart (France)


    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  12. Safety of VVER-440 reactors

    Slugen, Vladimir


    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  13. Organic chemistry and radiochemistry: study of chemical interactions between iodine and paint of French nuclear reactor in a severe accident situation; Chimie organique et radiochimie. Etude des interactions chimiques iode-peinture dans un reacteur nucleaire (de type R.E.P.) en situation d'accident grave

    Aujollet, Y. [Direction Generale de la Surete Nucleaire et de la Radioprotection, 75 - Paris (France)


    In Phebus (French in pile facility; PWR scale 1/5000) experiments, performed by the Institut de Radioprotection et de Surete Nucleaire, few quantities of organic iodides were registered after interaction between iodine and reactor containment paint. This study concerns all mechanisms of chemical reactions between iodine and the polymer of the paint in order to estimate the organic iodides released from the paint. At first, all the paint components had been identified. Several models of chemical sites of the polymer were synthesized and tested with iodine in different conditions of temperature and radiation. These experiments showed interactions between iodine and secondary or tertiary amines by charge transfer. In few cases, the complex of tertiary amines creates oxidation reactions. (author)

  14. Nuclear study of Melusine; Etude nucleaire de Melusine

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


    In this report are reviewed - with respect to starting of experiments - the main nuclear characteristics of a 20 per cent enriched uranium lattice, with light water as moderator and reflector. The reactor is to operate at 1 MW. 1) Study of various critical masses. 2) Control. Effectiveness of cadmium. Control rods and of a stainless steel regulating rod. 3) Study of the effect on reactivity of disturbances in the core center. 4) Study of xenon and samarium poisoning. 5) Temperature factor. 6) Heat exchanges in a fuel element. (author) [French] On etudie, dans ce rapport, les principales proprietes nucleaires d'un reseau a uranium enrichi (20 pour cent), dont le moderateur et le reflecteur sont l'eau legere en vue des experiences de demarrage. Ce reacteur devra fonctionner a 1 MW. 1) Etude de diverses masses critiques. 2) Controle. Efficacite des barres de controle en cadmium et d'une barre de reglage en acier inoxydable. 3) Etude de l'effet sur la reactivite de perturbation au centre du coeur. 4) Etude de l'empoisonnement xenon et samarium. 5) Coefficient de temperature. 6) Echanges thermiques dans un element. (auteur)

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

    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)


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

  16. 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

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


    projets de reacteurs futurs. - Il presente un concept de reacteur a neutrons rapides de 1000 MWe, pour la realisation duquel aucune impossibilite technologique de realisation n'apparait. - Il indique que le comportement dynamique appatait satisfaisant, bien que le coefficient de reactivite isotherme du au sodium soit positif. - Il essaie de situer le developpement des reacteurs a neutrons rapides dans le cadre de l'expansion de l'energie nucleaire. L'approvisionnement en matieres fissiles ne semble pas devoir freiner en France le developpement des reacteurs a neutrons rapides. Celui - ci dependra essentiellement a moyen terme de leur competitivite economique. (auteurs)

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



    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)

  18. Accident Management in VVER-1000

    F. D'Auria


    Full Text Available The present paper deals with the investigation study on accident management in VVER-1000 reactor type conducted in the framework of a European Commission funded project. The mentioned study involved both experimental and computational fields. The purpose of this paper is to summarize the main findings from the execution of a wide-range analysis focused on AM in VVER-1000 with main regard to the qualification of computational tools and the proposal for an optimal AM strategy for this kind of NPP.

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

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


    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)

  20. 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

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


    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)

  1. Production of nuclear graphite in France; Production de graphite nucleaire en France

    Legendre, P.; Mondet, L. [Societe Pechiney, 74 - Chedde (France); Arragon, Ph.; Cornuault, P.; Gueron, J.; Hering, H. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires


    The graphite intended for the construction of the reactors is obtained by the usual process: confection of a cake from coke of oil and tar, cooked (in a electric oven) then the product of cook is graphitized, also by electric heating. The use of the air transportation and the control of conditions cooking and graphitization have permitted to increase the nuclear graphite production as well as to better control their physical and mechanical properties and to reduce to the minimum the unwanted stains. (M.B.) [French] Le graphite destine a la construction des reacteurs est obtenu par le procede usuel: confection d'une pate a partir de coke de petrole et de brai, cuisson de cette pate (au four electrique) puis graphitation du produit cuit, egalement par chauffage electrique. L'usage du transport pneumatique et le controle des conditions cuisson et de graphitation ont permit d'augmenter la production de graphite nucleaire ainsi que de mieux controler ses proprietes physiques et mecaniques et de reduire au minimum les souillures accidentelles. (M.B.)

  2. Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Lavrenchuk, O. V.


    Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

  3. Modernization of existing VVER-1000 surveillance programs

    Kochkin, V.; Erak, D.; Makhotin, D. [NRC ' Kurchatov Inst.' , 1 Kurchatov Square, Moscow 123182 (Russian Federation)


    According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillance specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more. (authors)

  4. Comparison of attenuation coefficients for VVER-440 and VVER-1000 pressure vessels

    Marek, M.; Rataj, J.; Vandlik, S. [Reactor Physics Dept., Research Centre Rez, Husinec 130, 25068 (Czech Republic)


    The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit. (authors)

  5. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    A. Del Nevo


    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  6. VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

    Duo, J. I. [Radiation Engineering and Analysis, Westinghouse Electric Company LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)


    Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

  7. ASTEC and ICARE/CATHARE modelling improvement for VVERs

    Zvonarev, Yu [Russian Research Centre ' Kurchatov Institute' (RRC KI), NRI, Kurchatov Square 1, Moscow (Russian Federation); Volchek, A., E-mail: voltchek@nsi.kiae.r [Russian Research Centre ' Kurchatov Institute' (RRC KI), NRI, Kurchatov Square 1, Moscow (Russian Federation); Kobzar, V. [Russian Research Centre ' Kurchatov Institute' (RRC KI), NRI, Kurchatov Square 1, Moscow (Russian Federation); Chatelard, P.; Van Dorsselaere, J.P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Sadarache (France)


    ASTEC and ICARE/CATHARE computer codes, developed by IRSN (France) (the former with GRS, Germany), are used in RRC KI (Russia) for the analyses of accident transients on VVER-type NPPs. The latest versions of the codes were continuously improved and validated to provide a better understanding of the main processes during hypothetical severe accidents on VVERs. This paper describes modelling improvements for VVERs carried out recently in the ICARE common part of the above codes. These actions concern the important models of fuel rod cladding mechanical behaviour and oxidation in steam at high and very high temperatures. The existing models were improved basing on the experience in the field and latest literature data sources for Zr + 1%Nb material used for manufacture of VVERs fuel rod claddings. Best-fitted correlations for the Zr alloy oxidation through a broad temperature range were established, along with recommendations on model application in clad geometry and starvation conditions. A model for the creep velocity was chosen for the clad mechanical model and some cladding burst criteria were established as a function of temperature. After verification of modelling improvements on Separate Effect Tests, validation was carried out on integral bundle tests such as QUENCH, CODEX-CT, PARAMETER-SF (the application to the CORA-VVER experiments is not described in the present paper) and on the Paks-2 cleaning tank incident. The comparison of updated code results with experimental data demonstrated very good numerical predictions, which increases the level of code applicability to VVER-type materials.

  8. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)


    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

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

    Wohleber, X


    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)

  10. Can a nuclear reactor operate for 100 years?; Un reacteur nucleaire peut-il fonctionner cent ans

    Hertel, O.


    The TWR (Travelling Wave Reactor) concept was invented in the fifties, then forgotten and it reappeared in 2001 but it was considered too immature to be selected for the fourth generation of nuclear reactors, now an American company 'Terrapower' proposes one whose design is given in the article. This TWR operates with depleted uranium, only the lower part of the fuel rod involves uranium fuel with a civil enrichment ratio (less that 20%). The lower part of the fuel will ignite the fission reaction and enrich the part of fuel just above through neutron absorption. The burning part of the fuel will move up progressively. The main advantage of this reactor is that it can operate for decades without maintenance nor fuel loading. The principle is right on the paper but requires huge technological work to select materials and systems that will be able to withstand decades of operation time in harsh conditions. (A.C.)

  11. Nuclear fuel: from the mine to the reactor. Part. 1; Les combustibles nucleaire: de la mine au reacteur (1. partie)

    Maillard, D. [Direction Generale de l' Energie et des Matieres Premieres, Direction du Gaz, de l' Electricite et du Charbon, (STEEGB), 75 - Paris (France); Dujardin, Th. [OCDE, 92 - Issy les Moulineaux (France); Capus, G.; Durante, P.; Devos, L.; Canat, J.N. [AREVA NC, 13 --Saint-Paul-lez-Durance (France); Durret, L.F. [AREVA, Unit Enrichissement (United States); Delevallee, C. [AREVA, FBFC, 26 - Romans (France); Aullo, M.; Alonsa, J. [ENUSA, Industrias Avanzadas (Spain); Chapin, D.; Bellanger, Ph.; Buechel, R. [Westinghouse Electric Company (United States); Blanc, M. [Westinghouse Electric Company (Sweden)


    This issue contains a selection of articles from the S.F.E.N. convention 2006, held in Paris, June 13 and 14, 2006, which was dedicated to nuclear fuel. the second part will be published in R.G.N. 5 2006. (author)

  12. Inception and evolution of Oklo natural nuclear reactors; Genese et evolution des reacteurs nucleaires fossiles d'Oklo

    Bentridi, Salah-Eddine [UMR 7517, laboratoire d' hydrologie et de geochimie de Strasbourg, CNRS/universite de Strasbourg, 1, rue Blessig, 67084 Strasbourg (France); Laboratoire de l' energie et des systemes intelligents, CUKM, route de Theniet, El-Hed 44225 (Algeria); Gall, Benoit [UMR 7178, institut pluridisciplinaire Hubert-Curien, CNRS-IN2P3/universite de Strasbourg, 23, rue du Loess, 67037 Strasbourg (France); Gauthier-Lafaye, Francois [UMR 7517, laboratoire d' hydrologie et de geochimie de Strasbourg, CNRS/universite de Strasbourg, 1, rue Blessig, 67084 Strasbourg (France); Seghour, Abdeslam [Centre de recherches nucleaires d' Alger - CRNA, 2, boulevard Frantz-Fanon, 16000 Alger (Algeria); Medjadi, Djamel-Eddine [Ecole normale superieure, Vieux-Kouba, 16050 Alger (Algeria)


    The occurrence of more than 15 natural nuclear Reactor Zones (RZ) in a geological environment remains a mystery even 40 years after their discovery. The present work gives for the first time an explanation of the chemical and physical processes that caused the start-up of the fission reactions with two opposite processes, uranium enrichments and progressive impoverishment in {sup 235}U. Based on Monte-Carlo neutronics simulations, a solution space was defined taking into account realistic combinations of relevant parameters acting on geological conditions and neutron transport physics. This study explains criticality occurrence, operation, expansion and end of life conditions of Oklo natural nuclear reactors, from the smallest to the biggest ones. (authors)

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

    Wohleber, X


    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)

  14. KARATE - a code for VVER-440 core calculation

    Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.


    A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

  15. Positron Annihilation Studies of VVER Type Reactor Steels

    Brauer, G.


    A summary of recent positron annihilation work on Russian VVER type reactor steels is presented. Thereby, special attention is paid to the outline of basic processes that might help to understand the positron behaviour in this class of industrial material. The idea of positron trapping by irradiation-induced precipitates, which are probably carbides, is discussed in detail.

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

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


    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

  17. Siloette, Siloe mock-up; Siloette, modele nucleaire de siloe

    Delcroix, V.; Jeanne, G.; Mitault, G.; Schulhof, P. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires


    Siloette is the Siloe mock-up. The main installations are described: various tanks, building, auxiliaries, control systems... Precis ions are given about precautions taken for using spent fuel elements. (authors) [French] Siloette est le modele nucleaire de SILOE. On decrit ses diverses installations: bassins, batiments, auxiliaires, controle... Des precisions sont donnees sur les precautions prises pour y utiliser des elements uses. (auteurs)

  18. The nuclear power stations of the French atomic energy programme (1960); Les centrales nucleaires de puissance du programme francais (1960)

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


    After recalling the entry of nuclear energy into energy production in France, the paper emphasizes the evolution of techniques applied in the designing of French nuclear power plants and describes the means employed for reducing costs per kWh of EDF2 and EDF3 compared with EDF1: the electric power per ton of uranium varies from 493 kW/t for EDF1 to 970 kW/t for EDF3. For this purpose the thermal power and electric power of units are changed respectively from 290 MWt for EDF1 to 1200 or 1600 MWt for EDF3 and from 28 to 250 MW. The results are obtained by an improvement in neutronic characteristics, developments in nuclear fuel technology, and simplification of the system of charging the reactor, whose means of maintenance are increased; the EDF2 heat-exchangers have been so designed as to increase the unit power of the elements, which will attain 9 MWt, as against 3 for EDF1. For EDF3 an advance project forecasts a thermodynamic layout with only one pressure stage. The paper ends with a description of the burst-slug detection systems, and an appendix gives a detailed comparative table of EDF1, EDF2 and EDF3 plant characteristics. (author) [French] Apres avoir rappele l'integration de l'energie nucleaire parmi les moyens de production de l'energie en France, les auteurs se penchent surtout sur l'evolution des techniques appliquees dans l'equipement des centrales nucleaires francaises et decrivent les moyens mis en oeuvre pour reduire les prix de revient du kWh d'EDF2 et d'EDF3 par rapport a EDF1: la puissance electrique par tonne d'uranium varie de 493 kW/t pour EDF1 a 970 kW/t pour EDF3. C'est dans ce but que les puissances thermiques et la puissance unitaire des groupes turbo-alternateurs passent respectivement de 290 MWt pour EDF1 a 1200 ou 1600 MWt pour EDF3 et de 82 a 250 MW. Les resultats sont obtenus par une amelioration des caracteristiques neutroniques, des progres realises sur la technologie des elements

  19. Preparation macroconstants to simulate the core of VVER-1000 reactor

    Seleznev, V. Y.


    Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

  20. Thermal ageing mechanisms of VVER-1000 reactor pressure vessel steels

    Shtrombakh, Yaroslav I.; Gurovich, Boris A.; Kuleshova, Evgenia A.; Maltsev, Dmitry A.; Fedotova, Svetlana V.; Chernobaeva, Anna A.


    In this paper a complex of microstructural studies (TEM and SEM) and a comparative analysis of the results of these studies with the data of mechanical tests of temperature sets of VVER-1000 RPV surveillance specimens with exposure times up to ∼200,000 h were conducted. Special annealing of control and temperature sets of SS which provides the dissolution of grain boundary segregation was performed to clarify the mechanisms of thermal ageing. It was demonstrated that during long-term exposures up to 200,000 h at the operating temperature of about 310-320 °C thermal ageing effects reveal themselves only for the weld metal (Ni content ⩾ 1.35%) and are the result of grain boundary segregation accumulation (development of reversible temper brittleness). The obtained results improve the accuracy of prediction of the thermal ageing rate of VVER-1000 materials in case of RPV service life extension up to 60 years.

  1. Methodological studies on the VVER-440 control assembly calculations

    Hordosy, G.; Kereszturi, A.; Maraczy, C. [KFKI Atomic Energy Research Institute, Budapest (Hungary)


    The control assembly regions of VVER-440 reactors are represented by 2-group albedo matrices in the global calculations of the KARATE code system. Some methodological aspects of calculating albedo matrices with the COLA transport code are presented. Illustrations are given how these matrices depend on the relevant parameters describing the boron steel and steel regions of the control assemblies. The calculation of the response matrix for a node consisting of two parts filled with different materials is discussed.

  2. Coolability of ballooned VVER bundles with pellet relocation

    Hozer, Z.; Nagy, I.; Windberg, P.; Vimi, A. [AEKI, 49, Budapest, H-1525 (Hungary)


    During a LOCA incident the high pressure in the fuel rods can lead to clad ballooning and the debris of fuel pellets can fill the enlarged volume. The evaluation of the role of these two effects on the coolability of VVER type fuel bundles was the main objective of the experimental series. The tests were carried out in the modified configuration of the CODEX facility. 19-rod electrically heated VVER type bundle was used. The test section was heated up to 600 deg. C in steam atmosphere and the bundle was quenched from the bottom by cold water. Three series of tests were performed: 1. Reference bundle with fuel rods without ballooning, with uniform power profile. 2. Bundle with 86% blockage rate and with uniform power profile. The blockage rate was reached by superimposing hollow sleeves on all 19 fuel rods. 3. Bundle with 86% blockage rate and with local power peak in the ballooned area. The local power peak was produced by the local reduction the cross section of the internal heater bar inside of the fuel rods. In all three bundle configurations three different cooling water flow-rates were applied. The experimental results confirmed that a VVER bundle with even 86% blockage rate remains coolable after a LOCA event. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. Earlier tests on the coolability of ballooned bundles were performed only with Western type bundles with square fuel lattice. The present test series was the first confirmation of the coolability of VVER type bundles with triangular lattice. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front. The first tests indicated that the effect of local power peak was less significant on the delay of cooling down than the effect of ballooning. (authors)

  3. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.


    The possibility of using UO2-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO2-BeO fuel pellets are estimated.

  4. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)


    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  5. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)


    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

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

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


    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)

  7. Moessbauer study of EUROFER and VVER steel reactor materials

    Kuzmann, E., E-mail: [Eoetvoes University, Laboratory of Nuclear Chemistry, Institute of Chemistry (Hungary); Horvath, A. [Hungarian Academy of Sciences, Centre for Energy Research (Hungary); Alves, L.; Silva, J. F.; Gomes, U.; Souza, C. [Universidade Federal do Rio Grande do Norte (University) (Brazil); Homonnay, Z. [Eoetvoes University, Laboratory of Nuclear Chemistry, Institute of Chemistry (Hungary)


    {sup 57}Fe Moessbauer spectroscopy and X-ray diffractometry were used to study EUROFER or VVER ferritic reactor steels mechanically alloyed with TaC or NbC. Significant changes were found in the Moessbauer spectra and in the corresponding hyperfine field distributions between the ball milled pure steel and that alloyed with TaC or NbC. Spectral differences were also found in the case of use of same carbides with different origin, too. The observed spectral changes as an effect of ball milling of the reactor material steels with carbides can be associated with change in short range order of the constituents of steel.

  8. Fine structure behaviour of VVER-1000 RPV materials under irradiation

    Gurovich, B. A.; Kuleshova, E. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Chernobaeva, A. A.; Zabusov, O. O.


    Changes in the fine structure and mechanical properties of the base metal (BM) and weld metal (WM) of VVER-1000 pressure vessels during accumulation of neutron dose in the range of fluences ˜(3.2-15) × 10 23 m -2 ( E > 0.5 MeV) at 290 °C are studied using methods of transmission electron microscopy, fractographic analysis, and Auger electron spectroscopy. A correlation was found between the changes of mechanical properties and the micro- and nano-structures of the studied steels. Accumulation of neutron dose considerably raises the strength characteristics and transition temperature of VVER-1000 pressure vessel steels. The rate of changes in the mechanical properties of the weld metal is significantly higher than that of the base metal. The slower growth of strength characteristics and transition temperature shift of the base metal under irradiation as compared with the weld metal is due to the slower growth of the density of radiation defects and radiation-induced precipitates. The level of intergranular embrittlement under irradiation in the weld metal is not higher then in the base metal in spite of the higher content of nickel.

  9. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    Morozov, A. V.; Remizov, O. V.


    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  10. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.


    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  11. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)


    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

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

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


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

  13. Results of Post Irradiation Examinations of VVER Leaky Rods

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)


    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  14. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Bobrov Evgenii


    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  15. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors; Recherche sur l`energie nucleaire dans les domaines du cycle du combustible des reacteurs a eau legere et des reacteurs a neutrons rapides

    Barre, B.; Camarcat, N.


    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs.

  16. VVER Knowledge Preservation and Transfer within the Frame of CORONA Project Activities

    Mitev Mladen


    Full Text Available The CORONA project is funded by the European Commission under the FP7 programme with overall objective to establish a Regional Centre of Competence for VVER Technology and Nuclear Applications. The Centre will provide support and services for preservation and transfer of VVER-related nuclear knowledge as well as know-how and capacity building. Specific training schemes aimed at nuclear professionals and researchers, non-nuclear professionals and students are developed and implemented in cooperation with local, national and international training and educational institutions. Pilot trainings are executed for each specific target group to assess the applicability of the training materials. The training scheme implemented for nuclear professionals and researchers is focussed on the NPP Lifetime Management. The available knowledge on enhancing safety and performance of nuclear installations with VVER technology is used in the preparation of the training materials. The Online Multimedia Training Course on VVER Reactor Pressure Vessel Embrittlement and Integrity Assessment, developed by the joint effort of JRC-IET and IAEA is used in the training. The outcome collected from the trainees showed that the tool meets its primary goal of consolidating the existing knowledge on the VVER RPV Embrittlement and Integrity Assessment, provides adequate ground for transfer of this knowledge.

  17. VVER Knowledge Preservation and Transfer within the Frame of CORONA Project Activities

    Mitev, Mladen; Corniani, Enrico; Manolova, Maria; Pironkov, Lybomir; Tsvetkov, Iskren


    The CORONA project is funded by the European Commission under the FP7 programme with overall objective to establish a Regional Centre of Competence for VVER Technology and Nuclear Applications. The Centre will provide support and services for preservation and transfer of VVER-related nuclear knowledge as well as know-how and capacity building. Specific training schemes aimed at nuclear professionals and researchers, non-nuclear professionals and students are developed and implemented in cooperation with local, national and international training and educational institutions. Pilot trainings are executed for each specific target group to assess the applicability of the training materials. The training scheme implemented for nuclear professionals and researchers is focussed on the NPP Lifetime Management. The available knowledge on enhancing safety and performance of nuclear installations with VVER technology is used in the preparation of the training materials. The Online Multimedia Training Course on VVER Reactor Pressure Vessel Embrittlement and Integrity Assessment, developed by the joint effort of JRC-IET and IAEA is used in the training. The outcome collected from the trainees showed that the tool meets its primary goal of consolidating the existing knowledge on the VVER RPV Embrittlement and Integrity Assessment, provides adequate ground for transfer of this knowledge.

  18. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Gencheva, Rositsa Veselinova, E-mail:; Stefanova, Antoaneta Emilova, E-mail:; Groudev, Pavlin Petkov, E-mail:


    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  19. Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Zabusov, O.; Prikhodko, K.; Zhurko, D.


    The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E > 0.5 MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds - radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475 °C during 100 h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher - 565 °C during 100 h - to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.

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

    Jacquet, P.


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

    Perdu, F


    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)

  2. On needs for particular isotopes with future advanced nuclear reactors; Quelques besoins en isotopes particuliers pour les reacteurs nucleaires avances du futur

    Asou, M.; Porta, J. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Reacteurs


    This paper is concerned with two essential points with innovative approaches for nuclear reactors: utilization of {sup 15}N in uranium nitride fuels in order to avoid carbon 14 generation; utilization of rare earths, and especially gadolinium, for the control of the potential reactivity in the core in the case of a cycle lengthening from 12 to 18 months, and in the case of a water reactor operating without soluble boron. 1 tab., 17 refs.

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

    Dehaut, Ph


    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)

  4. Experimental reactor regulation: the nuclear safety authority's approach; Le controle des reacteurs experimentaux: la demarche de l'Autorite de surete nucleaire

    Rieu, J.; Conte, D.; Chevalier, A. [Autorite de Surete Nucleaire, 75 - Paris (France)


    French research reactors can be classified into 6 categories: 1) critical scale models (Eole, Minerve and Masurca) whose purpose is the study of the neutron production through the fission reaction; 2) reactors that produce neutron beams (Orphee, and the high flux reactor in Grenoble); 3) reactors devoted to safety studies (Cabri, Phebus) whose purpose is to reproduce accidental configurations of power reactors in reduced scale; 4) experimental reactors (Osiris, Phenix) whose purpose is the carrying-out of irradiation experiments concerning nuclear fuels or structure materials; 5) teaching reactors (Ulysse, Isis); and 6) reactors involved in defense programs (Caliban, Prospero, Apareillage-B). We have to note that 3 research reactors are currently being dismantled: Strasbourg University's reactor, Siloe and Siloette. Research reactors in France are of different types and present different hazards. Even if methods of control become more and more similar to those of power reactors, the French Nuclear Safety Authority (ASN) works to allow the necessary flexibility in the ever changing research reactor field while ensuring a high level of safety. Adopting the internal authorizations for operations of minor safety significance, under certain conditions, is one example of this approach. Another challenge in the coming years for ASN is to monitor the ageing of the French research reactors. This includes periodic safety reviews for each facility every ten years. But ASN has also to regulate the new research reactor projects such as Jules Horowitz Reactor, International Thermonuclear Experimental Reactor, which are about to be built.

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



    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.)

  6. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre; Defaillance electrique sur le reacteur n. 3 de la centrale nucleaire de Dampierre



    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

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

    Verdier, A


    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)

  8. 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

    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


    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.)

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

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


    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.

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

    Cognet, G


    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.

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

    Lelievre, F


    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)

  12. Etude de l'influence du champ magnetique dans une section d'essais thermohydraulique d'un canal de reacteur nucleaire CANDU 6

    Landry-Lavoie, Renaud

    This memoir deals with the effects of the magnetic fields present in a thermal hydraulic test section of the Canadian nuclear industry. This test section is used to determine the thermal hydraulic conditions that can lead to critical heat flux in a channel of a CANDU 6 nuclear reactor. To perform their series of experiments the STERN Company used strong electric currents to heat the simulation bundles with a thermal power similar to the one found in a channel of a CANDU reactor. The materials constituting the simulation channel and its supports are of ferromagnetic nature. The strong magnetic field generated by the bundles implies that they are subjected to a magnetostatic force due to the magnetization of the ferromagnetic materials. The nuclear industry wants to know if these efforts, combined with the force of gravity, are sufficient to maintain the bundles in place in the simulation channel. The question also arises whether or not the magnetic field present in the channel can affect the parameters of boiling heat transfer. To determine the magnetic field distribution in the simulation channel, we had recourse to the magnetostatic image method and the integral method of calculation of magnetization. The results of the calculations show that the magnetostatic forces exerted by the ferromagnetic elements of the test section are inferior in magnitude to the one estimated by the STERN laboratorie. We used the mechanistic model of Sullivan et al. (1964) to evaluate the possible influence of the magnetic fields on the departure diameter of the vapor bubbles. The deviation in the frequency of bubble emission was evaluated by using the correlations of Zuber et al. (1959) and Cole (1960). By introducing a magnetostatic force in the boiling model and in the correlations, we demonstrated that the magnetic field present in the STERN test section has a negligible effect on the bubble departure diameter and their emission frequency. We conclude that the conditions in the test section are consistent with those prevailing in the channel of a CANDU 6 reactor.

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



    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.)

  14. Functional approaches: a new view of nuclear reactors management; Les approches fonctionnelles: une nouvelle vision de la conduite des reacteurs nucleaires

    Papin, B


    Since many years a research program on the future reactors command, has been decided by the CEA, in collaboration with EDF and Framatome. This program aims to enhance the reactor safety by a better control of the installation in any exploitation situations. The paper presents the state of the art and the first reflexions. (A.L.B.)

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

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


    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.)

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

    Dadoun, O


    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)

  17. 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

    Dardour, S


    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.)

  18. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Voloschenko Andrey


    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  19. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Voloschenko, Andrey; Zaritskiy, Sergey; Egorov, Aleksander; Boyarinov, Viktor


    The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations) is demonstrated on several calculation and experimental tests.

  20. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors


    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  1. Exposure conditions of reactor internals of Rovno VVER-440 NPP units 1 and 2

    Grytsenko, O.V.; Pugach, S.M.; Diemokhin, V.L.; Bukanov, V.N. [Inst. for Nuclear Research, Kyiv, 03680 (Ukraine); Marek, M.; Vandlik, S. [Nuclear Research Inst. Rez Plc., Rez, 25068 (Czech Republic)


    Results of determination of irradiation conditions for vessel internals of VVER-440 reactor No. 1 and 2 at Rovno Nuclear Power Plant, obtained by specialists at Inst. for Nuclear Research Kyiv (Ukraine)), and Nuclear Research Inst. Rez (Czech Republic)), are presented. To calculate neutron transport, detailed calculation models of these reactors were prepared. Distribution of neutron flux functionals on the surface of reactor VVER-440 baffle and core barrel for different core loads was studied. Agreement between results obtained by specialists at Inst. for Nuclear Research and at Nuclear Research Inst. is shown. (authors)

  2. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)


    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  3. Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle

    Emmett, M.B.


    Standard problems were created to test the ability of American and Russian computational methods and data regarding the analysis of the storage and handling of Russian pressurized water reactor (VVER) mixed oxide fuel. Criticality safety and radiation shielding problems were analyzed. Analysis of American and Russian multiplication factors for fresh fuel storage for low-enriched uranium (UOX), weapons- (MOX-W) and reactor-grade (MOX-R) MOX differ by less than 2% for all variations of water density. For shielding calculations for fresh fuel, the ORNL results for the neutron source differ from the Russian results by less than 1% for UOX and MOX-R and by approximately 3% for MOX-W. For shielding calculations for fresh fuel assemblies, neutron dose rates at the surface of the assemblies differ from the Russian results by 5% to 9%; the level of agreement for gamma dose varies depending on the type of fuel, with UOX differing by the largest amount. The use of different gamma group structures and instantaneous versus asymptotic decay assumptions also complicate the comparison. For the calculation of dose rates from spent fuel in a shipping cask, the neutron source for UOX after 3-year cooling is within 1% and for MOX-W within 5% of one of the Russian results while the MOX-R difference is the largest at over 10%. These studies are a portion of the documentation required by the Russian nuclear regulatory authority, GAN, in order to certify Russian programs and data as being acceptably accurate for the analysis of mixed oxide fuels.

  4. Chemical composition effect on VVER-1000 RPV weld metal thermal aging

    Gurovich, B. A.; Chernobaeva, A. A.; Erak, D. Yu; Kuleshova, E. A.; Zhurko, D. A.; Papina, V. B.; Skundin, M. A.; Maltsev, D. A.


    Temperature and fast neutron flux simultaneously affect the material of welded joints of reactor pressure vessels under irradiation. Understanding thermal aging effects on the weld metal allows for an explanation of the mechanisms that govern an increase in the ductile-to-brittle transition temperature of the reactor pressure vessel materials under long term irradiation at operation temperature. This paper reports on new results and reassessment of the VVER-1000 weld metal surveillance specimen database performed at the National Research Center "Kurchatov Institute". The current database of VVER-1000 weld metal thermal aging at 310-320 °C includes 50 transition temperature values with the maximum holding time of 208,896 h. The updated database completed with the information on intergranular fracture shear and phosphorous content in the grain boundaries has allowed us to propose a new mechanism of VVER-1000 weld materials thermal aging at 310-320 °C and develop models of ductile-to-brittle transition temperature shift for VVER-1000 weld metal during a long-term exposure at 310-320 °C.

  5. Conformity Between LR0 Mock-Ups and Vvers Npp Rpv Neutron Flux Attenuation

    Belousov, Sergey; Ilieva, Krassimira; Kirilova, Desislava


    The conformity of the mock-up results and those for reactor pressure vessel (RPV) of nuclear power plants (NPP) has been evaluated in order to qualify if the mock-ups data could be used for benchmark's purpose only, or/and for simulating of the NPP irradiation conditions. Neutron transport through the vessel has been calculated by the three-dimensional discrete ordinate code TORT with problem oriented multigroup energy neutron cross-section library BGL. Neutron flux/fluence and spectrum shape represented by normalized group neutron fluxes in the multigroup energy structure, for neutrons with energy above 0.5 MeV, have been used for conformity analysis. It has been demonstrated that the relative difference of the attenuation factor as well as the group neutron fluxes did not exceed 10% at all considered positions for VVER-440. For VVER-1000, it has been obtained the same consistency, except for the location behind the RPV. The neutron flux attenuation behind the RPV is 18% higher than the mock-up attenuation. It has been shown that this difference arises from the dissimilarity of the biological shielding. The obtained results have demonstrated that the VVERs' mock-ups are appropriate for simulating the NPP irradiation conditions. The mock-up results for VVER-1000 have to be applied more carefully i.e. taking into account the existing peculiarity of the biological shielding and RPV attenuation azimuthal dependence.

  6. The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

    Ruscak, M.; Chvatal, P.; Karnik, D.


    One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

  7. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)


    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  8. Modelisation de la synthese reactive de poudres ultrafines dans un reacteur a plasma thermique

    Desilets, Martin

    La presente these s'inscrit dans le cadre de la modelisation mathematique des ecoulements a plasmas thermiques inertes et reactifs. Elle vise plus precisement a combler les lacunes des modeles existants en portant une attention particuliere aux phenomenes de transport multicomposant et a la prediction des transformations chimiques. Pour repondre a ces attentes et ainsi poursuivre le developpement dans ce domaine, un modele global a ete developpe. Il combine la resolution d'equations conservatives pour la masse, l'energie et le momentum. La generation d'un plasma inductif (h.f ) y est traitee au moyen d'equations representant les champs electromagnetiques. La nucleation et la croissance de poudres ultrafines sont incluses dans le modele via l'analyse des principaux moments de la distribution des tailles de particules. Enfin, tous les phenomenes physico-chimiques d'importance dans un milieu comme les plasmas thermiques, de meme que lem interactions, sont consideres. Le modele est applique ici a l'analyse de trois problematiques differentes et complementaires. La premiere concerne l'etude du melange gazeux d'un jet froid (He, N 2 ou O2), injecte au coeur d'une decharge d'argon/hydrogene ou d'argon/oxygene. La comparaison des predictions du modele avec des mesures experimentales obtenues par une sonde enthalpique permet une validation partielle de ce dernier. La deuxieme problematique a trait a l'etude numerique de la pyrolyse du methane en reacteur a plasma h.f. Elle met en evidence les difficultes de convergence de la methode numerique lorsque appliquee a la resolution d'ecoulements reactifs a haute temperature. Finalement, le dernier sujet aborde dans cette these, soit l'analyse systematique des principales conditions d'operation d'un reacteur h.f utilise pour la synthese reactive de poudres ultrafines de silicium, engage tous les elements theoriques du modele. Il implique en effet la decomposition thermique d'un precurseur gazeux, le tetrachlorure de silicium, la

  9. Assessment of computer codes for VVER-440/213-type nuclear power plants

    Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)


    Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

  10. Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

    Usheva, K. I.; Kuten, S. A.; Khruschinsky, A. A.; Babichev, L. F.


    A physical model of the radial and axial reflector of VVER-1200-like reactor core has been developed. Five types of radial reflector with different material composition exist for the VVER reactor core and 1D and 2D models were developed for all of them. Axial top and bottom reflectors are described by the 1D model. A two-group XS library for diffusion code DYN3D has been generated for all types of reflectors by using Serpent 2 Monte Carlo code. Power distribution in the reactor core calculated in DYN3D is flattened in the core central region to more extent in the 2D model of the radial reflector than in its 1D model.

  11. Evolution of microstructure and mechanical properties of VVER-1000 RPV steels under re-irradiation

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Erak, D.; Zhurko, D.


    This is a comprehensive study of microstructure and mechanical properties evolution at re-irradiation after recovery annealing of VVER-1000 RPV weld and base metals as well as the effect of annealing on the microstructure and properties of base metal in the zone of the temperature gradient that is implemented during annealing using special heating device. It is shown that the level of radiation-induced microstructural changes under accelerated re-irradiation of weld and base metal is not higher than for the primary irradiation. Thus, we can predict that re-embrittlement of VVER-1000 RPV materials considering the flux effect will not exceed the typical embrittlement rate for the primary irradiation.

  12. Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

    Saadatian-derakhshandeh Farahnaz


    Full Text Available One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

  13. 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

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


    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}{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}{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

  14. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J. [Nuclear Res. Inst., Rez (Czech Republic)


    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  15. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    Kostov, M.K. [Bulgarian Academy of Sciences, Sofia (BG). Central Lab. for Seismic Mechanics and Earthquake Engineering; Ma, D.C. [Argonne National Lab., IL (United States); Prato, C.A. [Univ. of Cordoba (AR); Stevenson, J.D. [Stevenson and Associates, Cleveland, OH (US)


    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made.

  16. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    Linge, I. I.; Mitenkova, E. F.; Novikov, N. V.


    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  17. 3D analysis of the reactivity insertion accident in VVER-1000

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)


    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  18. Embrittlement of low copper VVER 440 surveillance samples neutron-irradiated to high fluences

    Miller, M. K.; Russell, K. F.; Kocik, J.; Keilova, E.


    An atom probe tomography microstructural characterization of low copper (0.06 at.% Cu) surveillance samples from a VVER 440 reactor has revealed manganese and silicon segregation to dislocations and other ultrafine features in neutron-irradiated base and weld materials (fluences 1×10 25 m-2 and 5×10 24 m-2, E>0.5 MeV, respectively). The results indicate that there is an additional mechanism of embrittlement during neutron irradiation that manifests itself at high fluences.

  19. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján


    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  20. Thread Inspection Manipulator for Primary Loop Components of VVER 1000/1200 Nuclear Power Plants

    Rušev, Marko


    HRID developed special manipulator for inspection of different size of threads (M36, M48, M52, M60, M64, M100) on nuclear power plant (VVER 1000/1200) components with eddy current and ultrasonic methods. Manipulator is extremely easy to use reducing personnel time in radiation zone significantly. 95% of all assembling and disassembling activities can be performed manually without use of any tool. It allows quick inspection of threads with both methods in fully automatic mode.

  1. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Fekete, Balazs, E-mail: [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)


    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  2. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.


    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  3. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Zabusov, O.


    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  4. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Maltsev, D. A.; Fedotova, S. V.; Frolov, A. S.; Zhuchkov, G. M.


    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

  5. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Alekseev, P. N.; Bobrov, E. A., E-mail:; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)


    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  6. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Savander, V. I.; Shumskiy, B. E.; Pinegin, A. A.


    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  7. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.


    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  8. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.


    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  9. Conformity of nuclear construction codes with the requirements of the French order dated December 12, 2005 related to nuclear pressure equipment; Conformite des codes de construction nucleaires avec les exigences de l'arrete du 12 decembre 2005 relatif aux equipements sous pression nucleaires

    Grandemange, J.M.; Renaut, P. [Areva-NP, Tour AREVA, 92084 - Paris La Defense cedex, (France); Paris, D. [EDF-Ceidre 2 rue Ampere - 93206 SAINT-DENIS Cedex (France); Faidy, C. [EDF-Septen 12/14, Avenue Dutrievoz 69628 Villeurbanne Cedex (France)


    The French Decree dated December 13, 1999 transposing the Pressure Equipment Directive (PED) has replaced the fundamental texts on which up to now the regulation for pressure equipment important for the safety of nuclear reactors was also founded. By a Ministerial Order - called 'ESPN Order' - dated December 12, 2005, a new regulation has been issued for nuclear pressure equipment. This text makes reference to the Decree transposing the PED while completing these provisions by supplementary requirements having the objective to provide a very high level of integrity guarantee for equipments which are the most important for safety, and to cover the prevention of radioactive release risks. These regulatory evolutions are presented in the Plenary Session of the ESOPE conference. Referencing the Decree and thus the PED, and including specific provisions, the Ministerial Order implies that the Manufacturers update their documents and, if necessary, their prescriptions in the following two domains: - that of the conformity of Codes and Standards used, generally inspired from the ASME Code Section III, with the essential safety requirements of the PED, - that of the respect of the complementary provisions brought by the ESPN Order. This paper presents the more significant conclusions of this work and the resulting amendments of the RCC-M Code, introduced by the 2007 addendum to that Code. The analysis will lead to specify the same type of complementary requirements to Code when a manufacturer wishes to use the German KTA Rules or the ASME Code Section III. (authors) [French] Le decret du 13 decembre 1999 transposant la directive europeenne (DESP) relative aux equipements sous pression a remplace les textes fondamentaux sur lesquels se fondait egalement jusque la la reglementation des appareils a pression importants pour la surete des reacteurs nucleaires. Par arrete - dit 'arrete ESPN' - du 12 decembre 2005, une nouvelle reglementation a ete dictee. Ce

  10. Nuclear program of Iran. Towards de-escalation of a nuclear crisis. Advisory letter; Nucleair programma van Iran. Naar de-escalatie van een nucleaire crisis. Briefadvies



    The Dutch government, partly at the request of the House of Representatives (Second Chamber), the AIV asked to give an opinion about the position of Iran in the region and the role of the nuclear program of Iran in the geopolitical relations, in view of the most recent developments [Dutch] De Nederlandse regering heeft, mede op verzoek van de Tweede Kamer der Staten-Generaal, de AIV gevraagd advies uit te brengen over de positie van Iran in de regio en de rol van het nucleaire programma van Iran in de geopolitieke verhoudingen hierin, mede gelet op de meest recente ontwikkelingen.

  11. The differential characteristics of control rods of VVER-1000 core simulator at a low number of axial mesh points

    Bolsunov, A. A.; Karpov, S. A.


    An algorithm for refining the differential characteristics of the control rods (CRs) of the control and protection system (CPS) for a neutronics model of the VVER-1000 simulator at a low number of axial mesh points of the core is described. The problem of determining the constants for a cell with a partially inserted CR is solved. The cell constants obtained using the proposed approach ensure smoothing of the differential characteristics of an absorbing rod. The algorithm was used in the VVER-1000 simulators (Bushehr NPP, unit no. 1; Rostov NPP, unit no. 1; and Balakovo NPP, unit no. 4).

  12. Uncertainty-accounted calculational-experimental approach for improved conservative evaluations of VVER RPV radiation loading parameters

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N. [Scientific and Engineering Centre for Nuclear and Radiation Safety SEC NRS, Building 5, Malaya Krasnoselskaya Street, 2/8, 107140 Moscow (Russian Federation)


    The approach of improved uncertainty-accounted conservative evaluation of vodo-vodyanoi energetichesky reactor (VVER) (reactor-) pressure-vessel (RPV) radiation loading parameters has been proposed. This approach is based on the calculational-experimental procedure, which takes into account C/E ratio, depending on over- or underestimation, and uncertainties of measured and calculated results. An application of elaborated approach to the full-scale ex-vessel neutron dosimetry experiments on Russian VVERs combined with neutron-transport calculations has been demonstrated in the paper. (authors)

  13. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)


    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  14. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Töre, Candan; Ortego, Pedro


    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  15. Rod ejection simulation on VVER 1000/320 core using PARCS/TRACE

    Ruscak, Marek


    The rod ejection (RE) is a design basis accident in accordance with NUREG-0800 and usually studied using point kinetics. In this thesis a 3D kinetic model is prepared and coupled with a thermal hydraulic system code for simulating this accident scenario for general VVER 1000 technology. This topic has been defined by the Research Centre Rez of the Czech Republic as a part of a larger project concerning beyond design basis accident focused on the Station Black Out (SBO) and a Lo...

  16. Effect of Ni content on thermal and radiation resistance of VVER RPV steel

    Shtrombakh, Ya. I.; Gurovich, B. A.; Kuleshova, E. A.; Frolov, A. S.; Fedotova, S. V.; Zhurko, D. A.; Krikun, E. V.


    In this paper thermal stability and radiation resistance of VVER-type RPV steels for pressure vessels of advanced reactors with different nickel content were studied. A complex of microstructural studies and mechanical tests of the steels in different states (after long thermal exposures, provoking embrittling heat treatment and accelerated neutron irradiation) was carried out. It is shown that nickel content (other things being equal) determines the extent of materials degradation under influence of operational factors: steels with a lower nickel concentration demonstrate a higher thermal stability and radiation resistance.

  17. Mössbauer study of EUROFER and VVER steel reactor materials

    Kuzmann, E.; Horváth, Á.; Alves, L.; Silva, J. F.; Gomes, U.; Souza, C.; Homonnay, Z.


    57Fe Mössbauer spectroscopy and X-ray diffractometry were used to study EUROFER or VVER ferritic reactor steels mechanically alloyed with TaC or NbC. Significant changes were found in the Mössbauer spectra and in the corresponding hyperfine field distributions between the ball milled pure steel and that alloyed with TaC or NbC. Spectral differences were also found in the case of use of same carbides with different origin, too. The observed spectral changes as an effect of ball milling of the reactor material steels with carbides can be associated with change in short range order of the constituents of steel.

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

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


    - 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)

  19. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V


    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  20. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Frybort, J.


    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  1. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)


    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  2. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)


    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  3. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.


    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  4. Development and validation of calculation schemes dedicated to the interpretation of small reactivity effects for nuclear data improvement; Developpement et validation de schemas de calcul dedies a l'interpretation des mesures par oscillation pour l'amelioration des donnees nucleaires

    Gruel, A.


    Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here {sup 241}Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study. (author) [French] Les mesures de reactivite par la technique d'oscillation, comme celles effectuees dans le reacteur Minerve, permettent de tester de nombreux parametres neutroniques sur des materiaux, des combustibles ou des isotopes specifiques. Generalement, les effets attendus sont tres faibles, tout au plus de l'ordre de la dizaine de pcm. La modelisation de ces experiences doit donc etre particulierement precise, afin d'obtenir un retour fiable et precis sur les parametres cibles. En particulier, les biais de calcul doivent etre clairement identifies, quantifies et maitrises

  5. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)


    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  6. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Jardine, L J


    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  7. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    d’Auria Francesco


    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  8. CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model

    F. Moretti


    Full Text Available A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of EC TACIS project R2.02/02: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration at core inlet.” Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle, and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for nuclear reactor safety. Both a pretest and a posttest CFD simulations of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, resp.. The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

  9. Qualification of the APOLLO2 lattice physics code of the NURISP platform for VVER hexagonal lattices

    Hegyi, Gyoergy; Kereszturi, Andras; Tota, Adam [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.


    The experiments performed at the ZR-6 zero power critical reactor by the Temporary International Collective (TIC) and a burnup benchmark specified for depletion calculation of a VVER-440 assembly containing Gd burnable poison were used to qualify the APOLLO2.8-3.E (APOLLO2) code as a part of its ongoing validation activity. The work is part of the NURISP project, where KFKI AEKI undertook to develop and qualify some calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell, macro-cell and 2D calculations of selected regular and perturbed experiments are used for the validation. In the 2D cases, the radial leakage is also taken into account by the axial leakage represented by the measured axial buckling. Criticality parameter and reaction rate comparisons are presented. Although various sets of the experiments have been selected for the validation, good agreement of the measured and calculated parameters could be found by using the various options offered by APOLLO2. An additional mathematical benchmark - presented in the paper - also attests for the reliability of APOLLO2. All the test results prove the reliability of APOLLO2 for VVER core calculations. (orig.)

  10. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Fekete, Balazs; Trampus, Peter


    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  11. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science


    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  12. Test case for VVER-1000 complex modeling using MCU and ATHLET

    Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.


    The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

  13. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Jardine, L J


    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  14. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)


    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

  15. An experimental substantiation of the design functions imposed on the additional system for passively flooding the core of a VVER reactor

    Morozov, A. V.; Remizov, O. V.


    Results obtained from a research work on experimentally substantiating the serviceability of the additional system for passively flooding the core of a VVER reactor from the second-stage hydro accumulators are presented.

  16. Development and application of the coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER for safety analysis

    Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)


    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)

  17. Comprehensive survey of the Russian nuclear industry; Le panorama nucleaire russe



    This document presents the organization of nuclear activities in the Russian federation: Minatom and its replacement by the federal agency of atomic energy, personnel, nuclear power plants (VVER, RBMK, fast neutron and mixed reactors), availability and power production, export of activities (construction of nuclear power plants in Slovakia, Iran, China, India, project in Viet Nam), expansion of the nuclear power plants park (improvement of plants safety, increase of service life), completion of uncompleted plants, the construction of which was stopped after the Chernobyl accident and the reorganization of the former-USSR, construction of new generation power plants (VVER-640, -1000 and -1500), fuel cycle facilities (geographical distribution, production of natural uranium, conversion and enrichment), fuel fabrication, reprocessing processes and spent fuel storage, management of radioactive wastes (leasing), R and D activities (organizations and institutes), research programs of the international scientific and technical center, nuclear safety authority (Gosatomnadzor - GAN). (J.S.)

  18. A spatial kinetic model for simulating VVER-1000 start-up transient

    Kashi, Samira [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki, E-mail: [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, Majid [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)


    Research highlights: > A spatial kinetic model of a VVER-1000 reactor core is presented. > The reactor power is tracked using the point kinetic equations from 100 W to 612 kW. > The lamped parameter approximation is used for solving the energy balance equations. > The value of reactivity related to feedback effects of core elements is calculated. > The main neutronic parameters during the transient are calculated. - Abstract: An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means 'Reactor spatial

  19. Analytic index for nuclear physicians uses; Repertoire analytique a l'usage des physiciens nucleaires

    Ballini, R.; Barloutaud, R.; Bernas, R.; Bretonneau, P.; Chaminade, R.; Cohen, R.; Conjeaud, M.; Cotton, E.; Faraggi, H.; Grjebine, T.; Joffre, H.; Laboulaye, H. de; Lesueur, C.; Leveque, A.; Moreau, J.; Naggiar, V.; Papineau, L.; Prugne, P.; Schuhl, C.; Studinowski, FJ.; Netter, F.; Raievski, V.; Valladas, G. [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Marty, N.; Renard, G. [College de France, Lab. de Chimie Nucleaire (France)

    The problem of the documentation in nuclear physics becomes constantly more complex. Every week brings its share of new publications, always more numerous and more varied. To remedy to this facts that we tried, in the service of Nuclear Physics of the CEA, to give to the documentation a character of a collective and systematized work. The present report covers the literature appeared between first January 1950 and first July 1951. (Volume 1: CEA report number 120; Volume 2: CEA report number 184). (M.B.) [French] Le probleme de la documentation en physique nucleaire devient sans cesse plus complexe. Chaque semaine apporte son lot de publications, toujours plus nombreuses et plus diversifiees. C'est pour essayer de porter remede a cet etat de choses que nous avons essaye, au service de Physique Nucleaire du C.E.A., de donner a la documentation le caractere d'un travail collectif systematise. Le present rapport couvre la litterature parue entre le premier Janvier 1950 et le premier Juillet 1951. (Tome 1: Rapport CEA numero 120; Tome 2: Rapport CEA numero 184). (M.B.)

  20. Associative memories in nuclear physics; Les memoires associatives en physique nucleaire

    Blanca, E.; Carriere, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires


    Experiments in nuclear physics involve the use of large size 'memories'. After showing the difficulties arising from the use of such memories, the authors give the principles of the various programming methods which make it possible to operate the memories associatively thus benefiting from a reduction in size and better operational conditions. They attempt to estimate the shape and dimensions of an associative memory with cable connections which could be designed specially for nuclear research, contrary to those actually in service. (authors) [French] Les experiences de physique nucleaire necessitent l'emploi de 'memoires' de grandes dimensions. Apres avoir montre les inconvenients que presente l'utilisation de telles memoires, les auteurs exposent les principes des diverses methodes de programmation qui permettent d'assurer un fonctionnement des memoires sur le mode associatif donc une reduction de leurs dimensions et un meilleur usage. Ils tentent d'evaluer le format d'une memoire associative cablee qui, contrairement a celles qui existent actuellement, serait prevue specialement pour l'experimentation nucleaire. (auteurs)

  1. Nuclear biological studies in France; Les etudes de biologie nucleaires en France

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


    On the occasion of a colloquium on radiobiological research programmes, a number of documents dealing with French accomplishments and projects in this field were collected together. We felt that it would be useful to assemble these papers in one report; although they are brief and leave gaps to be filled in, they provide certain data, give an overall view of the situation, and can also suggest a rough plan for the general policy to adopt in the field of 'nuclear' biological research; i.e. research based on the nuclear tracer method or devoted to the action of ionising radiations. (author) [French] Un colloque sur les programmes de recherche en radiobiologie nous a donne l'occasion de reunir des documents sur les realisations et les projets francais dans ce domaine. Il nous a semble utile de reunir en un rapport l'ensemble de ces documents, qui, malgre leur brievete et malgre les lacunes qu'ils comportent, donnent un certain nombre d'informations, permettent une vue d'ensemble et peuvent dessiner aussi l'ebauche d'une politique coherente en matiere de recherches biologiques 'nucleaires', c'est-a-dire de recherches basees sur la methode des indicateurs nucleaires ou consacrees a l'action des rayonnements ionisants. (auteur)

  2. Technology of repair of selected equipment in the power plant type VVER 440

    Barborka, J.; Magula, V. [Welding Research Inst. (WRI), Bratislava (Slovakia)


    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored.

  3. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.


    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  4. Application of spectral tuning on the dynamic model of the reactor VVER 1000 support cylinder

    Musil A.


    Full Text Available The paper deals with the optimization of parameters of the dynamic model of the reactor VVER 1000 support cylinder. Within the model of the whole reactor, support cylinder appears to be a significant subsystem for its modal properties having dominant influence on the behaviour of the reactor as a whole. Relative sensitivities of eigenfrequencies to a change of the discrete parameters of the model were determined. Obtained values were applied in the following spectral tuning process of the (selected discrete parameters. Since the past calculations have shown that spectral tuning by the changes of mass parameters is not effective, the presented paper demonstrates what results are achieved when the set of the tuning parameters is extended by the geometric parameters. Tuning itself is then formulated as an optimization problem with inequalities.

  5. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Mohammed Saad Dwiddar


    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  6. Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

    V.I. Belozerov


    Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR. It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.

  7. Fracture mechanical investigation of a thermo shock scenario for a VVER-440 RPV

    Altstadt, E.; Abendroth, Martin [Forschungszentrum Dresden-Rossendorf (Germany)


    The paper describes the modelling and evaluation of a pressurized thermal shock (PTS) scenario in a VVER-440 reactor pressure vessel due to an emergency cooling. An axially oriented semi-elliptical crack is assumed to be located in the core welding seam. Two variants of fracture mechanical evaluation are performed: the analysis of a sub-cladding crack and of a surface crack. Three-dimensional finite element (FE) models are used to compute the global transient temperature and stress-strain fields. By using a three-dimensional submodel, which includes the crack, the local crack stress-strain field is obtained. Within the subsequent postprocessing using the J-integral technique the stress intensity factors K{sub I} along the crack front are obtained. The FE results are compared to analytical calculations proposed in the VERLIFE code. The stress intensity factors are compared to the fracture toughness curve of the weld material. (orig.)

  8. CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

    I. Spasov


    Full Text Available The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT and include asymmetric vessel flow transients and main steam line break (MSLB transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

  9. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.


    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  10. Neutron Dosimetry in Edf Experimental Surveillance Programme for VVER-440 Nuclear Power Plants

    Brumovsky, Milan; Erben, Oldrich; Zerola, Ladislav; Hogel, Josef; Massoud, Jean-Paul; Trollat, Christophe


    Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. For the absolute fluence values evaluation account was taken of the time history of the reactor power and of local changes of the neutron flux along the reactor core height, and of correction factors due to the orientation of monitors with respect to the reactor core centre. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis at the axial positions of the sample centres and fluence values in the geometric centre of the samples were calculated making use the exponential attenuation model of the incident neutron beam.


    Tymur Foshch


    Full Text Available This study represents the improved mathematical and imitational allocated in space multi-zone model of VVER-1000 which differs from the known one. It allows to take into account the energy release of 235U nuclei fission as well as 239Pu . Moreover, this model includes sub-models of simultaneous control impact of the boric acid concentration in the coolant of the first circuit and the position of 9th group control rods which allows to consider it as the model with allocated parameters and also allows to monitor changes in the mentioned technological parameters by reactor core symmetry sectors, by layers of reactor core height and by fuel assembly group each symmetry sector. Moreover, this model allows to calculate important process-dependent parameters of the reactor (including axial offset as quantitative measure of its safety. As the mathematical and imitational models were improved, it allows to take into account intrinsic properties of the reactor core (including transient processes of xenon and thus reduce the error of modelling static and dynamic properties of the reactor.The automated control method of power change of the NPP unit with VVER-1000 was proposed for the first time. It uses three control loops. One of which maintains the regulatory change of reactor power by regulating the concentration of boric acid in the coolant, the second circuit keeps the required value of axial offset by changing the position of control rods, and the third one holds constant the coolant temperature mode by regulating the position of the main turbo generator valves.On the basis of the above obtained method, two control programs were improved. The first one is the improved control program that implements the constant temperature of the coolant in the first circuit and the second one is the improved control program that implements the constant steam pressure in the second circuit.

  12. Sequence of decommissioning of the main equipment in a central type VVER 440 V-230; Secuencia de desmantelamiento de los equipos principales de una central Tipo VVer 440 V-230

    Andres, E.; Garcia Ruiz, R.


    IBERDROLA Ingenieria y Construccion S.A.U., leader of consortium with Empresarios Agrupados and INDRA, has developed the Basic Engineering for the decommissioning of contaminated systems and building of a VVER 440 V-230 Nuclear Power Plant, establishing the sequence and methodology for the main equipment fragmentation. For that, it has been designed dry and wet cutting zones to be set up in the area where steam generators, main cooling pumps and pressurizer are located; these components will be dismantled previously. (Author)

  13. Evolution of structure and properties of VVER-1000 RPV steels under accelerated irradiation up to beyond design fluences

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Maltsev, D.; Frolov, A.; Zabusov, O.; Erak, D.; Zhurko, D.


    In this paper comprehensive studies of structure and properties of VVER-1000 RPV steels after the accelerated irradiation to fluences corresponding to extended lifetime up to 60 years or more as well as comparative studies of materials irradiated with different fluxes were carried out. The significant flux effect is confirmed for the weld metal (nickel concentration ⩾1.35%) which is mainly due to development of reversible temper brittleness. The rate of radiation embrittlement of VVER-1000 RPV steels under operation up to 60 years and more (based on the results of accelerated irradiation considering flux effect for weld metal) is expected not to differ significantly from the observed rate under irradiation within surveillance specimens.

  14. Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core

    Gorodkov, S.S.; Kalugin, M.A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)


    Up to now core calculations with Monte Carlo provided only average cross-sections of mesh cells for further use either in finite difference calculations or as benchmark ones for approximate spectral algorithms. Now MCU code is capable to handle functions, which may be interpreted as average diffusion coefficients. Subsequently the results of finite difference calculations with cells characteristic sets obtained in such a way can be compared with Monte Carlo results as benchmarks, giving reliable information on quality of production code under consideration. As an example of such analysis, the results of mesh calculations with 1-, 2-, 4-, 8- and 12 neutron groups of some model VVER fuel assembly are presented in comparison with the exact Monte Carlo solution. As a second example, an analysis is presented of water gap approximate enlargement between fuel assemblies, allowing VVER core region be covered by regular mesh.

  15. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)


    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  16. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Panferov Pavel


    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  17. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey


    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  18. Development of a cross-section methodology and a real-time core model for VVER-1000 simulator application

    Georgieva, Emiliya Lyudmilova


    The novel academic contributions are summarized as follows. A) A cross-section modelling methodology and a cycle-specific cross-section update procedure are developed to meet fidelity requirements applicable to a cycle-specific reactor core simulation, as well as particular customer needs and practices supporting VVER-1000 operation and safety. B) A real-time version of the Nodal Expansion Method code is developed and implemented into Kozloduy 6 full-scope replica control room simulator.

  19. SANS response of VVER440-type weld material after neutron irradiation, post-irradiation annealing and reirradiation

    Ulbricht, Andreas; Bergner, Frank; Boehmert, Juergen; Valo, Matti; Mathon, Marie-Helene; Heinemann, Andre


    Abstract It is well accepted that the reirradiation behaviour of reactor pressure vessel (RPV) steel after annealing can be different from the original irradiation behaviour. We present the first small-angle neutron scattering (SANS) study of neutron irradiated, annealed and reirradiated VVER440-type RPV weld material. The SANS results are analysed both in terms of the size distribution of irradiation-induced defect/solute atom clusters and in terms of the ratio of total and nuclea...

  20. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    V. Sánchez


    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  1. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K. [Russian Research Center ' Kurchatov Institute' , 1., Kurchatov sq., 123182 Moscow (Russian Federation)


    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of FAs of state-of-the-art designs. This technique is employed in the VVER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

  2. The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

    Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.


    The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

  3. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Gy. Ézsöl


    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  4. The French civilian nuclear: connections and stakes; Le nucleaire civil francais: filieres et enjeux



    This document (18 power point slides) gives an overview of the French civilian nuclear industry and research and development: importance of the nuclear power generation in France, excellence of the education in nuclear sciences, organization of the nuclear connection (CEA, Areva, EDF, IRSN), the role of the French International Nuclear Agency (AFNI), the requirements for a renewal of human resources (French and foreign engineers) in the field of nuclear energy, the degree course for a diploma, examples of engineer and university diplomas, the educational networks in various regions of France, presentation of the Institut National des Sciences et Techniques Nucleaires (Nuclear Sciences and Techniques National Institute) and its master degrees, organization of the French education system in nuclear sciences with strong relations with the research and development programs

  5. The nuclear law: safety. 2006-2010; Le droit nucleaire: la surete 2006-2010

    Bringuier, P. [Montpellier-1 Univ., UMR 5815, 34 (France)


    The author discusses the legal evolutions related to nuclear safety between 2006 and 2010. He identifies three main topics of unequal importance. Firstly, he comments the implementation of an international reference framework which has been completed at the European level and which aims at the harmonization of safety and security rules. Secondly, he comments the creation of the French Nuclear Safety Authority (ASN, Autorite de Surete Nucleaire). Thirdly, he comments the recast of the standard framework in order to update the French law with respect to the international reference framework. This leaded to a new distribution of power and authority, to more complete and constraining procedures, and to the definition of procedures for each step of an installation life cycle

  6. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey


    A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  7. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Košťál Michal


    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  8. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)


    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  9. French research in the field of nuclear agronomy; Les recherches francaises en agronomie nucleaire

    Guerin De Montgareuil, P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires


    , industrial firms, university laboratories scientific institutes. The role of the Commissariat a l'Energie Atomique is defined: on the one hand it supplies information and support, and on the other hand it takes charge of specifically nuclear aspects of the work. Its part in the field has recently found expression in the creation, within the Biology Department, of a Radio-agronomy Section; its objective are described,, as well as the, means placed att its disposal at the Centre d'etudes Nucleaires, Cadarache. (author) [French] On propose un bilan des travaux les plus significatifs effectues en France depuis la deuxieme conference internationale en matiere d'agronomie nucleaire et qui vont d'une recherche apparemment desinteressee a l'application la plus directe. Une telle differenciation recouvre de moins en moins, au fur et a mesure de l'evolution des programmes, la distinction qui est faite dans l'expose entre l'action biologique des rayonnements et les autres emplois des techniques nucleaires. C'est ainsi que les recherches do radiogenetique agricole sont poursuivies dans deux directions: d'un point de vue theorique et methodologique avec l'etude comparative de l'action des divers types de rayonnements, l'influence du debit de dose et de la temperature, l'action des agents mutagenes chimiques, la production de chimeres sous irradiation gamma; et d'autre part, sous un aspect pratique aboutissant a la creation de varietes nouvelles plus resistantes ou plus precoces (riz, mil, arachide). Les problemes de destruction des insectes (eradication) et de conservation des denrees sous irradiation se trouvent egalement abordes par des voies et avec des objectifs tres divers. A la demarche globale representee par une irradiation pure et simple (grains humides, pommes de terre...) sont parfois associees des etudes souvent originales, d'ordre biochimique ou microbiologique (par exemple: alteration de l

  10. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Journeau, Ch


    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  11. Stress corrosion cracking in the vessel closure head penetrations of French PWR`s; Fissuration par corrosion sous contrainte de penetrations de couvercle de cuve de reacteur nucleaire francais a eau pressurisee

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.


    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR`s in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR`s are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs.

  12. Material effects on multiphase phenomena in late phases of severe accidents of nuclear reactors; Effets des materiaux sur les phenomenes multiphasiques se produisant lors des phases avancees d'accident grave de reacteur nucleaire

    Seiler, J.M.; Froment, K


    This paper reviews and presents work carried out in the French Atomic Energy Commission (CEA) on the subject of nuclear severe accidents, i.e. those which are accompanied by melting of the nuclear core material. The emphasis is on the (crucial) thermodynamic and material behaviour of corium melts in the solidus-liquidus temperature interval, which is linked to the thermal hydraulic description. A global model approach is proposed. The work is presented in the context of the overall international effort in the area. (authors)

  13. Marine dispersion of radioactive elements susceptible to be released by the reactors of the Kursk damaged nuclear submarine; Dispersion marine des elements radioactifs susceptibles d'etre liberes par les reacteurs du sous-marin nucleaire accidente Koursk

    Calmet, D. [CEA/Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, IPSN, 92 (France); Lepicard, S. [Centre d' Etude sur l' Evaluation de la Protection dans le Domaine Nucleaire, 92 - Fontenay-aux-Roses (France)


    The Kursk nuclear submarine has been damaged on the 12. august 2000. It is on the bottom of the Barents sea. The loss of watertightness of the two nuclear reactors, that contain some hundred of kilograms of fuel (enriched uranium) would lead to the release of radioactive elements. This report specifies the general conditions of circulation of water mass susceptible to be concerned by an eventual radioactive contamination. evaluates the times of water transit that could be contaminated and then evaluates the activities contributions susceptible to be added with time to the sea waters and sea products from the French coasts of Atlantic and Channel. (N.C.)

  14. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems; Congres sur les reacteurs a sels fondus (RSF) pyrochimie et cycles des combustibles nucleaires du futur

    Brossard, Ph. [GEDEON, Groupement de Recherche CEA CNRS EDF FRAMATOME (France); Garzenne, C.; Mouney, H. [and others


    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and precipitation processes); cold salt: potentiality and preliminary results; TOPIC: redox control of MSR fuel (MSR: nominal operating conditions for the reprocessing process and redox control); technical aspects of R and D of some advanced non-aqueous reprocessing technologies for MSR systems (promising innovative separation and partitioning processes for the MSR fuel cycle); nominal operating conditions for MSR reprocessing process - data base needed and experiments for reprocessing validation; corrosion and materials for MSR and for pyro-chemistry processes; MSR reactor physics - dynamic behaviour; what safety principles for MSR? (MSR and integrated cycle (IFR) safety approach); experimental programmes in the frame of the SPHINX project of MS transmuter (programme of irradiated probes BLANKA, experimental facilities (MSR)); ISTC 1606 status - experimental study of molten salt technology for safe, low-waste and proliferation resistant treatment of radioactive waste and plutonium in accelerator-driven and critical systems. (J.S.)

  15. Jules Horowitz reactor - Complementary safety assessment in the light of the Fukushima accident; Reacteur Jules Horowitz - Evaluation complementaire de la surete au regard de l'accident survenu a la centrale nucleaire de Fukushima



    This CSA (Complementary Safety Assessment) analyses the robustness of the Jules Horowitz reactor (RJH) to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. RJH is being built on the Cadarache CEA's site. Robustness is the ability for the facility to withstand events beyond the level for which the facility was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence (cliff edge effect). Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like RJH's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. This report is divided into 9 main chapters: 1) main features of the RJH facility, 2) identification of cliff edge risks and of equipment essential for safety, 3) earthquake risk, 4) flood risk, 5) risks due to other extreme natural disasters, 6) the loss of electrical power supplies and of cooling systems, 7) management of severe accidents, 8) subcontracting policy, 9) synthesis and list of improvements. This study shows a globally good robustness of the RJH for the considered risks. Nevertheless it can considered relevant to increase the robustness of the plant on a few points: -) to increase the seismic safety margins of some pieces of equipment, -) to increase the robustness of the internal electrical power supplies, -) to increase the fuel cooling capacity, and -) to improve the management of the post-accidental period. (A.C.)

  16. Exploitation continuation of Fessenheim nuclear plant nr 1 reactor after thirty years of operation; Poursuite d'exploitation du reacteur n.1 de la centrale nucleaire de fessenheim apres trente annees de fonctionnement



    After having recalled the regulatory framework, this report indicates how the Fukushima accident has been taken into account by the French nuclear safety authority (ASN) for the decision of keeping on operating the Fessenheim nuclear plant. Then, after a general presentation of nuclear installations, the report describes some peculiarities of the Fessenheim power plant with respect to the other French nuclear plants. It comments and discusses various issues: reactor exploitation, fuel management, vessel exploitation, exploitation of the main secondary circuits, of the confinement enclosure, and of other equipment. It recalls significant events, exploitation rules, and modifications brought to the reactor. It gives a global assessment. The authors report the safety re-examination (approach, compliance examination, security re-assessment), controls performed during decennial inspection (main controls and tests, implementation of modifications foreseen by safety re-examination, significant events, monitoring by the ASN, reactor restarting after the third decennial inspection). Perspectives are then discussed for the ten following years in terms of maintenance policy, ageing management, reactor vessel serviceability, and additional actions within the frame of ageing management. The operation continuation is then discussed

  17. The EPR in a few words: all you need to know about the EPR nuclear reactor; L'EPR en bref: ce qu'il faut savoir sur le reacteur nucleaire EPR



    After a brief presentation of the EPR (European - or Evolutionary - Pressurized Reactor) type nuclear reactor, this paper, proposed by the collective group 'Stop EPR', develops the following points: EPR is as dangerous as other reactors; EPR flouts democracy; France's energy demand do not need the construction of EPRs; the construction of EPRs is not a factor of economical and social development; EPR should not be constructed neither in France nor elsewhere and the present building sites should be cancelled; the EPR will not help France to increase its energy independence and protect itself from oil price increases; choosing the EPR is incompatible with the large investments to be made in energy conservation and renewable energies; the EPR is not a solution to climate change; the VHV line corridor that will starts at Flamanville is not justified and poses risks to the environment and public health

  18. Thermal radiation modeling inside a degraded reactor core in presence of steam and water droplets; Modelisation du rayonnement thermique dans un coeur de reacteur nucleaire degrade en presence de vapeur et de gouttes d'eau

    Chahlafi, Miloud


    This work aims at modelling thermal radiation in a nuclear reactor, in the course of a severe accident leading to its degradation. Because the reactor coolant is water, radiative heat transfer occurs in presence of steam and water droplets. The 3D geometry of a fuel bundle with 21 damaged rods has been characterized from {gamma}-tomography images. The degradation of the rods has been simulated in the experimental small-scale facility PHEBUS. The homogenized radiative properties of the considered configurations with a transparent fluid phase have been completely characterized by both the extinction cumulated distribution function G{sub ext} and the scattering phase functions p. G{sub ext} strongly differs from the exponential function associated with the Beer law and p strongly depends on both the incidence and the scattering directions. By using the radiative transfer equation generalized for non Beerian porous media by Taine et al. the radiative conductivity tensor has been first determined with a transparent fluid phase, by a numerical perturbation method. Only the diagonal radial and axial components of this tensor are not equal to zero. They have been fitted by a simple law only depending on the porosity, the specific area and the wall absorptivity. In a second step, a radiative transfer equation based on three temperatures is established. This model takes into account a semi transparent fluid phase by coupling the radiative properties of fluid and solid phases. The radiative properties of water steam and droplets are calculated respectively with the CK approach and Mie theory, in typical thermal hydraulics conditions of reactor accidents. The radiative fluxes verify the Fourier law and are characterized by radiative coupled conductivity tensors associated with the temperatures of each phase. The radiative powers exchanged between phases per unit volume are also calculated from this model. (author)

  19. Contribution to development of SPNDs for instantaneous and selective measurement of different radiation fields in nuclear reactors; Contribution au developpement de collectrons pour la mesure instantanee et selective des differents champs de rayonnements en reacteurs nucleaires

    Blandin, Christophe [Institut National Polytechnique, 38 - Grenoble (France)


    The objective of this work was conceiving and experimentally optimizing the SPNDs (Self-Powdered Neutron Detector) able to control fast power transients in test reactors and also to cope with requirements of surveillanceand protection of EDF reactors. Thus, different SPND emitters of platinum, gadolinium, hafnium and cobalt were provided according to their nature with sheathing and stainless steel plugs as well as with zirconium over-sheathing in order to render them faster, more selective and adapted for wear checking. Special experimental devices were designed for measuring inside the Siloe reactor the promptness of the signals from SPND, on one hand, and their sensitivity to thermal and epithermal neutrons as well as to gamma rays, on the other hand. The follow-up of power transients in test reactors is ensured by the instantaneous measurement of thermal and epithermal neutron flux as well as of gamma field by means of three special SPND with gadolinium, hafnium and platinum. Also, we have defined the characteristics of a new SPND with cobalt, that delivers a current of unique neutronic origin, able to ensure the surveillance and protection of a power reactor over a period of at least six years.

  20. Modeling of the thermal transfer inside a porous environment: application to nuclear reactors in accident situation; Modelisation du transfert thermique dans un milieu poreux: application aux reacteurs nucleaires en situation accidentelle

    Rubiolo, P.R


    The purpose of this report is to simulate heat exchanges occurring by conduction, by convection and by radiating in a porous medium made up of opaque particles in a semi-transparent fluid. Usually the determination of the macroscopic equations is based on homogenization techniques, but in the case of a major accident, the complexity of the problem is so overwhelming that semi-empirical methods are used to determine macroscopic coefficients. The author develops a new method to determine these coefficients, this method is based on the calculation of different tensors: the equivalent conductivity tensor, the radiative conductivity tensor, the thermal conductivity tensor and the heat exchange coefficient (h{sub sf}) between the solid phase and the fluid one. The first chapter briefly describes energy, impulse and mass balances. In the case of the energy balance the solid phase is not supposed to be in thermal equilibrium with the liquid phase. The second chapter presents an application of the porous media method to a one-dimensional and stationary problem, this application to a simple problem gives an idea of the performance of the method. The model allowing the calculation of h{sub sf} is developed, it is a wide range model. The second chapter ends with the presentation of the model allowing the computing of the effective conductivity of fuel rods. A comparison between results given by this new method and other numeric calculations or experimental data coming from benchmarks is presented in the third chapter. This chapter ends with the simulation of a reactor core in accidental situation, 2 cases are presented: with and without the presence of water steam. (A.C.)

  1. Earthquake resistance of the PTR and ASG reservoirs of reactors 1 and 2 of the Fessenheim nuclear power plant; Resistance au seisme des reservoirs PTR et ASG des reacteurs 1 et 2 de la centrale nucleaire de Fessenheim



    This decision from the French authority of nuclear safety (ASN) concerns the improvement of the earthquake resistance of the PTR and ASG water tanks of the Fessenheim power plant. The PTR tank (refueling water storage tank) is used to fill up the reactor pool during the loading and unloading of the fuel and it ensures the cooling of the core during some accidental conditions by supplying the safety injection and spray systems. The ASG tank supplies the steam generators in case of failure of the normal water supply systems. A two-times restoration of the tanks has been considered as acceptable: a resistance to a maximum historically probable earthquake must be warranted by July 31, 2001, while a resistance to dimensioning earthquakes must be warranted by November 30, 2001. (J.S.)

  2. Modeling of delayed strains of concrete under biaxial loadings. Application to the reactor containment of nuclear power plants; Modelisation des deformations differees du beton sous sollicitations biaxiales. application aux enceintes de confinement de batiments reacteurs des centrales nucleaires

    Benboudjema, F


    The prediction of delayed strains is of crucial importance for durability and long-term serviceability of concrete structures (bridges, containment vessels of nuclear power plants, etc.). Indeed, creep and shrinkage cause cracking, losses of pre-stress and redistribution of stresses, and also, rarely, the ruin of the structure. The objective of this work is to develop numerical tools, able to predict the long-term behavior of concrete structures. Thus, a new hydro mechanical model is developed, including the description of drying, shrinkage, creep and cracking phenomena for concrete as a non-saturated porous medium. The modeling of drying shrinkage is based on an unified approach of creep and shrinkage. Basic and drying creep models are based on relevant chemo-physical mechanisms, which occur at different scales of the cement paste. The basic creep is explicitly related to the micro-diffusion of the adsorbed water between inter-hydrates and intra-hydrates and the capillary pores, and the sliding of the C-S-H gel at the nano-porosity level. The drying creep is induced by the micro-diffusion of the adsorbed water at different scales of the porosity, under the simultaneous effects of drying and mechanical loadings. Drying shrinkage is, therefore, assumed to result from the elastic and delayed response of the solid skeleton, submitted to both capillary and disjoining pressures. Furthermore, the cracking behavior of concrete is described by an orthotropic elastoplastic damage model. The coupling between all these phenomena is performed by using effective stresses which account for both external applied stresses and pore pressures. This model has been incorporated into a finite element code. The analysis of the long-term behavior is also performed on concrete specimens and prestressed concrete structures submitted to simultaneous drying and mechanical loadings. (author)

  3. Modelling turbulent fluid flows in nuclear and fossil-fired power plants; La modelisation des ecoulements turbulents rencontres dans les reacteurs nucleaires et dans les centrales thermiques a flamme

    Viollet, P.L.


    The turbulent flows encountered in nuclear reactor thermal hydraulic studies or fossil-fired plant thermo-aerodynamic analyses feature widely varying characteristics, frequently entailing heat transfers and two-phase flows so that modelling these phenomena tends more and more to involve coupling between several branches of engineering. Multi-scale geometries are often encountered, with complex wall shapes, such as a PWR vessel, a reactor coolant pump impeller or a circulating fluidized bed combustion chamber. When it comes to validating physical models of these flows, the analytical process highlights the main descriptive parameters of local flow conditions: tensor characterizing the turbulence anisotropy, characteristic time scales for turbulent flow particle dynamics. Cooperative procedures implemented between national or international working parties can accelerate validation by sharing and exchanging results obtained by the various organizations involved. With this principle accepted, we still have to validate the products themselves, i.e. the software used for the studies. In this context, the ESTET, ASTRID and N3S codes have been subjected to a battery of test cases covering their respective fields of application. These test cases are re-run for each new version, so that the sets of test cases systematically benefit from the gradually upgraded functionalities of the codes. (author). refs., 3 figs., 6 tabs.

  4. Neutronic modelling of the reflector for the calculation of pressurized water reactors: application to EPR; Modelisation neutronique du reflecteur pour le calcul des coeurs des reacteurs nucleaires a eau pressurisee: application a l'EPR

    Sandrin, Ch.


    This PhD Thesis aims to achieve a method for the modelling of the reflector surrounding the core for neutronics core calculations. This method should consider the EPR reactor specificities (steel reflector) and the increased demand in precision. In neutronics core calculations, the reflector can be represented either by albedos boundary conditions (current ratios) or by one or several media, surrounding the core, characterised by homogenized parameters. Those parameters (cross sections and diffusion coefficients) should be obtained using equivalence so that they allow a good reproduction of the reference albedos in a representative situation. During this PhD, such an equivalence method has been developed in the APOLLO-2 code with the minimization of a functional of the differences between the reference albedos and those computed with the equivalent parameters. Because of the positiveness constraints, a local minimization, such as Newton-like methods, is not always possible and we have therefore also implemented a Particle Swarm Optimization Algorithm for more than two energy groups' problems. The parameters obtained have been used in two dimensions EPR core calculations with the CRONOS-2 code for various fuel loadings in two to eight groups diffusion. Those core calculation have been validated against reference Monte-Carlo calculations and against core calculations with albedos boundary conditions. In addition to the increased easiness of utilization, the implemented equivalence method has yielded an improvement of the results for the two groups calculation. With a higher energy groups number, the use of a unique equivalent reflector does not account correctly for the two dimensions effects; a modelling with different reflector meshes has improved the results. The modelling of the reflector by two dimensions albedos boundary conditions is the more suited for the representation of the boundary conditions and, therefore, should the two dimensions albedos calculation be developed in a deterministic code, it is the advocated method in the future. (author)

  5. Study of water radiolysis in relation with the primary cooling circuit of pressurized water reactors; Etude sur la radiolyse de l`eau en relation avec le circuit primaire de refroidissement des reacteurs nucleaires a eau sous pression

    Pastina, B


    This memorandum shows a fundamental study on the water radiolysis in relation with the cooling primary circuit of PWR type reactors. The water of the primary circuit contains boric acid a soluble neutronic poison and also hydrogen that has for role to inhibit the water decomposition under radiation effect. In the aim to better understand the mechanism of dissolved hydrogen action and to evaluate the impact of several parameters on this mechanism, aqueous solutions with boric acid and hydrogen have been irradiated in a experimental nuclear reactor, at 30, 100 and 200 Celsius degrees. It has been found that, with hydrogen, the water decomposition under irradiation is a threshold phenomenon in function of the ratio between the radiation flux `1` B(n, )`7 Li and the gamma flux. When this ratio become too high, the number of radicals is not sufficient to participate at the chain reaction, and then water is decomposed in O{sub 2} and H{sub 2}O{sub 2} in a irreversible way. The temperature has a beneficial part on this mechanism. The iron ion and the copper ion favour the water decomposition. (N.C.). 83 refs.

  6. Economical aspects of multiple plutonium and uranium recycling in VVER reactors

    Alekseev, P.N.; Bobrov, E.A.; Dudnikov, A.A.; Teplov, P.S. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)


    The basic strategy of Russian Nuclear Energy development is the formation of the closed fuel cycle based on fast breeder and thermal reactors, as well as the solution of problems of spent nuclear fuel accumulation and availability of resources. Three options of multiple Pu and U recycling in VVER reactors are considered in this work. Comparison of MOX and REMIX fuel recycling approaches for the closed fuel cycle involving thermal reactors is presented. REMIX fuel is supposed to be fabricated from non-separated mixture of uranium and plutonium obtained in spent fuel reprocessing with further makeup by enriched U. These options make it possible to recycle several times the total amount of Pu and U obtained from spent fuel. The main difference is the full or partial fuel loading of the core by assemblies with recycled Pu. The third option presents the concept of heterogeneous arrangement of fuel pins made of enriched uranium and MOX in one fuel assembly. It should be noted that fabrication of all fuel assemblies with Pu requires the use of expensive manufacturing technology. These three options of core loading can be balanced with respect to maximum Pu and U involvement in the fuel cycle. Various physical and economical aspects of Pu and U multiple recycling for selected options are considered in this work.

  7. Generation and Testing of XS Libraries for VVER Using APOLLO2 and TRIPOLI4

    Zheleva, Nonka; Petrov, Nikolay; Todorova, Galina; Kolev, Nikola


    MOC based calculation schemes with APOLLO2 were used to generate few-group cross-section libraries for VVER-1000 at the nodal and pin level. This paper presents an overview of the testing of the schemes and the libraries, as well as the computational aspects. Two major ameliorations are considered: application of new developments in APOLLO2 and multicore computation for an acceptable trade-off between accuracy and efficiency. Two-level Pij-MOC industrial calculation schemes were tested against TRIPOLI4 reference results. Benchmarking of the schemes shows that the higher-order linear surface method of characteristics (LS MOC) is an efficient option for cross-section library generation. There is a significant potential for further refinement of the MOC energy mesh and the MOC parameters with the progress in distributed computing. A multi-parameter cross-section library for MSLB analysis with homogenized nodes was tested in 2D core simulation with COBAYA3 vs. whole-core TRIPOLI4 solutions on the CEA CCRT HPC system. Pin-by-pin cross-sections and interface discontinuity factors of Black Box Homogenization type were tested in diffusion calculations with COBAYA3 pin-by-pin against transport reference solutions. Good agreement is displayed.

  8. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  9. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch


    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  10. Simulation of the VVER-Type bundle experiment QUENCH-12 with ATHLET-CD

    Bratfisch, Christian; Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics (LEE)


    To ensure the coolability of an overheated core, reflood of the uncovered fuel elements is an essential accident management measure to terminate a severe accident transient in Light Water Reactors (LWR) and therefore to avoid further core degradation. From analysis of the TMI-2 accident it is known that an enhanced oxidation of the zircaloy cladding may occur before the water succeeds in cooling the fuel rods. This oxidation process results in a sharp temperature increase, hydrogen generation and can finally after fuel rod cladding failure lead to fission product release. For further development and validation of the program ATHLET-CD, post-test calculations of experiments creating a reflood scenario in a controlled and defined environment are used. One of these experiments is QUNECH-12 in which Zr1%Nb (E 110) fuel rod claddings are used. Typically, fuel rods of VVER reactors are made from this material. In this work, the QUENCH-12 experiment and its conduct will be presented followed by explanations of the modeling in ATHLET-CD version 2.2A. Results of ATHLET-CD simulating the test will be discussed in order to validate the code's ability to adequately calculate phenomena like hydrogen production and melt oxidation during reflooding of uncovered fuel rods of E 110. (orig.)

  11. Irradiation-induced structural changes in surveillance material of VVER 440-type weld metal

    Grosse, M.; Denner, V.; Böhmert, J.; Mathon, M.-H.


    The irradiation-induced microstructural changes in surveillance materials of the VVER 440-type weld metal Sv-10KhMFT were investigated by small angle neutron scattering (SANS) and anomalous small angle X-ray scattering (SAXS). Due to the high fluence, a strong effect was found in the SANS experiment. No significant effect of the irradiation is detected by SAXS. The reason for this discrepancy is the different scattering contrast of irradiation-induced defects for neutrons and X-rays. An analysis of the SAXS shows that the scattering intensity is mainly caused by vanadium-containing (VC) precipitates and grain boundaries. Both types of scattering defects are hardly changed by irradiation. Neutron irradiation rather produces additional scattering defects of a few nanometers in size. Assuming these defects are clusters containing copper and other foreign atoms with a composition according to results of atom probe field ion microscopy (APFIM) investigations, both the high SANS and the low SAXS effect can be explained.

  12. Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

    Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)


    At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

  13. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    Frybort, Jan


    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  14. PCA-based ANN approach to leak classification in the main pipes of VVER-1000

    Hadad, Kamal; Jabbari, Masoud; Tabadar, Z. [Shiraz Univ. (Iran, Islamic Republic of). School of Mechanical Engineering; Hashemi-Tilehnoee, Mehdi [Islamic Azad Univ., Aliabad Katoul (Iran, Islamic Republic of). Dept. of Engineering


    This paper presents a neural network based fault diagnosing approach which allows dynamic crack and leaks fault identification. The method utilizes the Principal Component Analysis (PCA) technique to reduce the problem dimension. Such a dimension reduction approach leads to faster diagnosing and allows a better graphic presentation of the results. To show the effectiveness of the proposed approach, two methodologies are used to train the neural network (NN). At first, a training matrix composed of 14 variables is used to train a Multilayer Perceptron neural network (MLP) with Resilient Backpropagation (RBP) algorithm. Employing the proposed method, a more accurate and simpler network is designed where the input size is reduced from 14 to 6 variables for training the NN. In short, the application of PCA highly reduces the network topology and allows employing more efficient training algorithms. The accuracy, generalization ability, and reliability of the designed networks are verified using 10 simulated events data from a VVER-1000 simulation using DINAMIKA-97 code. Noise is added to the data to evaluate the robustness of the method and the method again shows to be effective and powerful. (orig.)

  15. A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core

    Saarinen, Simo [Fortum Nuclear Services Ltd., P.O. Box 100, 00048 Fortum (Finland)


    Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pin-power reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretized using piecewise linear polynomials. The core is discretized using one node per fuel pin cell. The axial discretization can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modelling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented. HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretizations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. (author)

  16. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center


    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  17. Contribution to the study of nuclear resonance in magnetic media (1963); Contribution a l'etude de la resonance nucleaire dans les milieux magnetique (1963)

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


    An attempt is made to interpret the results of nuclear magnetic resonance experiments made by various workers on ferro and ferrimagnetic compounds of the iron group. The problems encountered are the following: effects of the dipolar fields and the hyperfine structure anisotropy; signal intensity; frequency pulling due to the Suhl-Nakamura interaction between nuclear spins ; nuclear relaxation and ferrimagnetic resonance in single domain samples of impure YIG; nuclear relaxation in the Bloch walls of insulators. The results of our calculations are generally in good agreement with experiment. (author) [French] On se propose d'interpreter les resultats d'experiences de resonance magnetique nucleaire fates par divers auteurs sur des composes ferro et ferrimagnetiques du groupe du fer. Les problemes abordes sont les suivants: effets des champs dipolaires et de l'anisotropie de structure hyperfine; intensite des signaux; deplacement de frequence du a l'interaction de Suhl-Nakamura entre spins nucleaires; relaxation nucleaire et resonance ferrimagnetique dans les echantillons monodomaines de grenat de fer et d'yttrium impur; relaxation nucleaire dans les parois de Bloch des isolants. Les resultats des calculs sont generalement en bon accord avec l'experience. (auteur)

  18. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.


    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  19. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    Hirschberg, G


    Full Text Available of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg G abor Hirschberg a,P al Baradlai a,K alm an Varga a,*, Gerrit Myburg b, J anos Schunk c,P eter Tilky c, Paul Stoddart d a Department of Radiochemistry, University...-cooled nuclear reactors is of great importance for a number of practical reasons. For instance, under normal operating conditions (when there is no ?ssion product release due to fuel cladding failure) the majority of radioactive contamination in the pri- mary...

  20. Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor.

    Košt'ál, Michal; Švadlenková, Marie; Milčák, Ján


    The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density.

  1. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others


    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  2. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others


    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  3. Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

    Miller, M. K.; Chernobaeva, A. A.; Shtrombakh, Y. I.; Russell, K. F.; Nanstad, R. K.; Erak, D. Y.; Zabusov, O. O.


    A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10 23 m -2 ( E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10 23 m -2 ( E > 0.5 MeV). High number densities of ˜2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT 41 J ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.

  4. The Design of PSB-VVER Experiments Relevant to Accident Management

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  5. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.


    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  6. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    S.M. Dmitriev


    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  7. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Dzhalandinov A.


    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  8. A comprehensive approach to selecting the water chemistry of the secondary coolant circuit in the projects of nuclear power stations equipped with VVER-1200 reactors

    Tyapkov, V. F.


    The paper presents the results obtained from studies on selecting the water chemistry of the secondary coolant circuit carried out for the project of a nuclear power station equipped with a new-generation VVER-1200 reactor on the basis of case calculations and an analysis of field experience gained at operating nuclear power stations.

  9. French research in the field of nuclear agronomy; Les recherches francaises en agronomie nucleaire

    Guerin De Montgareuil, P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires


    , industrial firms, university laboratories scientific institutes. The role of the Commissariat a l'Energie Atomique is defined: on the one hand it supplies information and support, and on the other hand it takes charge of specifically nuclear aspects of the work. Its part in the field has recently found expression in the creation, within the Biology Department, of a Radio-agronomy Section; its objective are described,, as well as the, means placed att its disposal at the Centre d'etudes Nucleaires, Cadarache. (author) [French] On propose un bilan des travaux les plus significatifs effectues en France depuis la deuxieme conference internationale en matiere d'agronomie nucleaire et qui vont d'une recherche apparemment desinteressee a l'application la plus directe. Une telle differenciation recouvre de moins en moins, au fur et a mesure de l'evolution des programmes, la distinction qui est faite dans l'expose entre l'action biologique des rayonnements et les autres emplois des techniques nucleaires. C'est ainsi que les recherches do radiogenetique agricole sont poursuivies dans deux directions: d'un point de vue theorique et methodologique avec l'etude comparative de l'action des divers types de rayonnements, l'influence du debit de dose et de la temperature, l'action des agents mutagenes chimiques, la production de chimeres sous irradiation gamma; et d'autre part, sous un aspect pratique aboutissant a la creation de varietes nouvelles plus resistantes ou plus precoces (riz, mil, arachide). Les problemes de destruction des insectes (eradication) et de conservation des denrees sous irradiation se trouvent egalement abordes par des voies et avec des objectifs tres divers. A la demarche globale representee par une irradiation pure et simple (grains humides, pommes de terre...) sont parfois associees des etudes souvent originales, d'ordre biochimique ou microbiologique (par exemple: alteration de l

  10. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi


    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  11. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail:


    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  12. Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

    Sabotinov, Luben; Chevrier, Patrick

    The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

  13. Study of thick, nuclear-compensated silicon detectors; Etude des detecteurs epais au silicium compense nucleairement

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


    A study is made here, from the point of view of the realization and the performance, of thick nuclear-compensated silicon detectors. After recalling the need for compensation and reviewing the existing methods, the author describes in detail the controlled realization of thick detectors by nuclear compensation from the theoretical and experimental points of view. The practical precautions which should be observed are given: control of the homogeneity of the starting material, control of the evolution of the compensation, elimination of parasitic processes. The performances of the detectors obtained are then studied: electrical characteristics (current, life-time) on the one hand, detection and spectrometry of penetrating radiations on the other hand. The results show, that the compensated diodes having an effective thickness of two millimeters operate satisfactorily as detectors for applied voltages of about 500 volts. The resolutions observed are then about 2 per cent for mono-energetic electrons and about 4 per cent for the gamma; they can be improved by the use of a pre-amplifier of very low background noise. (author) [French] Les detecteurs epais au silicium compense nucleairement sont etudies ici du double point de vue realisation et performances. Apres un rappel sur la necessite de la compensation et les procedes existants, la realisation controlee des detecteurs epais par compensation nucleaire est decrite en detail sous l'aspect theorique et l'aspect experimental. On met en evidence les precautions a prendre dans la pratique: controle de l'homogeneite du materiau de base, controle de l'evolution de la compensation, elimination des processus parasites. On etudie ensuite les performances de detecteurs obtenus : caracteristiques electriques (courant, duree de vie) d'une part, d'autre part detection et spectrometrie des rayonnements penetrants. Les resultats montrent que les diodes compensees ayant une epaisseur utile de deux

  14. Analytic catalog for the use of the nuclear physicists; Repertoire analytique a l'usage des physiciens nucleaires

    Ballini, R.; Barloutaud, R.; Bernas, R.; Chaminade, R.; Cohen, R.; Conjeaud, M.; Cotton, E.; Faraggi, H.; Grjebine, T.; Laboulaye, H. de; Lehmann, P.; Leveque, A.; Levi, C.; Moreau, J.; Naggiar, V.; Olkowsky, J.; Papineau, L.; Papineau, L.; Prugne, P.; Schuhl, C.; Szteinsznaider, D.; Tzara, C.; Valladas, G. [Commissariat a l' Energie Atomique, Saclay(France). Centre d' Etudes Nucleaires; Marty, N.; Renard, G. [College de France, Lab. de Chimie Nucleaire (France)


    The problem of the documentation in nuclear physics becomes constantly more complex. Every week brings its share of publications, always more numerous and more varied. To remedy to this problem, we tried, at the Nuclear Physics Services of the CEA, to give to the documentation the character of a collective and systematized work. The present report covers the literature appeared between January 1, 1950 and July 1, 1951. (Volume 1: CEA report number 120; Volume 2: CEA report number 184). (M.B.) [French] Le probleme de la documentation en physique nucleaire devient sans cesse plus complexe. Chaque semaine apporte son lot de publications, toujours plus nombreuses et plus diversifiees. C'est pour essayer de porter remede a cet etat de choses que nous avons essaye, au service de Physique Nucleaire du C.E.A., de donner a la documentation le caractere d'un travail collectif systematise. Le present rapport couvre la litterature parue entre le 1 Janvier 1950 et le 1 Juillet 1951. (Tome 1: Rapport CEA numero 120; Tome 2: Rapport CEA numero 184). (M.B.)

  15. Decontamination of nuclear liquid wastes using ion exchange. Case of the treatment of the water of the storage pond of the EDF Bugey 1 nuclear reactor; Decontamination d'effluents nucleaires par echange d'ions. Cas du traitement de l'eau de la piscine d'entreposage du reacteur nucleaire EDF Bugey-1

    Doury-Berthod, M. [CEA Saclay, 91 - Gif-sur-Yvette (France). Institut National des Sciences et Techniques Nucleaires


    The paper explains in detail the works carried out in order to decontaminate before dismantling the water contained in a cooling pond where were stored spent nuclear fuels from the Bugey 1 reactor. The studies included laboratory tests, the design and the operation of pilot and industrial units. The process was based on the use of ion exchange resins on several steps in order to completely fix the contaminating elements, especially cesium, strontium and plutonium.The procedure applied and the detailed results are given. (authors)

  16. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor; Cahier des charges specifiques pour la formation du personnel de categorie A ou B travaillant dans les installations nucleaires. Option reacteur nucleaire



    This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  17. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor-borne; Cahier des charges specifiques pour la formation du personnel de categorie A ou B travaillant dans les installations nucleaires. Option reacteur nucleaire embarque



    This document describes the specific dispositions relative to the nuclear reactor-borne domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor-borne and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  18. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    Naudet, R. [CEA, Paris (France)


    Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

  19. Security report on Siloe - the descriptive part. (1963); Rapport de surete de Siloe - partie descriptive (1963)

    Ageron, P.; Chatoux, J.; Denielou, G.; Jacquemain, M.; Mitault, G.; Robien, E. de; Rossillon, F. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires


    This report is a full description of the site, the reactor, the building and the experimental facilities. It gives the nuclear, thermodynamic and hydrodynamic characteristics of the core. (authors) [French] Ce rapport decrit completement le site, le reacteur, les batiments et les installations experimentales. Il donne les caracteristiques nucleaires, thermodynamiques et hydrodynamiques du coeur. (auteurs)

  20. A mathematical model for cost of maritime transport. Application to competitiveness of nuclear vessels; Modele mathematique du cout de transport maritime application a la competitivite du navire nucleaire

    Dorval, C. [Commissariat a l' Energie Atomique, 75 - Paris (France)


    In studying the competitiveness of a nuclear merchant vessel, economic assessments in terms of figures were discarded in favor of a simplified model, which gives a clearer idea of the mechanism of the comparison between alternative vessels and the particular influence of each parameter. An expression is formulated for the unit cost per ton carried over a given distance as a function of the variables (speed and deadweight tonnage) and is used to determine the optima for conventional and nuclear vessels. To represent the freight market involved in the optimization studies, and thus in the competitiveness computation, two cases are taken into account: the tonnage to be carried annually is limited, and the tonnage to be carried annually is not limited. In both cases the optima are calculated and compared for a conventional and a nuclear vessel. Competitiveness curves are plotted as a function of the ratios of nuclear and conventional fuel costs and nuclear and conventional marginal power costs. These curves express the limiting values of the above two ratios for which the transport costs of the nuclear and conventional vessels are equal. The competitiveness curves vary considerably according to the hypothesis adopted for the freight market and the limit of tonnage carried annually. (author) [French] Pour etudier la competitivite du navire marchand nucleaire, plutot que de nous livrer a des evaluations economiques chiffrees, discutables dans l'etat actuel des etudes, nous utilisons un modele simplifie permettant de mieux saisir le mecanisme de la comparaison des navires et l'influence particuliere de chaque parametre. Nous etablissons une expression du cout unitaire de la tonne transportee sur un parcours donne en fonction des variables vitesse et port en lourd. Et nous l'utilisons pour determiner les optima des navires classiques et nucleaires. Pour representer le marche du fret qui intervient dans les etudes d'optimisation, et donc dans la

  1. A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

    Bezrukov Yury Alekseevich


    Full Text Available The paper covers the results of VVER core reflooding studies in fuel assembly (FA mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 °C and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation.

  2. The Change of Austenitic Stainless Steel Elements Content in the Inner Parts of VVER-440 Reactor during Operation

    Smutný, Vladimír; Hep, Jaroslav; Novosad, Petr


    Neutron activation induces the element transmutation in materials surrounding the reactor active core. The objective of the present paper is to calculate and evaluate the change of austenitic stainless steel 08Ch18N10T elements content through neutron induced activation, in inner parts of VVER-440 - in the baffle and in the barrel. Particularly the content changes of Mn in austenitic stainless steel. The neutron flux density and then the neutron activation of austenitic stainless steel elements in parts at the core are calculated. Neutron activation represents a measure of austenitic stainless steel elements transmutation. The power distribution is determined as an average value of several cycles power distribution in the middle of a cycle for the NPP Dukovany. The power distribution is calculated with the code MOBY-DICK [1]. The neutron flux density is calculated with the code TORT [2]. The neutron activation of austenitic stainless steel elements in the baffle and in the barrel is calculated with the system EASY-2007 containing the code FISPACT-2007 [3]. The calculation of the changing austenitic stainless steel elements content is performed depending on the moment of the supposed end of reactor operation - 40 years. There is also necessary monitoring and benchmarking of steel element content change, because the neutron flux calculation, particularly in thermal region, shows a considerable uncertainty, e.g. [4]. The motivation for this work is the study focused to stress corrosion cracking of austenitic stainless steels induced by radiation inside PWR and BWR, e.g. [5]. The paper could be a suggestion to estimation of austenitic stainless steel corrosion damage induced by neutrons in inner parts of VVER-440 reactor.

  3. 1: the atom. 2: radioactivity. 3: man and radiations. 4: the energy. 5: nuclear energy: fusion and fission. 6: the operation of a nuclear reactor. 7: the nuclear fuel cycle; 1: l'atome. 2: la radioactivite. 3: l'homme et les rayonnements. 4: l'energie. 5: l'energie nucleaire: fusion et fission. 6: le fonctionnement d'un reacteur nucleaire. 7: le cycle du combustible nucleaire



    This series of 7 digest booklets present the bases of the nuclear physics and of the nuclear energy: 1 - the atom (structure of matter, chemical elements and isotopes, the four fundamental interactions, nuclear physics); 2 - radioactivity (definition, origins of radioelements, applications of radioactivity); 3 - man and radiations (radiations diversity, biological effects, radioprotection, examples of radiation applications); 4 - energy (energy states, different forms of energy, characteristics); 5 - nuclear energy: fusion and fission (nuclear energy release, thermonuclear fusion, nuclear fission and chain reaction); 6 - operation of a nuclear reactor (nuclear fission, reactor components, reactor types); 7 - nuclear fuel cycle (nuclear fuel preparation, fuel consumption, reprocessing, wastes management). (J.S.)

  4. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.


    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  5. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Carbajo, J.J.


    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  6. Determination of power distribution in the VVER-440 core on the basis of data from in-core monitors by means of a metric analysis

    Kryanev, A. V.; Udumyan, D. K.; Kurchenkov, A. Yu.; Gagarinskiy, A. A.


    Problems associated with determining the power distribution in the VVER-440 core on the basis of a neutron-physics calculation and data from in-core monitors are considered. A new mathematical scheme is proposed for this on the basis of a metric analysis. In relation to the existing mathematical schemes, the scheme in question improves the accuracy and reliability of the resulting power distribution.

  7. A RSM Method for Nonlinear Probabilistic Analysis of the Reinforced Concrete Structure Failure of a Nuclear Power Plant - Type VVER 440

    Králik, Juraj


    This paper describes the reliability analysis of a concrete containment for VVER 440 under a high internal overpressure. The probabilistic safety assessment (PSA) level 3 aims at an assessment of the probability of the concrete structure failure under the excessive overpressure. The non-linear analysis of the concrete structures was considered. The uncertainties of the loads level (long-time temperature and dead loads), the material model (concrete cracking and crushing, behavior of the reinf...

  8. General phenomenology of underground nuclear explosions; Phenomenologie generale des explosions nucleaires souterraines

    Derlich, S.; Supiot, F. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes


    An essentially qualitatively description is given of the phenomena related to underground nuclear explosions (explosion of a single unit, of several units in line, and simultaneous explosions). In the first chapter are described the phenomena which are common to contained explosions and to explosions forming craters (formation and propagation of a shock-wave causing the vaporization, the fusion and the fracturing of the medium). The second chapter describes the phenomena related to contained explosions (formation of a cavity with a chimney). The third chapter is devoted to the phenomenology of test explosions which form a crater; it describes in particular the mechanism of formation and the different types of craters as a function of the depth of the explosion and of the nature of the ground. The aerial phenomena connected with explosions which form a crater: shock wave in the air and focussing at a large distance, and dust clouds, are also dealt with. (authors) [French] On donne une description essentiellement qualitative des phenomenes lies aux explosions nucleaires souterraines (explosion d'un seul engin, d'engins en ligne et explosions simultanees). Dans un premier chapitre sont decrits les phenomenes communs aux explosions contenues et aux explosions formant un cratere (formation et propagation d'une onde de choc provoquant la vaporisation, la fusion et la fracturation du milieu). Le deuxieme chapitre decrit les phenomenes lies aux tirs contenus (formation d'une cavite et d'une cheminee). Le troisieme chapitre est consacre a la phenomenologie des tirs formant un cratere et decrit notamment le mecanisme de formation et les differents types de crateres en fonction de la profondeur d'explosion et de la nature du terrain. Les phenomenes aeriens lies aux explosions formant un cratere: onde de pression aerienne et focalisation a grande distance, nuages de poussieres, sont egalement abordes. (auteurs)

  9. Assessment of the TiO2/water nanofluid effects on heat transfer characteristics in VVER-1000 nuclear reactor using CFD modeling

    Seyed Mohammad Mousavizadeh


    Full Text Available The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid (TiO2/water on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR, etc. were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

  10. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Tyapkov, V. F.


    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  11. Assessment of the TiO{sub 2}/water nanofluid effects on heat transfer characteristics in VVER-1000 nuclear reactor using CFD modeling

    Mousavizadeh, Seyed Mohammad; Ansarifar, Gholam Reza; Talebi, Mansour [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)


    The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid (TiO2/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

  12. 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

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


    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

  13. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.


    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

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

    Berger, F.; Bertrand, J


    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)

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

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


    the beam extraction 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

  16. Contribution to the study of several chemical hazards in the Centre d'Etudes Nucleaires of Fontenay-aux-Roses; Contribution a l'etude de quelques nuisances chimiques au centre d'etudes nucleaires de Fontenay-aux-Roses

    Megemont, C.; Grau, C. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires


    From the checking of 2750 index cards of hazards, the study relates the distribution of the chemical hazards in the Centre d'Etudes Nucleaires of Fontenay-aux-Roses. Those concerning the greatest number of agents in the Centre are classified according to the categories corresponding to the different conditions of working. Thus, the most important are put forward. Then, the authors rapidly make a review of hazards which may have some special interest because they appear more specific of the nuclear energy or because they are the most frequently noted on the index cards of hazards. The case of the tributylphosphate is studied more precisely. (authors) [French] A partir de l'examen de 2750 fiches de nuisances, l'etude porte sur la repartition des nuisances chimiques au Centre d'Etudes Nucleaires de Fontenay-aux-Roses. Celles qui concernent le plus grand nombre d'agents du Centre sont classees selon les categories correspondant aux differentes conditions de travail. Les plus importantes d'entre elles sont ainsi mises en evidence. | Les auteurs passent ensuite en revue, rapidement, les nuisances qui peuvent presenter un interet particulier soit parce qu'elles semblent plus specifiques de l'Energie Nucleaire, soit parce qu'on les rencontre le plus frequemment sur les fiches de nuisances. Le cas du tributylphosphate est envisage de facon plus detaillee. (auteurs)

  17. Flica: a code for the thermodynamic study of a reactor or a test loop; Programme FLICA etude thermodynamique d'un reacteur ou d'une boucle d'essai

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


    This code handles the thermal problems of water loops or reactor cores under the following conditions: High or low pressure, steady state or transient behavior, one or two phases - Three-dimensional thermodynamic study of the flow in cylindrical geometry - Unidimensional study of heat transfer in heating elements - Neutronic studies can be coupled and a schematic representation of the safety rod behavior is given. The number of cells described in a flow cross-section is presently less than 20. This code is the logical following of FLID and CACTUS of which it constitutes a synthesis. (author) [French] Ce code permet de traiter les problemes thermiques d'une boucle ou d'un coeur de reacteur a eau dans les conditions suivantes: - Haute ou basse pression, regime permanent ou transitoire, simple ou double phase - Etude thermodynamique de l'ecoulement a 3 dimensions dans une geometrie cylindrique - Etude unidimensionnelle du transfert de chaleur dans les masses chauffantes - Possibilite de couplage avec la neutronique (reacteur point) et d'une representation schematique des actions de securite. Ce code dans lequel le nombre de cellules decrites dans une section droite de l'ecoulement est actuellement limite a 20 est la suite logique des codes FLID et CACTUS dont il constitue la synthese. (auteur)

  18. Continuous dissolution of irradiated nuclear fuels; Dissolution continue des combustibles nucleaires irradies

    Michel, P.; Talmont, X.; Tarnerq, M. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires


    In the case of the continuous dissolution of nuclear fuels, the equations for the calculation of the fuel concentration of the solution flowing out of a pot dissolver have been written. Nitric acid feed flow rates have been calculated in order to obtain an adjusted solution when starting or stopping a dissolution, or when changing the number of rods introduced per hour. Then some transient states brought on by perturbations, have been studied: a) sudden change in nitric acid flow rate; b) continuous drift of the latter; c) sudden change in nitric acid feed concentration; d) transition from a fuel concentration to another by changing the flow rate of nitric acid feed. It has been shown that some transient states cannot be solved with general equations. Computer calculation programs would be probably more useful. (authors) [French] L'etude se rapporte a la dissolution dans l'acide nitrique des combustibles nucleaires irradies, en vue de la recuperation de la matiere fissile qu'ils contiennent. On a etabli, dans le cas de la dissolution continue, les differentes equations permettant le calcul de la concentration en combustible a la sortie d'un dissolveur du type 'marmite'. On a etudie les regimes du debit d'alimentation en acide nitrique a imposer lors du demarrage, de l'arret d'une dissolution, ou lors d'un changement de cadence d'introduction des barreaux, de facon a obtenir une solution ajustee. On a etudie ensuite differents regimes transitoires consecutifs a des perturbations: changement brusque du debit d'acide d'alimentation, derive continue de ce debit, changement brusque de la concentration de l'acide d'alimentation, passage d'une concentration en combustible a une autre par changement du debit d'acide d'alimentation. On a pu montrer que certains regimes transitoires ne peuvent se traiter par des equations generales, et necessiteraient plustot l'etablissement d

  19. Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

    Hölgye, Z; Filgas, R


    Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

  20. The concept of extending the service life of the VVER-440-based power units at the Novovoronezh nuclear power plant

    Asmolov, V. G.; Povarov, V. P.; Vitkovskii, S. L.; Berkovich, V. Ya.; Chetverikov, A. E.; Mozul', I. A.; Semchenkov, Yu. M.; Suslov, A. I.


    Basic statements of the Concept of Extending the Service Life of the VVER-440-Based Power Units at the Novovoronezh NPP beyond 45 years are considered. This topic is raised in connection with the fact that that in December 2016 and in December 2017 the extended service lives of Units 3 and 4 at this NPP will expire. The adopted concept of repeatedly extending the service life of the Novovoronezh NPP Unit 4 implies fitting the power unit with additional reactor core cooling systems with a view to extend the (ultimate) design-basis accidents (which have hitherto been adopted to be a loss of coolant accident involving a leak of reactor coolant through a break with a nominal diameter of 100 mm) to a reactor coolant leak equivalent to rupture of the main reactor coolant pipeline. The modified Unit 4 will also use the safety systems of Unit 3 that is going to be decommissioned. Preliminary calculated assessments of the new design-basis accident scenario involving rupture of the reactor coolant pipeline in Unit 4 fitted with a new configuration of safety systems confirmed the correctness of the adopted concept of repeatedly extending the service life of Unit 4.

  1. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.


    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  2. Neutron and gamma field investigations in the VVER-1000 mock-up concrete shielding on the reactor LR-0

    Zaritsky, S.; Egorov, A. [National Research Center, Kurchatov Inst., Moscow 123182 (Russian Federation); Osmera, B.; Marik, M.; Rypar, V. [Research Centre Rez Ltd., Rez 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, Brno 61200 (Czech Republic); Kolros, A. [Czech Technical Univ., Prague 18000 (Czech Republic)


    Two sets of neutron and gamma field investigations were carried out in the dismountable model of radiation shielding of the VVER-1000 mock-up on the LR-0 reactor. First, measurements and calculations of the {sup 3}He(n,p)T reaction rate and fast neutrons and gamma flux spectra in the operational neutron monitor channel inside a concrete shielding for different shapes and locations of the channel (cylindrical channel in a concrete, channels with collimator in a concrete, cylindrical channel in a graphite). In all cases measurements and calculations of the {sup 3}He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the {sup 3}He(n,p)T reaction rate spatial distribution inside a concrete. The {sup 3}He(n,p)T reaction rate measurements and calculations were carried out exploring the relative thermal neutron density in the channels and its space distribution in the concrete. Fast neutrons and gamma measurements were carried out with a stilbene (45 x 45 mm) scintillation spectrometer in the energy regions 0.5-10 MeV (neutrons) and 0.2-10 MeV (gammas). (authors)

  3. Statistical pulses generator. Application to nuclear instrumentation (1962); Generateur d'impulsions aleatoires. Application a l'instrumentation nucleaire (1962)

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


    This report describes a random pulses generator adapted to nuclear instrumentation. After a short survey on the statistical nature of electronic signals, the different ways for generating pulses with a Poisson's time-distribution are studied. The final generator built from a gaseous thyratron in a magnetic field is then described. Several tests are indicated : counting-rate stability, Pearson's criterion, distribution of time-intervals. Applications of the generator in 'whole testing' of nuclear instrumentation are then indicated for sealers, dead time measurements, time analyzers. In this application, pulse-height spectrums have been made by Poissonian sampling of a recurrent or random low-frequency signal. (author) [French] Cette etude decrit un generateur d'impulsions aleatoires et ses applications a l'instrumentation nucleaire. Apres un bref rappel sur la nature statistique des signaux en electronique nucleaire, sont passes en revue les principaux moyens d'obtenir des impulsions distribuees en temps suivant une loi de Poisson. Le generateur utilisant un thyratron a gaz dans un champ magnetique est ensuite decrit; diverses methodes de test sont appliquees (stabilite du taux de comptage, criterium de Pearson, spectre des intervalles ds temps). Les applications du generateur a l'electronique nucleaire dans le domaine des 'essais globaux' sont indiques: test des echelles de comptage et mesure des temps morts, test des analyseurs en temps apres division du taux de comptage par une puissance de deux, test des analyseurs multicanaux en amplitude. Pour cette derniere application, on a realise des spectres d'amplitudes suivant une loi connue, par echantillonnage poissonien d'un signal basse frequence recurrent ou aleatoire. (auteur)

  4. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    Osadchaya, D. Yu.; Fuks, R. L.


    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  5. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.


    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  6. Specification of a VVER-1000 SFAT device prototype. Interim report on Task FIN A 1073 of the Finnish Support Programme to IAEA Safeguards

    Nikkinen, M. [Radiation and Nuclear Safety Authority, Helsinki (Finland); Tiitta, A. [VTT Chemical Technology, Espoo (Finland); Iievlev, S.; Dvoeglazov, M.; Lopatin, S. [Ministry of Environmental Protection and Nuclear Safety, Kiev (Ukraine)


    The project to specify the optimal design of the Spent Fuel Attribute Tester (SFAT) for Ukrainian VVER-1000 facilities was run under Finnish Support Programme for IAEA Safeguards under the task FIN A1073. This document illustrates the optimum design and takes into account the special conditions at the Ukrainian facilities. The requirement presented here takes into account the needs of the user (IAEA), nuclear safety authority (NRA) and facilities. This document contains the views of these parties. According to this document, the work to design the optimal SFAT device can be started. This document contains also consideration for the operational procedures, maintenance and safety. (orig.) 5 refs.

  7. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor; Spectrometrie gamma haute resolution et hauts taux de comptage sur primaire de reacteur de type generation 4 au sodium liquide

    Coulon, R.


    Sodium-cooled Fast Reactors are under development for the fourth generation of nuclear reactor. Breeders reactors could gives solutions for the need of energy and the preservation of uranium resources. An other purpose is the radioactive wastes production reduction by transmutation and the control of non-proliferation using a closed-cycle. These thesis shows safety and profit advantages that could be obtained by a new generation of gamma spectrometry system for SFR. Now, the high count rate abilities, allow us to study new methods of accurate power measurement and fast clad failure detection. Simulations have been done and an experimental test has been performed at the French Phenix SFR of the CEA Marcoule showing promising results for these new measurements. (author) [French] Les reacteurs a neutrons rapides refroidis au sodium sont en developpement en vue d'assurer une quatrieme generation de reacteurs repondant a la demande energetique, tout en assurant la preservation des ressources d'uranium par un fonctionnement en surgenerateur. L'objectif de la filiere est egalement d'ameliorer la gestion de la radiotoxicite des dechets produits par transmutation des actinides mineurs et de controler la non-proliferation par un fonctionnement en cycle ferme. Une instrumentation de surveillance et de controle de ce type de reacteur a ete etudiee dans cette these. La spectrometrie gamma de nouvelle generation permet, par les hauts taux de traitement aujourd'hui accessibles, d'envisager de nouvelles approches pour suivre avec une precision accrue la puissance neutronique et de detecter plus precocement des ruptures de gaine combustible. Des simulations numeriques ont ete realisees et une campagne d'essai a ete menee a bien sur le reacteur Phenix de Marcoule. Des perspectives prometteuses ont ete mises en exergue pour ces deux problematiques

  8. Nuclear developments at the international inter govern mental level (1961); Developpements nucleaires sur le plan international intergouvernemental (1961)

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


    The United Nations organisation and nuclear energy rose simultaneously, in 1945, to occupy an important place in the public eye. The spiritual succession of the League of Nations which had foundered during the war was taken up by the new organisation which sought to implant its political ideal in a more tangible reality, so that it might thereby be inspired by concrete and substantial objectives. This is one of the reasons for the existence of the dozen specialized agencies created by the family of the United Nations and dealing with Health, Culture, Agriculture, Finance, etc. Nuclear energy is one of these techniques. Becoming suddenly an important power factor and exploiting for itself the prestige of Science, it became the favorite domain for the growth of this new spirit, as much in its universal form in 1945 as in its more regional form which it was later to adopt. The achievements are numerous and of varying importance; they deserve te be studied carefully. (author) [French] L'organisation des Nations Unies et le nucleaire ont ete places simultanement, en 1945, au premier plan de la scene mondiale. La Societe des Nations ayant sombre pendant la guerre, son heritage spirituel fut recueilli par la nouvelle organisation qui chercha a enraciner son ideal politique dans une realite plus materielle, de facon a y puiser une nourriture concrete et substantielle. C'est une des raisons d'existence de la douzaine d'institutions specialisees gravitant dans la famille des Nations Unies et s'occupant de Sante, de Culture, d'Agriculture, de Finances, etc. Le nucleaire est l'une de ces techniques. Devenu soudainement un facteur primordial de puissance, Cristallisant a son benefice le prestige de la Science, c'etait un terrain de predilection pour le developpement du nouvel esprit, aussi bien sous la forme universelle de 1945, que sous les formes regionales qui ont vu le jour ensuite. Les travaux realises que nombreux, d

  9. 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)

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


    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)

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

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


    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

  11. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Sanchez-Espinoza, Victor Hugo


    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  12. Laser-induced mobilization of dust produced during fusion reactors operation; Mise en suspension par laser de poussieres generees lors du fonctionnement des reacteurs de fusion

    Vatry, A.


    During tokamak operation, plasma-wall interactions lead to material erosion process and dusts production. These dusts are mainly composed by carbon and tungsten, with sizes ranging from 10 nm to 100 {mu}m. For safety reasons and to guarantee an optimum reactor functioning, the dusts have to be kept in reasonable quantity. The dusts mobilization is a first step to collect them, and the laser is a promising technique for this application. To optimize the cleaning, physical mechanisms responsible for dust ejection induced by laser have been identified. Some particles, such as aggregates, are directly ablated by the laser. The metal droplets are ejected intact by an electrostatic force, induced by the photoelectrons. We also characterized the particles ejection to choose an appropriate collection device. (author) [French] Lors du fonctionnement d'une machine de fusion, les interactions plasma-parois conduisent a des processus d'erosion des materiaux et a la production de particules. Ces poussieres sont principalement composees de carbone et de tungstene. Pour des raisons de surete et afin de garantir un fonctionnement optimum du reacteur, il est important de garder en quantite raisonnable les poussieres dont la taille varie entre 10 nm et 100 {mu}m. La mise en suspension de ces poussieres est une etape preliminaire a leur recuperation, et le laser est une technique prometteuse pour cette application. Afin d'optimiser le nettoyage, les mecanismes physiques a l'origine de l'ejection induite par laser de ces poussieres ont ete identifies. Les agregats sont directement ablates par le laser et les gouttelettes metalliques sont ejectees intactes par une force electrostatique induite par les photoelectrons. Nous avons egalement caracterise l'ejection des particules pour choisir un systeme de recuperation adapte

  13. Koroze slitin Zr–Nb a Zr–Sn za simulovaných podmínek reaktoru VVER

    Krausová A.


    Full Text Available Oxidační kinetiku lze v případě koroze zirkoniových slitin rozdělit do dvou stádií, která jsou od sebe oddělená transientním stavem. Výskyt transientního stavu je z korozního hlediska klíčový, neboť při něm dochází ke zvýšení rychlosti oxidace. V této práci byly pomocí in-situ elektrochemické impedanční spektroskopie studovány slitiny zirkonia Zr–Nb a Zr–Sn při teplotě 340 °C a tlaku 15 MPa v prostředí, které simulovalo chladivo reaktoru VVER. Cílem dlouhodobých experimentů (až 9000 h bylo charakterizovat oxidační kinetiku v závislosti na expoziční době a typu zirkoniové slitiny. Výsledky ukázaly, že za uvedených experimentálních podmínek dochází u slitiny zirkonia Zr–Sn k dřívějšímu dosažení přechodu, tzv. transientnímu stavu.

  14. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    Hirschberg, Gábor; Baradlai, Pál; Varga, Kálmán; Myburg, Gerrit; Schunk, János; Tilky, Péter; Stoddart, Paul

    Formation, presence and deposition of corrosion product radionuclides (such as 60Co, 51Cr, 54Mn, 59Fe and/or 110mAg) in the primary circuits of water-cooled nuclear reactors (PWRs) throw many obstacles in the way of normal operation. During the course of the work presented in this series, accumulations of such radionuclides have been studied at austenitic stainless steel type 08X18H10T (GOST 5632-61) surfaces (this austenitic stainless steel corresponds to AISI 321). Comparative experiments have been performed on magnetite-covered carbon steel (both materials are frequently used in some Soviet VVER type PWRs). For these laboratory-scale investigations a combination of the in situ radiotracer `thin gap' method and voltammetry is considered to be a powerful tool due to its high sensitivity towards the detection of the submonolayer coverages of corrosion product radionuclides. An independent technique (XPS) is also used to characterize the depth distribution and chemical state of various contaminants in the passive layer formed on austenitic stainless steel. In the first part of the series the accumulation of 110mAg has been investigated. Potential dependent sorption of Ag + ions (cementation) is found to be the predominant process on austenitic steel, while in the case of magnetite-covered carbon steel the silver species are mainly depleted in the form of Ag 2O. The XPS depth profile of Ag gives an evidence about the embedding of metallic silver into the entire passive layer of the austenitic stainless steel studied.

  15. Some aspects of the nuclear fission process; Quelques aspects du processus de fission nucleaire

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


    In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U{sup 233}, U{sup 235}, Pu{sup 239}, U{sup 238} are described at the beginning of this work. It appears that for U{sup 233} there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U{sup 239} than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U{sup 235}. Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [French] Le present travail debute par un apercu de l'etat actuel de nos connaissances sur le processus de fission nucleaire, notamment sur le passage par le point-seuil. Puis sont evoques des aspects lies au niveau d'energie d'excitation auquel est porte le noyau qui subit la fission. Les mesures de sections efficaces de fission induite dans {sup 233}U, {sup 235}U, {sup 239}Pu et {sup 238}U par des neutrons rapides effectuees a Saclay sont decrites en premier lieu; elles font apparaitre pour {sup 233}U une ondulation caracteristique du role des etats collectifs d'excitation du noyau deforme au point-seuil. Des experiences sur la fission avec emission de particules de long parcours confirment cet aspect tout en demontrant que

  16. Some features of the effect the pH value and the physicochemical properties of boric acid have on mass transfer in a VVER reactor's core

    Gavrilov, A. V.; Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.


    Certain features of the effect of boric acid in the reactor coolant of nuclear power installations equipped with a VVER-440 reactor on mass transfer in the reactor core are considered. It is determined that formation of boric acid polyborate complexes begins under field conditions at a temperature of 300°C when the boric acid concentration is equal to around 0.065 mol/L (4 g/L). Operations for decontaminating the reactor coolant system entail a growth of corrosion product concentration in the coolant, which gives rise to formation of iron borates in the zones where subcooled boiling of coolant takes place and to the effect of axial offset anomalies. A model for simulating variation of pressure drop in a VVER-440 reactor's core that has invariable parameters during the entire fuel campaign is developed by additionally taking into account the concentrations of boric acid polyborate complexes and the quantity of corrosion products (Fe, Ni) represented by the ratio of their solubilities.

  17. On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

    Zizin, M. N.; Ivanov, L. D.


    In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

  18. Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors

    Alipchenkov, V. M.; Belikov, V. V.; Davydov, A. V.; Emel'yanov, D. A.; Mosunova, N. A.


    Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.

  19. Public debate about the EPR nuclear power plant at Flamanville; Debat public sur la centrale nucleaire EPR a Flamanville



    The project of building of he EPR reactor at Flamanville (Manche, France) has been submitted to the public debate. This document includes a presentation of the project and of the rules of the public debate, a synthesis of the file made by the prime contractor (EDF), a synthesis of the collective book of national actors concerned by the project (a group of associations for environment protection, Areva company, the ministries of economy and ecology, Global Chance, association of pro-nuclear ecologists (AEPN), 'Sortir du Nucleaire' (out-of nuclear) network, group of scientists for the information about nuclear (GSIEN), association for the promotion of the Flamanville site (Proflam), French nuclear energy society (SFEN) in association with 'Sauvons le Climat' (let's save climate), regional collective association 'EPR non merci, ni ailleurs, ni ici' (EPR, no thanks, neither elsewhere, nor here), NegaWatt), and 5 detailed books of actors: ACRO (association for the control of radioactivity in Western France), CFDT and CGT syndicates, the economic and social council of Basse Normandie region, and Proflam. (J.S.)

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

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


    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

  1. Influence of operation factors on brittle fracture initiation and critical local normal stress in SE(B) type specimens of VVER reactor pressure vessel steels

    Kuleshova, E. A.; Erak, A. D.; Kiselev, A. S.; Bubyakin, S. A.; Bandura, A. P.


    A complex of mechanical tests and fractographic studies of VVER-1000 RPV SE(B) type surveillance specimens was carried out: the brittle fracture origins were revealed (non-metallic inclusions and structural boundaries) and the correlation between fracture toughness parameters (CTOD) and fracture surface parameters (CID) was established. A computational and experimental method of the critical local normal stress determination for different origin types was developed. The values of the critical local normal stress for the structural boundary origin type both for base and weld metal after thermal exposure and neutron irradiation are lower than that for initial state due to the lower cohesive strength of grain boundaries as a result of phosphorus segregation.

  2. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    Shchelik, S. V.; Pavlov, A. S.


    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  3. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)


    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  4. Verification Calculation Results to Validate the Procedures and Codes for Pin-by-Pin Power Computation in VVER Type Reactors with MOX Fuel Loading

    Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.


    One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.

  5. Spent VVER fuel characterisation combining a fork detector with gamma spectrometry. Interim report on Task JNT A 1071 FIN of the Finnish Support Programme to IAEA Safeguards

    Tiitta, A.; Hautamaeki, J. [VTT Chemical Technology, Espoo (Finland)


    According to the IAEA's criteria a partial defect verification of spent fuel assemblies has to be performed before they become difficult to access. A partial defect test for spent fuel should be able to detect if half or more of the fuel pins have been removed from an assembly or possibly replaced by dummies. Therefore a partial defect test procedure needs to be developed by evolving the measurement systems and the analysis methods of the measurement data. 18 VVER assemblies were measured with an enhanced fork detector at the Loviisa KPA Store in May 2000. This measurement campaign is a follow-on to the campaigns conducted in 1999 at the Olkiluoto KPA Store, where BWR assemblies were measured using the same instrument. The validity of correction methods developed in data analysis of Olkiluoto measurements was investigated in the analysis of Loviisa measurements. The share of {sup 244}Cm neutron source out of the total neutron counts is derived from the results calculated with the PYVO code. The enrichment correction method to the neutron data corresponding to that used for BWR assemblies was applied for VVER assemblies. The contribution of other gamma emitting nuclides than {sup 137}Cs was eliminated from the gross gamma signal with the help of gamma spectroscopy using the method developed for the BWR data. All these corrections were found to improve the essential correlations. An assembly may have off-reactor cycles between irradiation cycles. The measured {sup 137}Cs gamma signal can be corrected for off cycles using the recipes introduced in this report. Also the effect of off cycles to the neutron signal is contemplated. The correction for off cycles may be very important for a correct burnup verification of those assemblies, which have not been irradiated in sequential cycles. (orig.)

  6. Estimation of thermal loads on the VVER vessel under conditions of inversion of the stratified molten pool in a severe accident

    Loktionov, V. D.; Mukhtarov, E. S.


    Analysis of the thermal state of molten pools that can be formed on the vessel bottom of the VVER-600 medium-power reactor during a severe anticipated accident with melting of the core is represented. Two types of the molten pool of core materials, with the two-layer and inverse three-layer stratification, are considered. Thermal loads acting on the reactor vessel from the melt are estimated depending on its formation time. Features of the thermal state of the melt in the case of its inverse stratification are analyzed. It is shown that thermal loads on the reactor vessel exceed the critical heat flux (CHF) when forming the two-layer stratified molten pool 10 and 24 h after its shutdown, and the thermal load is close to the corresponding CHF or somewhat exceeds it in 72 h. In the case of the formation of the inverse structure of the melt, one can observe a decrease by more than 2.5 times (in comparison with the two-layer stratified structure) in the thermal load on the reactor vessel in the region of its contact with the upper layer of the steel melt. Analysis of results showed that maximum densities of heat flux to the reactor vessel from the bottom metallic layer with the melt inversion did not exceed corresponding CHFs 24 and 72 h after the reactor shutdown. Because the thermal load on the reactor vessel can be localized in the region of its bottom, where the CHF is relatively small, during the inverse stratification of the melt, there is a need to carry out further in-depth experimental and analytical investigations of conditions for formation of the stratified molten pool and to obtain corrected experimental CHFs for conditions and outlines of cooling the external surface of the VVER-600 vessel in a severe accident.

  7. Decree n. 2007-534 of the 4. april 2007 allowing the creation of the base nuclear installation named Flamanville 3, including a EPR type reactor, on the site of Flamanville (Manche); Decret no 2007-534 du 10 avril 2007 autorisant la creation de l'installation nucleaire de base denommee Flamanville 3, comportant un reacteur nucleaire de type EPR, sur le site de Flamanville (Manche)



    This decree gives the authorization to EDF to create on the site of Flamanville a nuclear installation including a PWR type reactor for a power of 4500 MW and devoted to the electric production. This reactor will can use uranium oxide or a mixture of uranium oxide and plutonium oxide. Considerations concerning the safety are given, as well as the control of the impact of this exploitation on the populations and the environment. (N.C.)

  8. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)


    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  9. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L


    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  10. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Solonin, V. I.; Perevezentsev, V. V.


    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  11. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Borodkin Pavel


    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  12. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay


    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  13. Development and application of an information-analytic system on the problem of flow accelerated corrosion of pipeline elements in the secondary coolant circuit of VVER-440-based power units at the Novovoronezh nuclear power plant

    Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Kiselev, A. N.; Shepelev, S. V.; Galanin, A. V.


    Specific features relating to development of the information-analytical system on the problem of flow-accelerated corrosion of pipeline elements in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh nuclear power plant are considered. The results from a statistical analysis of data on the quantity, location, and operating conditions of the elements and preinserted segments of pipelines used in the condensate-feedwater and wet steam paths are presented. The principles of preparing and using the information-analytical system for determining the lifetime to reaching inadmissible wall thinning in elements of pipelines used in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered.

  14. 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

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


    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

  15. Decision no. 2011-DC-0216 of the French nuclear safety authority from May 5, 2011, ordering the Laue Langevin Institute to proceed to a complementary safety evaluation of its basic nuclear facility (high flux reactor - INB no. 67) in the eyes of the Fukushima Daiichi nuclear power plant accident; Decision no. 2011-DC-0216 de l'Autorite de surete nucleaire du 5 mai 2011 prescrivant a l'Institut Laue Langevin (ILL) de proceder a une evaluation complementaire de la surete de son installation nucleaire de base (Reacteur a Haut Flux - INB n.67) au regard de l'accident survenu a la centrale nucleaire de Fukushima Daiichi



    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the Laue Langevin Institute, operator of the high flux research reactor (RHF) of Grenoble (France). (J.S.)

  16. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)


    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  17. Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

    Mun, Ch


    Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

  18. INB 40 OSIRIS and ISIS reactors: complementary safety evaluation in the light of the Fukushima Daiichi nuclear power station accident; INB 40 reacteurs OSIRIS et ISIS evaluation complementaire de la surete au regard de l'accident servenu a la centrale nucleaire de Fukushima Daiichi



    This report proposes a complementary safety evaluation of the Osiris and Isis reactors (INB 40) in Saclay, one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents some characteristics of the installation (brief description, buildings, effluents, control-command, radioprotection), identifies the risks of cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme events, combination of events, earthquake or flooding with a higher level than that for which the installation is designed, measures to prevent a cliff effect). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, training and exercises, available intervention means, robustness of available means, measures for the protection of the confinement integrity after fuel damage. It discusses the conditions of the use of subcontractors

  19. The Jules Horowitz reactor: complementary safety evaluation in the light of the Fukushima 1 nuclear power station accident; Reacteur Jules Horowitz evaluation complementaire de la surete au regard de l'accident servenu a la centrale nucleaire de Fukushima 1



    This report proposes a complementary safety evaluation of the Jules Horowitz reactor in Cadarache (INB 172), one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents the main characteristics of the installation, identifies the risks of a cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme meteorological conditions related to flooding, earthquake or flooding with a higher level than that for which the installation is designed). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, available intervention means, robustness of available means. It discusses the conditions of the use of subcontractors

  20. Safety approach for a facility coupling a nuclear reactor to a chemical plant, generic principles and application to a hydrogen production process; Approche de surete d'une installation associant un reacteur nucleaire a une usine chimique, principes generiques et application a un procede de production d'hydrogene

    Bertrand, F.; Barbier, D.; Bassi, A. [CEA Cadarache, Direction de l' Energie Nucleaire, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs


    The aim of this paper is to propose an overall safety approach devoted to the coupling on a same site of a nuclear reactor to a plant of hydrogen production. Such facilities depend on their own safety principles and practices and are submitted to their own regulation. Therefore, the approach presented here takes into account the aforementioned constraints and takes into consideration the various risks on the site in the design process of the coupling system. This approach relying on the defence in depth concept declined in five levels led to a generalization of the notion of physical barriers and safety functions applied in the French nuclear safety approach. Three main safety functions can be considered for the whole coupled facility : the control of the nuclear and chemical reactivity, the power extraction and the confinement of hazardous materials. Moreover, according to the concept of defence in depth, different plant conditions (normal, incidents and accidents) have been analyzed for the whole facility. Furthermore, the safety approach proposed for the chemical plant is aimed to select reference scenarios taking into account their probability and their consequences on the basis of the methodology presented in the ARAMIS European project. Finally, the purpose of the safety analysis of the chemical plant is the assessment of adequate safety distances to protect people outside of the site as well as the coupling system and, above everything, the nuclear reactor containment. In other respects, a progressive response aiming to avoid the reactor scram is proposed to manage with incidents. (authors)

  1. SARNET european excellence network on nuclear reactor major accidents. Display and realizations after a year of operating; Sarnet reseau d'excellence europeen sur les accidents graves de reacteur nucleaire. Son deploiement et ses realisations apres une annee de fonctionnement



    The Sarnet (Severe Accident Research NETwork of excellence) is devoted to the research on major accidents of water cooled reactors. The developed knowledge will be integrated in a simulation tool ASTEC co-developed with the IRSN and the GRS. This evaluation report presents the context, the objectives and the program of the Sarnet network. It discusses the network operating and the ASTEC simulation code. Some examples of experimental programs are provided. (A.L.B.)

  2. A numerical study on a lumped-parameter model and a CFD code coupling for the analysis of the loss of coolant accident in a reactor containment; Etude numerique 0D-multiD pour l'analyse de perte de refrigerant dans une enceinte de confinement d'un reacteur nucleaire

    Choi, Y.J.


    In the case of PWR severe accident (Loss of Coolant Accident, LOCA), the inner containment ambient properties such as temperature, pressure and gas species concentrations due to the released steam condensation are the main factors that determine the risk. For this reason, their distributions should be known accurately, but the complexity of the geometry and the computational costs are strong limitations to conduct full three-dimensional numerical simulations. An alternative approach is presented in this thesis, namely, the coupling between a lumped-parameter model and a CFD. The coupling is based on the introduction of a 'heat transfer function' between both models and it is expected that large decreases in the CPU-costs may be achieved. First of all, wall condensation models, such as the Uchida or the Chilton-Colburn models which are implemented in the code CAST3M/TONUS, are investigated. They are examined through steady-state calculations by using the code TONUS-0D, based on lumped parameter models. The temperature and the pressure within the inner containment are compared with those reported in the archival literature. In order to build the 'heat transfer function', natural convection heat transfer is then studied by using the code CAST3M for a partitioned cavity which represents a simplified geometry of the reactor containment. At a first step, two-dimensional natural convection heat transfer without condensation is investigated only. Either the incompressible-Boussinesq fluid flow model or the asymptotic low Mach model are considered for solving the time dependent conservation equations. The SUPG finite element method and the implicit scheme are applied for the numerical discretization. The computed results are qualified by the second-order Richardson extrapolation method which allows obtaining the so-called 'Exact values', i.e. grid size independent values. The computations are also validated through non-partitioned cavity case studies. The discussion is focused on heat transfer characteristics such as the variations of the average Nusselt number (Nu-bar) versus the dimensionless thickness of the partition (0.01 {<=} {gamma} {<=} 0.2) and conductivity ratio of the partition wall to the fluid (1 {<=} {sigma} {<=} 10{sup 5}). Finally, a 'heat transfer function' is suggested based upon the thermal resistance of the partition wall. The validity of the model is assessed thanks to comparisons with 'half-cavity' simulations. (author)

  3. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident; Developpement d`un element fini coque. Application au comportement thermo-viscoplastique d`une cuve de reacteur nucleaire (REP) en situation d`accident grave

    Diaz, V


    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author) 113 refs.

  4. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor; Analyse des techniques d'homogeneisation spatiale et energetique dans la resolution de l'equation du transport des neutrons dans les reacteurs nucleaires

    Magat, Ph


    Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)

  5. Nuclear data covariances and sensitivity analysis, validation of a methodology based on the perturbation theory; application to an innovative concept: the molten thorium salt fueled reactor; Analyses de sensibilite et d'incertitude de donnees nucleaires. Contribution a la validation d'une methodologie utilisant la theorie des perturbations; application a un concept innovant: reacteur a sels fondus thorium a spectre epithermique

    Bidaud, A


    Neutron transport simulation of nuclear reactors is based on the knowledge of the neutron-nucleus interaction (cross-sections, fission neutron yields and spectra...) for the dozens of nuclei present in the core over a very large energy range (fractions of eV to several MeV). To obtain the goal of the sustainable development of nuclear power, future reactors must have new and more strict constraints to their design: optimization of ore materials will necessitate breeding (generation of fissile material from fertile material), and waste management will require transmutation. Innovative reactors that could achieve such objectives - generation IV or ADS (accelerator driven system) - are loaded with new fuels (thorium, heavy actinides) and function with neutron spectra for which nuclear data do not benefit from 50 years of industrial experience, and thus present particular challenges. After validation on an experimental reactor using an international benchmark, we take classical reactor physics tools along with available nuclear data uncertainties to calculate the sensitivities and uncertainties of the criticality and temperature coefficient of a thorium molten salt reactor. In addition, a study based on the important reaction rates for the calculation of cycle's equilibrium allows us to estimate the efficiency of different reprocessing strategies and the contribution of these reaction rates on the uncertainty of the breeding and then on the uncertainty of the size of the reprocessing plant. Finally, we use this work to propose an improvement of the high priority experimental request list. (author)

  6. A homogenization method for the modal analysis of a nuclear reactor with internal structures modelling and fluid-structure interaction coupling; Une methode d'homogeneisation pour l'analyse modale d'un reacteur nucleaire avec modelisation des structures internes et de l'interaction fluide / structure

    Sigrist, J.F. [DCN Propulsion, Service Technique et Scientifique, 44 - La Montagne (France); Broc, D. [CEA Saclay, Lab. d' Etude Mecanique et Sismique, 91 - Gif-sur-Yvette (France)


    A homogenization method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a 'reduced' numerical model accounting for inertial fluid-structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenization techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to reactor internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenization approach to the case of reactor internals is then exposed: it is shown that in such case, confinement effects can be modelled by a suitable modification of classical fluid-structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a 'reduced' model with homogenized fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenization approach is proved to be efficient from the numerical of view point and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted. (authors)

  7. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor; Simulation numerique et analyse du declenchement et du developpement des instabilites axiales dans les turbomachines: application a un transitoire de breche dans un reacteur nucleaire a helium

    Tauveron, N


    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge development is very limited (no more than few seconds). It is also shown that in the case of a break event, the turbomachine can have a significant contribution to decay heat removal from the nuclear core. At last, such a device is autonomous for a certain time only, and that this time is sensitive to some parameters such as break location and back pressure value. (author)

  8. Measurement of the neutron capture cross section of U{sup 234} in n-TOF at CERN for Generation IV nuclear reactors; Mesure de la section efficace de capture neutronique de l'{sup 234}U a n-TOF au CERN pour les reacteurs nucleaires de generation 4

    Dridi, W


    Accurate and reliable neutron capture cross sections are needed in many research areas, including stellar nucleosynthesis, advanced nuclear fuel cycles, waste transmutation, and other applied programs. In particular, the accurate knowledge of U{sup 234}(n,{gamma}) reaction cross section is required for the design and realization of nuclear power plants based on the thorium fuel cycle. We have measured the neutron capture cross section of U{sup 234}, with a 4{pi} BaF{sub 2} Total Absorption Calorimeter, at the recently constructed neutron time-of-flight facility n-TOF at CERN in the energy range from 0.03 eV to 1 MeV. Monte-Carlo simulations with GEANT4 and MCNPX of the detector response have been performed. After the background subtraction and correction with dead time and pile-up, the capture yield from 0.03 eV up to 1.5 keV was derived. The analysis of the capture yield in terms of R-matrix resonance parameters is discussed. We have identified 123 resonances and measured the resonance parameters in the energy range from 0.03 eV to 1.5 keV. The mean radiative width <{gamma}{sub {gamma}}> is found to be (38.2 {+-} 1.5) meV and the mean spacing parameter is (11.0 {+-} 0.2) eV, both values agree well with recommended values.

  9. 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

    Elias, P.


    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.

  10. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 13, 2011, 7:00 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011 (13 mars 2011 - Point a 7h)



    This situation note is established according to the information gained on March 13, 2011, at 7:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site is briefly presented with the progress of the accident management actions. (J.S.)

  11. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 16, 2011, 2:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 16 mars 2011 a 14 heures



    This situation note is established according to the information gained on March 16, 2011, at 2:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of all 6 reactors of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  12. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 18, 2011, 2:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 18 mars 2011 a 14 heures



    This situation note is established according to the information gained on March 18, 2011, at 2:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  13. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 10:30 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du mardi 15 mars 2011 a 10h30



    This situation note is established according to the information gained on March 15, 2011, at 10:30 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  14. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 10:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011 (15 mars 2011 - point a 22h00 heures)



    This situation note is established according to the information gained on March 15, 2011, at 10:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 4, 5 and 6 of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  15. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 15, 2011, 3:30 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 15 mars 2011 a 15h30



    This situation note is established according to the information gained on March 15, 2011, at 3:30 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 5 and 6 of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  16. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 17, 2011, 6:00 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 17 mars 2011 a 06 heures



    This situation note is established according to the information gained on March 17, 2011, at 6:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  17. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 12, 2011, 8:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur servenu le 11 mars 2011. Point de situation du 12 mars 2011 a 20 heures



    This situation note is established according to the information gained on March 12, 2011, at 8:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site is briefly presented with the progress of the accident management actions. The operation principles of a BWR-type reactor and of a PWR-type reactor are presented in appendix as well as the confinement principle specific to Mark I-type BWR reactors designed by General Electric. The meteorological forecasts of the day are presented in a figure. (J.S.)

  18. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 16, 2011, 9:00 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 16 mars 2011 a 9 heures



    This situation note is established according to the information gained on March 16, 2011, at 9:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2, 3 and 4 and of the spent fuel pools of reactors No. 4, 5 and 6 of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  19. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 20, 2011, 6:00 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 20 mars 2011 a 06 heures



    This situation note is established according to the information gained on March 20, 2011, at 6:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  20. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 16, 2011, 7:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 16 mars 2011 a 19 heures



    This situation note is established according to the information gained on March 16, 2011, at 7:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  1. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 13, 2011, 7:00 PM status - updated at 11:00 PM; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 13 mars 2011 a 19 heures - Mis a jour a 23 heures



    This situation note is established according to the information gained on March 13, 2011, at 7:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site (Dai-ichi), of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  2. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 19, 2011, 6:00 AM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 19 mars 2011 a 06 heures



    This situation note is established according to the information gained on March 19, 2011, at 6:00 AM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  3. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 17, 2011, 3:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur survenu le 11 mars 2011. Point de situation du 17 mars 2011 a 15 heures



    This situation note is established according to the information gained on March 17, 2011, at 3:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of all 6 reactors of the Fukushima I site (Dai-ichi) and of their spent fuel pools, as well as the situation of the reactors No. 1, 2, 3 and 4 of the Fukushima II site (Daini), and of the Onagawa and Tokai power plants is briefly presented with the progress of the accident management actions. (J.S.)

  4. 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

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


    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)

  5. Modifikace utěsnění víka primárního kolektoru PG VVER 440

    Blažková, Eva


    Cílem této diplomové práce je řešení problematiky, týkající se utěsnění víka primárního kolektoru parogenerátoru bloku jaderné elektrárny typu VVER 440. Tyto parogenerátory jsou těsněny v původním provedení niklovými kroužky. Modifikací stávajícího způsobu těsnění za nový druh těsnícího materiálu, zejména z expandovaného grafitu, lze podstatně snížit tlak v těsnící ploše a tím i napjatost svorníků a přírubového spoje. Nové řešení utěsnění spoje kolektoru a víka by mělo zvýšit životnost spoje ...

  6. 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

    Girardin, G.


    was paid to have each individual fuel sub-assembly and CA represented, in order to allow the analysis of local deformations of the 3D distributions of power and safety related parameters, such as the coolant, cladding and fuel temperatures. The validation of the coupled full core models was performed against reference ERANOS-VARIANT calculations. (author) [French] Le travail de recherche a ete mene dans le contexte du RNR-G de reference a 2400 MWth. Le but principal a ete de developper et de qualifier le design de l'assemblage de controle et le schema d'implantation correspondant a ce systeme. Le travail a ete realise en trois phases successives et complementaires: (1) la validation des outils neutroniques, (2) le developpement du design de l'assemblage de controle et les etudes statiques correspondantes, (3) les etudes du comportement dynamique du coeur durant des transitoires hypothetiques lies aux assemblages de controle. Pendant la premiere phase de la these, le reseau test de reference de PROTEUS utilise dans ces experiences, a ete reinterprete avec ERANOS et la bibliotheque nucleaire ajustee ERALIB1. De maniere complementaire, des calculs de references ont ete realises avec le code de Monte Carlo MCNPX, visant a verifier les resultats deterministes, et a analyser la sensibilite des resultats aux differentes modernes librairies. Pour les taux de reactions principaux, la nouvelle analyse du reseau de reference GCFR-PROTEUS donne generalement des resultats en bon accord - dans l'incertitude experimentale 1{sigma} - avec les resultats experimentaux, ainsi qu'avec les simulations de Monte Carlo. L'utilisation d'ERANOS-2.0/ERALIB1 comme outil neutronique de reference a ainsi pu etre demontree pour l'analyse du RNR-G. Dans une seconde phase de cette recherche, le design de l'assemblage de controle (AC) du GFR a ete developpe, base sur des calculs a caractere iteratif, neutroniques et thermo-hydrauliques. En premier, des

  7. Is it possible to recycle nuclear wastes? Costs, risks and stakes of the plutonium industry; Peut-on recycler les dechets nucleaires? Couts, risques et enjeux de l'industrie du plutonium



    This document, published by the French association 'Sortir du nucleaire' (Get out of nuclear), gives some information on the chain reaction from uranium to plutonium, the difference between reprocessing (which does not reduce waste volumes but multiply waste types) and recycling, the high risks associated with plutonium transport, La Hague as the most dangerous nuclear site in France, reprocessing as the alibi for the French nuclear industry, Areva as an expert in propaganda, reprocessing as an absurd world strategy, plutonium as a fuel for proliferation, the myth of unlimited energy with the breeder reactors, and so on

  8. Factors and uncertainties in the profitability of using nuclear energy in desalination of water; Facteurs et incertitudes de la rentabilite du recours a l'energie nucleaire dans le dessalement des eaux

    Thiriet, L.; Lievre, P. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires


    One of the economic advantages of nuclear energy consists of the small proportional element in its cost structure. Economies of scale favour the nuclear station as compared with the conventional thermal one, and when the demand for electricity and heat, in particular for desalination, are sufficient, nuclear energy may, subject to certain conditions, prove advantageous. The object of this paper is to discuss the validity of the conclusions reached according to the hypotheses adopted. In the first part, the different kind of uncertainties connected with technical, economic and financial data (the various transmission coefficients, the life of equipment according to the choice of materials, changes in prices, the form of price functions and interest rates), and with the various constraints, are examined and discussed. In the second part the uncertainties connected with the method of optimisation used and the criterion of selection adopted are examined and discussed. It is shown thereby that it is usually extremely difficult to assume absolutely the competitiveness, or conversely the non-competitiveness, of using nuclear energy in the desalination of water, and that a large number of aspects have to be carefully examined. (author) [French] On sait que l'un des avantages economiques de l'energie nucleaire reside dans la faible part proportionnelle dans la structure de son cout. Les economies d'echelle favorisent le nucleaire par rapport au thermique classique, et lorsque les demandes d'electricite et de chaleur, notamment pour le dessalement, sont suffisantes on peut envisager favorablement, sous certaines hypotheses, le recours a l'energie nucleaire. L'objet de cette communication est de discuter la validite des conclusions auxquelles on parvient selon les hypotheses envisagees. Dans une premiere partie, on etudie et on discute les differentes sortes d'incertitudes, liees aux donnees techniques, economiques et financieres (les divers

  9. Relationship between the nuclear resonance of cobalt metal and its ferromagnetic properties; Relations entre la resonance nucleaire du cobalt metallique et ses proprietes ferromagnetiques

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


    Theoretical study of nuclear magnetic resonance in ferromagnetic metals shows the near dependence of ferromagnetic properties and unusual feature of this nuclear resonance. This results from a strong interaction between nuclei and magnetic electrons. They excite the nuclei, and, in Bloch walls, submit them to a RF field much stronger than those directly applied. The parameters of the resonance are determined from wall movement and depend consequently of ferromagnetic constants. The theory is enable to provide quantitatively some peculiar effects, specially those of a continuous magnetic field and of temperature. Experimental study was made on cobalt powders, and is in good agreement with theory. However one must take the skin-effect into consideration and accordingly adjust, the theory. This can explain some observed divergences, as well as the influence at particles size and magnetic field over the line shape. Original informations have been obtained about some typical ferromagnetic properties of cobalt, when studying magnetic field effect, and it has been able to apply this method to other ferromagnetic materials. In consideration of the peculiar characteristics of this nuclear resonance, which occurs without external magnetic field and whose line width is large, new models of spectrographs have been realized and have permitted accurate measures of the line shape. The weak intensity of the signals obtained in some cases, has induced the elaboration of an original method of extraction whose theory and practical uses are described here. The whole of this experiment reveals the nuclear resonance as a strong way for the study of ferromagnetism, which is able to detect microscopic phenomenons, not easily accessible by classical methods. (author) [French] L'etude theorique de la resonance magnetique nucleaire dans les metaux ferromagnetiques revele l'etroite liaison entre les proprietes ferromagnetiques et l'aspect inhabituel de cette resonance. Ceci

  10. Optimized planning of in-service inspections of local flow-accelerated corrosion of pipeline elements used in the secondary coolant circuit of the VVER-440-based units at the Novovoronezh NPP

    Tomarov, G. V.; Povarov, V. P.; Shipkov, A. A.; Gromov, A. F.; Budanov, V. A.; Golubeva, T. N.


    Matters concerned with making efficient use of the information-analytical system on the flow-accelerated corrosion problem in setting up in-service examination of the metal of pipeline elements operating in the secondary coolant circuit of the VVER-440-based power units at the Novovoronezh NPP are considered. The principles used to select samples of pipeline elements in planning ultrasonic thickness measurements for timely revealing metal thinning due to flow-accelerated corrosion along with reducing the total amount of measurements in the condensate-feedwater path are discussed.

  11. Calculation of reactivities using ionization chamber currents with different sets of kinetic parameters for reduced scram system efficiency in the VVER-1000 of the third unit of the Kalinin nuclear power plant at the stage of physical start-up

    Zizin, M. N., E-mail: [Russian Research Centre Kurchatov Institute (Russian Federation); Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A. [JSC VNIIAES (Russian Federation)


    The effectiveness of the VVER-1000 reactor scram system is analyzed using ionization chamber currents with different sets of kinetic parameters with allowance for the isotopic composition in the calculation of these parameters. The most 'correct, aesthetically acceptable' results are obtained using the eight-group constants of the ROSFOND (BNAB-RF) library. The difference between the maximum and minimum values of the scram system effectiveness calculated with different sets of kinetic parameters slightly exceeds 2{beta}. The problems of introducing corrections due to spatial effects are not considered in this study.

  12. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Rahmani, Yashar, E-mail: [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)


    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  13. Some problems on the aqueous corrosion of structural materials in nuclear engineering; Problemes de corrosion aqueuse de materiaux de structure dans les constructions nucleaires

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


    The purpose of this report is to give a comprehensive view of some aqueous corrosion studies which have been carried out with various materials for utilization either in nuclear reactors or in irradiated fuel treatment plants. The various subjects are listed below. Austenitic Fe-Ni-Cr alloys: the behaviour of austenitic Fe-Ni-Cr alloys in nitric medium and in the presence of hexavalent chromium; the stress corrosion of austenitic alloys in alkaline media at high temperatures; the stress corrosion of austenitic Fe-Ni-Cr alloys in 650 C steam. Ferritic steels: corrosion of low alloy steels in water at 25 and 360 C; zirconium alloys; the behaviour of ultrapure zirconium in water and steam at high temperature. (authors) [French] On presente un ensemble d'etudes de corrosion en milieu aqueux effectuees sur des materiaux utilises, soit dans la construction des reacteurs soit pour la realisation des usines de traitement des combustibles irradies. Les differents sujets etudies sont les suivants. Les alliages austenitiques Fer-Nickel-Chrome: comportement d'alliages austenitiques fer-nickel-chrome en milieu nitrique en presence de chrome hexavalent; Corrosion sous contrainte d'alliages austenitiques dans les milieux alcalins a haute temperature; Corrosion sous contrainte dans la vapeur a 650 C d'alliages austenitiques fer-nickel-chrome. Les aciers ferritiques; Corrosion d'aciers faiblement allies dans l'eau a 25 et 360 C; le zirconium et ses alliages; Comportement du zirconium tres pur dans l'eau et la vapeur a haute temperature. (auteurs)

  14. Differential measurement of the earth's magnetic field by nuclear magnetic resonance; Mesure differentielle du champ magnetique terrestre par resonance magnetique nucleaire

    Robach, F. [Commissariat a l' Energie Atomique, 38 - Grenoble (France). Centre d' Etudes Nucleaires


    MNR transducers using proton dynamic polarisation allows to convert into a phase measurement any variation of the earth magnetic field. There exist several versions of the instrument corresponding to various models of MNR transducers, which the author analyses in detail, devoting an important place to influence of their alignment with respect to the earth's magnetic field. The sensibility obtained is of one hundredth of a gamma over a bandwidth of (0-0,1 Hz). - This instrument is designed for measuring field gradients in airborne magnetic surveying, for detecting nearly magnetic anomalies, and for distinguishing between nearly and distant magnetic phenomena. (author) [French] L'emploi de capteurs, bases sur la resonance magnetique nucleaire des protons en presence de polarisation dynamique, permet de traduire une difference de champ magnetique terrestre en une mesure de phase. L'appareil existe sous plusieurs versions avec des capteurs de modeles differents dont l'auteur fait une analyse detaillee en accordant une part importante a l'influence de l'orientation des capteurs par rapport au champ magnetique terrestre. La sensibilite est de 1/100 {gamma} pour une bande passante de (0 - 0,1 Hz). Cet appareil s'applique a la mesure du gradient en prospection magnetique aeroportee, a la detection d'anomalies magnetiques proches, a la differentiation d'effets magnetiques proches et lointains. (auteur)

  15. Damage caused to houses and equipment by underground nuclear explosions; Degats dus aux explosions nucleaires souterraines sur les habitations et les equipements

    Delort, F.; Guerrini, C. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes


    A description is given of the damaged caused to various structures, buildings, houses, mechanical equipment and electrical equipment by underground nuclear explosions in granite. For each type of equipment or building are given the limiting distances for a given degree of damage. These distances have been related to a parameter characterizing the movement of the medium; it is thus possible to generalize the results obtained in granite, for different media. The problem of estimating the damage caused at a greater distance from the explosion is considered. (authors) [French] Les degats sur diverses structures, constructions, habitations, equipements mecaniques et materiels electriques provoques par des explosions nucleaires souterraines dans le granite sont decrits. On a indique pour chaque type de materiel ou de construction, les distances limites correspondant a un degre de gravite de dommage observe. Ces distances ont ete reliees a un parametre caracterisant le mouvement du milieu, permettant ainsi de generaliser les resultats obtenus dans le granite, a differents milieux. Le probleme de la prevision des degats en zone lointaine a ete aborde. (auteurs)

  16. Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

    Parchevsky, V. M.; Guryanova, V. V.


    A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

  17. Types and analysis of defects in welding junctions of the header to steam generator shells on power-generating units with VVER-1000

    Ozhigov, L. S.; Voevodin, V. N.; Mitrofanov, A. S.; Vasilenko, R. L.


    Investigation objects were metal templates, which were cut during the repair of welding junction no. 111 (header to the steam generator shell) on a power-generating unit with VVER-1000 of the South-Ukraine NPP, and substances of mud depositions collected from walls of this junction. Investigations were carried out using metallography, optical microscopy, and scanning electron microscopy with energy dispersion microanalysis by an MMO-1600-AT metallurgical microscope and a JEOL JSM-7001F scanning electron microscope with the Shottky cathode. As a result of investigations in corrosion pits and mud depositions in the area of welding junction no. 111, iron and copper-enriched particles were revealed. It is shown that, when contacting with the steel header surface, these particles can form microgalvanic cells causing reactions of iron dissolution and the pit corrosion of metal. Nearby corrosion pits in metal are microcracks, which can be effect of the stress state of metal under corrosion pits along with revealed effects of twinning. The hypothesis is expressed that pitting corrosion of metal occurred during the first operation period of the power-generating unit in the ammonia water chemistry conditions (WCC). The formation of corrosion pits and nucleating cracks from them was stopped with the further operation under morpholine WCC. The absence of macrocracks in metal of templates verifies that, during operation, welding junction no. 111 operated under load conditions not exceeding the permissible ones by design requirements. The durability of the welding junction of the header to the steam generator shell significantly depends on the technological schedule of chemical cleaning and steam generator shut-down cooling.

  18. 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



    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

  19. Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

    Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.


    The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

  20. Narrow gap mechanised arc welding in nuclear components manufactured by AREVA NP; Le soudage mecanise a l'arc en chanfrein etroit dans les constructions nucleaires realisees par AREVA NP

    Peigney, A. [Departement Soudage AREVA Centre Technique - 71380 Saint-Marcel (France)


    Nuclear components require welds of irreproachable and reproducible quality. Moreover, for a given welding process, productivity requirements lead to reduce the volume of deposited metal and thus to use narrow gap design. In the shop, narrow gap Submerged Arc Welding process (SAW) is currently used on rotating parts in flat position for thicknesses up to 300 mm. Welding is performed with one or two wires in two passes per layer. In Gas Tungsten Arc Welding process (GTAW), multiple applications can be found because this process presents the advantage of allowing welding in all positions. Welding is performed in one or two passes per layer. The process is used in factory and on the nuclear sites for assembling new components but also for replacing components and for repairs. Presently, an increase of productivity of the process is sought through the use of hot wire and/or two wires. Concerning Gas Metal Arc Welding process (GMAW), its use is growing for nuclear components, including narrow gap applications. This process, limited in its applications in the past on account of the defects it generated, draws benefit from the progress of the welding generators. Then it is possible to use this efficient process for high security components such as those of nuclear systems. It is to be noted that the process is applicable in the various welding positions as it is the case for GTAW, while being more efficient than the latter. This paper presents the state of the art in the use of narrow gap mechanised arc welding processes by AREVA NP units. (author) [French] Les constructions nucleaires necessitent des soudures de qualite irreprochable et reproductible. Par ailleurs les imperatifs de productivite conduisent, pour un procede donne, a reduire le volume de metal a deposer et donc a utiliser des chanfreins etroits. En atelier, le soudage fil-flux en chanfrein etroit est couramment utilise sur des pieces tournantes en position a plat pour des epaisseurs atteignant 300 mm. On

  1. Nuclear energy; Le nucleaire



    This digest document was written by members of the union of associations of ex-members and retired people of the Areva group (UARGA). It gives a comprehensive overview of the nuclear industry world, starting from radioactivity and its applications, and going on with the fuel cycle (front-end, back-end, fuel reprocessing, transports), the nuclear reactors (PWR, BWR, Candu, HTR, generation 4 systems), the effluents from nuclear facilities, the nuclear wastes (processing, disposal), and the management and safety of nuclear activities. (J.S.)

  2. Nuclear wastes; Dechets nucleaires



    Here is made a general survey of the situation relative to radioactive wastes. The different kinds of radioactive wastes and the different way to store them are detailed. A comparative evaluation of the situation in France and in the world is made. The case of transport of radioactive wastes is tackled. (N.C.)

  3. The International Atomic Energy Agency: orientations for the 21. century. Nuclear facilities exploitation: three questions to Remy Carle. The civil nuclear and the electric power generation in Germany. Usa: the electric power marker deregulation and the perspective of the nuclear energy. The situation of nuclear energy in Japan. Finland.. debate about the 5. reactor. The electronuclear development in China. The last act of the swedish nuclear saga. The Korean nuclear programme. The civil nuclear energy in Eastern Europe... in brief; L'Agence Internationale de l'Energie Atomique: orientations pour le 21. siecle. Exploitation des centrales nucleaires: trois questions a Remy Carle. Le nucleaire civil et la production d'electricite en Allemagne. Etats-Unis: la deregulation du marche de l'electricite et les perspectives du nucleaire. La situation du nucleaire au Japon. Finlande... sur fond de debat a propos du 5. reacteur. Le developpement electronucleaire de la Chine.

    Carle, R.; Thiebaud, Ph. [CEA, Direction des Relations Interantionales, 75 - Paris (France); Heuraux, Ch.; Tinturier, B.; Lavergne, B. de [EDF, 75 - Paris (France); Forum Atomique Allemand, Bonn (Germany); Lavigne, J.J. [Ambassade de France au Japon (Japan); Soyer, B. [Ambassade de Chine en France (France); Edin, K.A. [Institut Politique et Social de Stockholm (Sweden); Chaucheprat, P. [Ambassade de France en Coree (Korea, Republic of)


    This issue is dedicated to nuclear programmes throughout the world. A few articles give an overview of the major trends of nuclear industry, from the standpoints of economy, industry, environment and international regulations. A few articles give an overview of the major trends of nuclear industry, from the standpoints of economy, industry, environment and international regulations. A few specific countries or groups of countries are then highlighted, together with a typical economic and political backgrounds. The countries dealt with are Germany, China, Korea, the U.S., Finland, Japan and Sweden. Some specific data are given on nuclear industry in Eastern Europe. (author)

  4. Activity report 1999; Rapport d'activites 1999



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

  5. Activity report 1998; Rapport d'activites 1998



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

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

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


    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

  7. Some Aspects of Diffusion Theory

    Pignedoli, A


    This title includes: V.C.A. Ferraro: Diffusion of ions in a plasma with applications to the ionosphere; P.C. Kendall: On the diffusion in the atmosphere and ionosphere; F. Henin: Kinetic equations and Brownian motion; T. Kahan: Theorie des reacteurs nucleaires: methodes de resolution perturbationnelles, interactives et variationnelles; C. Cattaneo: Sulla conduzione del calore; C. Agostinelli: Formule di Green per la diffusione del campo magnetico in un fluido elettricamente conduttore; A. Pignedoli: Transformational methods applied to some one-dimensional problems concerning the equations of t

  8. Micro manometer and pitot tube for measuring the velocity distribution in a natural convection water stream between two vertical parallel plates (1961); Micromano metre et tube de pitot destines a l'exploration du profil de vitesse dans un ecoulement d'eau de convection naturelle entre deux plaques verticales paralleles (1961)

    Santon, L.; Vernier, Ph. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires


    For heat transfer studies in certain cases of cooling in swimming-pool type nuclear reactors, a knowledge of the distribution of the velocities between two heating elements is of prime importance. A Pitot tube and a micro-manometer have been developed for making these measurements on an experimental model. (authors) [French] Pour l'etude du transfert de chaleur dans certains cas de refroidissement des reacteurs nucleaires du type piscine, la connaissance de la repartition des vitesses entre deux elements chauffants est primordiale. On a mis au point un tube de Pitot et un micromanometre pour effectuer ces mesures sur une maquette experimentale. (auteurs)

  9. Origin of elements of the Uranium-235 family observed in the Ellez river near the EL-4 experimental nuclear reactor in dismantling (Monts d'Arree- Finistere department); Origine des elements de la famille de l'uranium-235 observes dans la riviere Ellez a proximite du reacteur nucleaire experimental EL4 en cours de demantelement (Mont d'Arree - departement du Finistere). Resultats et premiers constats annee 2006



    In a previous study which concerned the catchment basin of the harbour of Brest, the A.C.R.O. put in evidence a marking by artificial radioelements around the power plant of Brennilis which can be imputed without ambiguities to the nuclear installation. It also put in evidence abnormalities concerning the natural radioactivity which justifies this new study. In the area of the Monts d'Arree, actinium 227 ({sup 227}Ac), non born by its ascendents which are {sup 235}U and {sup 231}Pa is observed. This phenomenon is characterized by mass activities superior to these ones of {sup 235}U and able to reach these ones of {sup 238}U. Its presence corresponds with the drainage of the Ellez river since the former channel of radioactive effluents releases from the nuclear power plant EL-4 up to the reservoir Saint-Herblot situated 6 km downstream. The strongest values of radioactivity are registered near the disused power plant, at this place a relationship exists between the level of actinium 227 and this one of the artificial radioactivity as it exists a relationship with the decay products of radon exhaled from the subsoil ({sup 210}Pb). But its presence is not limited to a part of the Ellez river, it is equally observed in terrestrial medium, in places in priori not influenced by the direct liquid effluents of the power plant. This place is situated at more than 4 km and without any connection with the Ellez waters. At this stage of the study, it is not possible to answer with certainty the question of the origin of this phenomenon. A new reorientation is considered indispensable to clarify definitively the origin of this unknown phenomenon in the scientific publications and the environmental monitoring. (N.C.)

  10. Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions; Etude de l'origine des elements de la famille de l'uranium-235 observes en exces dans les environs du reacteur nucleaire experimental EL4 en cours de demantelement. Enseignements retires a ce jour et conclusion



    This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

  11. Health physics problems in the context of the development of industrial uses of nuclear energy; Les problemes de radioprotection devant le developpement de l'utilisation industrielle de l'energie nucleaire

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


    fate of radioactive ions in the hydrosphere may be determined. - experiments and studies linked with the possibilities for disposal a underground storage of wastes. All these studies cover a complementary field, that of radioprotection, which forms the necessary link between nuclear safety proper, which concerns the actual operation of nuclear installations and medical supervision which directly concerns the health of the individual and the population. At the same time each one of the Health Physics disciplines possesses an independence which increases their efficiency on the safety level. (authors) [French] Le developpement de l'utilisation industrielle de l'energie nucleaire a mis l'accent sur la necessite de promouvoir une doctrine rationnelle et coherente en matiere de securite des installations atomiques et plus particulierement en matiere de securite des sites nucleaires. Le but principal de la securite est de diminuer, voire d'annuler les risques d'irradiation ou de contamination des travailleurs et de la population, consecutifs au fonctionnement normal ou accidentel des installations. La securite est obtenue par la mise en oeuvre d'un ensemble complexe de moyens. Parmi ces moyens les auteurs considerent essentiellement ceux qui presentent une certaine independance a l'egard du type d'installation: moyens d'alerte et de mesure et leur mise en oeuvre - preparation psychologique et technique des individus - evaluation de la capacite d'un site a absorber les rejets radioactifs sans dommage pour la population. Il est clair en effet que les consequences d'un accident sont diminuees par l'augmentation de la valeur des moyens de surveillance, de mesure et d'alerte, l'elevation du niveau technique et l'entrainement du personnel, l'education de la population. Ceci est particulierement vrai aupres d'une installation nucleaire ou l'etude des dangers presente un caractere abstrait et

  12. 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

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


    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)

  13. Contribution to the study of the evolution of radiation induced He in Be O; Etudes sur l'evolution de l'helium produit par reactions nucleaires dans l'oxyde de beryllium

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


    The purpose of this work in-pile investigation of He formed by (n, 2n) and (n, {alpha}) nuclear reactions, released from irradiated BeO, in the temperature range 1000 - 1350 deg. C. The experimental results show that, for an instantaneous neutron fast flux of 10{sup 13} n cm{sup -2} sec{sup -1}, an equilibrium is attempted, after several days, for a part lower than 20 per cent of the quantity of He formed, theoretically calculated from the neutron cross sections of nuclear reactions, and from the analytical form of the neutron fast flux, releases from the solid. The validity of the values of calculated helium and the gas chromatographic analytical method are also verified by dissolution of the BeO pellets in cryolite. A new fast neutron flux measuring method may be so defined. The discussion of the experimental results enables to establish that the processus of He release is characterized by two phenomena: the first one which controls the release of He atoms out of the solid and the second which reveals a capture processus, connected to the irradiation and probably due to the vacancies induced in the lattice. (author) [French] On etudie en pile le degagement de l'helium forme par reactions nucleaires (n, 2 n) et (n, {alpha}) dans l'oxyde de beryllium irradie entre 1000 et 1350 deg. C. Les resultats experimentaux montrant que, pour un flux rapide instantane de 10{sup 13} n{sub r} cm{sup -2} s{sup -1}, on aboutit, au bout de quelques jours, a un etat d'equilibre pour lequel une partie, inferieure a 20 pour cent de la quantite d'helium forme, calculee theoriquement a partir des sections efficaces des reactions nucleaires et de la forme analytique du flux rapide, s'echappe du solide. On verifie egalement par dissolution des echantillons de BeO dans la cryolithe la validite du calcul de l'helium et de la methode de dosage par chromatographie en phase gazeuse. On peut ainsi definir une methode nouvelle de mesure des flux rapides. La

  14. Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

    Rebreyend, P.; Burel, J.P. [Rolls-Royce Civil Nuclear SAS (France); Spoc, J. [Skoda JS (Czech republic); Karasek, A. [CEZ a.s.(Czech republic)


    Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

  15. Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation: material: 15 Ch2MFA {approx} 0, 15 C-2,5 Cr-0, 7Mo-0,3 V; Veraenderungen der Mikrostruktur in Grundwerkstoff, WEZ und Schweissgut eines 440-VVER-Reaktordruckbehaelters, verursacht durch Neutronenbestrahlung im langzeitigen Betrieb; Werkstoff: 15 Ch2MFA {approx} 0,15 C-2,5 Cr-0, 7Mo-0,3 V

    Maussner, G.; Scharf, L.; Langer, R. [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Gurovich, B. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)


    Within the scope of the Tacis `91/1.1 project of the European Community, ``Reactor Vessel Embrittlement``, specimens were taken from the heavily irradiated circumferential welds of a VVER pressure vessel. The cumulated fast neutron fluence in the specimens amounts to up to 6.5 x 10{sup 19} cm{sup -}2 (E > 0.5 MeV). For the multi-laboratory, coordinated study, the specimens were cutted for mechanical testing as well as analytical, microstructural and microanalytical examinations in the base metal, HAZ and weld metal with respect to the effects of reactor operatio and post-irradiation annealing as well as thermal treatment (475 C, 560 C). The analytical transmission electron microscopy (200 kV) revealed the alterations found in the mechanical properties to be due to the formation of black dots and irradiation-induced segregations and accumulations of copper and carbides. These effects, caused by operation, (neutron radiation, temperature), are much more significant in the HAZ than in the base metal. (orig./CB) [Deutsch] Im Rahmen des von der Europaeischen Union beauftragten Tacis `91/1.1 Programms `Reactor Vessel Embrittlement` wurden Bohrkerne aus dem hochbestrahlten Rundnahtbereich eines VVER-Reaktordruckbehaelters entnommen. Die kumulierte schnelle Neutronenfluenz in diesen Proben betraegt bis zu 6,5 x 10{sup 19} cm{sup -2} (E>0,5 MeV). In einer gemeinschaftlichen Untersuchung wurden mechanisch-technologische, chemische sowie mirkostrukturelle Untersuchungen an Grundwerkstoff-, WEZ- und Schweissgutproben im vergleichbaren Ausgangs-, bestrahlten und anschliessend waermebehandelten (475 C, 560 C) Werkstoffzustand durchgefuehrt. Die analytische Durchstrahlelektronenmikroskopie (200 kV) laesst als Ursache fuer die festgestellten Veraenderungen der mechanischen Eigenschaften die Bildung von Versetzungsringen (black dots) sowie von bestrahlungsinduzierten Ausscheidungen und Anreicherungen von Kupfer in den Karbiden erkennen. Diese Effekte, als Folge der betrieblichen

  16. Reactor Bolshoi Moshchnosti Kalani; Reacteurs RBMK

    Bastien, D. [Conservatoire National des Arts et Metiers (CNAM), 75 - Paris (France)


    The Reactor Bolshoi Molshchnosti Kalani (RBMK) are pressure tubes reactor, boiling light water cooled. Exported since 1990 from the ex-USSR, they are today in three independent countries: Russian, Ukraine and Lithuania. Since this date, data exchange with the occident allowed the better knowledge of this reactor type. The design, the technical description (core, fuel, primary system), the safety and the improvement since Chernobyl are detailed. (A.L.B.)

  17. Heavy Water Reactor; Reacteurs a eau lourde

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)


    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  18. The EPR reactor; Le reacteur EPR

    Lacoste, A.C.; Dupuy, Ph.; Gupta, O.; Perez, J.R.; Emond, D. [Direction Generale de la Surete Nucleaire et de la Radioprotection, 75 - Paris (France); Cererino, G.; Rousseau, J.M.; Jeffroy, F.; Evrard, J.M. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Dir. de la Surete des Reacteurs, 92 - Clamart (France); Seiler, J.M. [CEA Cadarache (DEN/DTN), 13 - Saint-Paul-lez-Durance (France); Azarian, G. [FRAMATOME ANP, 92 - Paris-La-Defence (France); Chaumont, B. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DSR), 92 - Fontenay-aux-Roses (France); Dubail, A. [Electricite de France (EDF), 78 - Chatou (France); Fischer, M. [Framatome ANP GmbH, Erlangen (Germany); Tiippana, P.; Hyvarinen, J. [Stuk, Autorite de Surete Nucleaire et de Radioprotection (Finland); Zaleski, C.P.; Meritet, S. [Paris-9 Univ. Dauphine, Centre de Geopolitique de l' Energie et des Matieres Premieres, 75 (France); Iglesias, F.; Vincent, C. [Direction Generale de l' Energie et des Matieres Premieres, 75 - Paris (France); Massart, S.; Graillat, G. [Electricite de France (EDF), 75 - Paris (France); Esteve, B. [AREVA/Framatome, 75 - Paris (France); Mansillon, Y. [Commission Nationale de Debat Public, 75 - Paris (France); Gatinol, C. [Assemblee Nationale, 75 - Paris (France); Carre, F. [CEA, Dir. de Programme Systemes du Futur, France (France)


    This document reviews economical and environmental aspects of the EPR project. The following topics are discussed: role and point of view of the French Nuclear Safety Authority on EPR, control of design and manufacturing of EPR by the French Nuclear Safety Authority, assessment by IRSN of EPR safety, research and development in support of EPR, STUK safety review of EPR design, standpoint on EPR, the place of EPR in the French energy policy, the place of EPR in EDF strategy, EPR spearhead of nuclear rebirth, the public debate, the local stakes concerning the building of EPR in France at Flamanville (Manche) and the research on fourth generation reactors. (A.L.B.)

  19. Thermohydraulics of reactors; Thermohydraulique des reacteurs

    Delhaye, J.M


    This scientific and technical handbook about PWR reactors thermohydraulics is the result of many years of teaching in the framework of the CEA-INSTN's atomic engineering training courses, in engineering schools and during continuing training activities. Its main goal is to present in a rigorous and pedagogical way the basic knowledge necessary for the understanding and modeling of single phase and two-phase thermohydraulic phenomena encountered during the design and operation of nuclear reactors. In particular, heat transfers in two-phase flows are presented in a detailed way. Most chapters include some nuclear engineering examples of application of the studied concepts, and some exercises aiming at mastering these concepts. Each example or exercise is accompanied by its detailed solution. Content: - thermohydraulic characteristics of reactors; - design and thermal dimensioning of reactors; - thermal engineering of the fuel element; - two-phase flow configurations in ducts; - recalls about single-phase flow equations; - basic equations for two-phase flows; - modeling of two-phase flows inside ducts; - pressure drops in ducts; - boiling and vapor condensation heat transfers; - two-phase flow instabilities in ducts; - two-phase flow blocking; thermohydraulics of naval propulsion reactors.

  20. Nuclear medicine; La medecine nucleaire

    Sibille, L. [Hopital Lapeyronie CHU Montpellier, Medecine Nucleaire, 34 - Montpellier (France); Nalda, E.; Collombier, L.; Kotzki, P.O.; Boudousq, V. [CHU de Nimes, Service de Medecine Nucleaire et de biophysique, 30 - Nimes (France)


    Nuclear medicine is a medical specialty using the properties of radioactivity. Radioactive markers associated with vectors are used as a tracer or radiopharmaceutical for diagnostic purposes and/or therapy. Since its birth more than half a century ago, it has become essential in the care of many patients, particularly in oncology. After some definitions, this paper presents the main nuclear techniques - imaging for diagnostic, radiopharmaceuticals as therapeutic agents, intra-operative detection, technique of radioimmunoassay - and the future of this field. (authors)

  1. Information sur les technologies nucleaires


    "La base Inis (International Nuclear Information System) de l'Agence Internationale d'Energie Atomique (IAEA) dispose desormais de liens vers les textes complets de documents internationaux" (1 paragraph).

  2. Pressure vessel calculations for VVER-440 reactors.

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E


    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  3. Pressure Vessel Calculations for VVER-440 Reactors

    Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.


    Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

  4. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.


    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  5. Decision nr 2011-DC-0231 of the 4 July 2001 by the ASN specifying to EDF - limited company (EDF-SA) - the additional specifications applicable to the Fessenheim (Haut-Rhin) electronuclear site according to the conclusions of the third safety re-examination of reactor nr 1 of the nr 75 INB; Decision de l'Autorite de surete nucleaire n. 2011-DC-0231 du 4 juillet 2011 fixant a Electricite de France - Societe Anonyme (EDF-SA) les prescriptions complementaires applicables au site electronucleaire de Fessenheim (Haut Rhin) au vu des conclusions du troisieme reexamen de surete du reacteur n.1 de l'INB n.75



    This document defines the specifications EDF must comply with for the exploitation of the Fessenheim reactor nr 1. The specifications concern the safety policy and management for operations submitted to a declaration or to an agreement by the ASN, the management of accident risks (use of radioactive materials or susceptible to generate a nuclear reaction, other risks), the management and elimination of wastes and used fuels in a basic nuclear installation (waste and used fuel warehousing), the management of accidental chemical pollutions, the management of water sampling and effluent releases, the production of wastes within the installation

  6. Cooling and spreading of corium during its fall into water in a pressurised water nuclear plant severe accident: description of mechanical and thermal interactions in a three phase flow during spreading of cold or heated spheres in a liquid pool; Refroidissement et dispersion du corium lors de sa chute dans l'eau pendant un accident severe de reacteur nucleaire a eau pressurisee: description des interactions mecaniques et thermiques en ecoulement triphasique lors de la dispersion de spheres solides froides ou chaudes dans un bain liquide

    Duplat, F


    In the frame of nuclear safety studies about corium and water interactions, we address spreading and cooling stage of corium fragments in a liquid pool. Considering the complexity of encountered flow regimes and mechanical and thermal interactions coupling, modelling validation is based on a thermal-hydraulic computer code (MC3D). A bibliographical study shows that classical modelling of three phase flow is based on constitutive laws already established in the case of two phase flow. The present study states a complete analysis of BILLEAU experiments and defines a characterisation method for a sphere cloud. Some complementary QUEOS experiments are also described. Mechanical interaction terms such as added mass, lift and turbulent dispersion have been presented in the frame of a three phase flow and their influence has been tested in numerical simulations of BILLEAU tests. The effect of film vapour overheat, as well as particle diameter evolution have been studied. Moreover a radiative heat transfer modelling developed in Karlsruhe research centre (FZK) has been analysed and completed. Numerical simulations achieved for this study show that mechanical and thermal behaviour of the system are actually coupled. Taking into account lift and turbulent dispersion terms as well as heat transfer modifications all wed better results. This study also presents some considerations about flow regimes identification as a preliminary for studies about numerical diffusion that was already estimated in the present state of the computer code MC3D. (author)

  7. Contribution to the study of {beta} disintegration and of nuclear structure using experiments on certain {beta}-{gamma} cascades: 198{sub Au}, 86{sub Rb}, 170{sub Tm}; Contribution a l'etude de la desintegration beta et a l'etude de la structure nucleaire a l'aide d'experiences sur certaines cascades beta-gamma: 198{sub Au}, 86{sub Rb}, 170{sub Tm}

    Lachkar, J. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes; Paris-11 Univ., fabulte des Sciences 91 - Orsay (France)


    {beta}{gamma} directional angular correlations and shapes of inner beta spectra leading to the first excited level of the final nucleus enable one to determine the nuclear matrix elements typical of the {beta} transition. In the three observed first forbidden cases: {sup 170}Tm, {sup 86}Rb, {sup 198}Au, these matrix elements do not confirm the independent shell model theory. Other hypotheses are then suggested and discussed. (author) [French] Les experiences de correlation angulaire {beta}{gamma} et l'etude du spectre {beta} conduisant au premier niveau excite du noyau final permettent de determiner les elements de matrices nucleaires caracteristiques de cette transition. Dans les trois cas etudies (transitions une fois interdites): {sup 170}Tm, {sup 86}Rb, {sup 198}Au, ces elements de matrices ne peuvent etre retrouves a l'aide du modele en couches et a particules independantes. D'autres hypotheses sont alors emises et discutees. (auteur)

  8. Activity report 1998/1999; Rapport d'activite 1998/1999



    The aim of this report is to outline the main developments of the ''Departement d'Etudes des Combustibles'' (DEC), during the period 1998-1999: developments in terms of structure, staff, scientific research programs, publications and contracts. DEC is one of the department of the ''Departement des Reacteurs Nucleaires'' (DRN), itself depending of the CEA Institution. This report is divided in three parts. The first part gathers information on staff, budget and management. The part 2 presents the scientific programs in the domains of the nuclear Park competition improvement and renewal, the minor actinides incineration, thermo-mechanical codes and atomic calculations on fuel. The last part is devoted to publications,communication and training. (A.L.B.)

  9. Scientific report 1998; Rapport scientifique 1998



    The aim of this report is to outline the main developments of the ''Departement des Reacteurs Nucleaires'', (DRN) during the year 1998. DRN is one of the CEA Institution. This report is divided in three main parts: the DRN scientific programs, the scientific and technical publications (with abstracts in english) and economic data on staff, budget and communication. Main results of the Department, for the year 1998, are presented giving information on the reactors technology and safety, the neutronics, the transmutation and the hybrid systems, the dismantling and the sites improvement, the nuclear accidents, the nuclear matter transport, the thermonuclear fusion safety, the fuel cladding materials and radioactive waste control. (A.L.B.)

  10. Contribution to the study of molecular movements in cyclohexane by electron spin resonance and electron-nuclear double resonance using a radical probe; Contribution a l'etude des mouvements moleculaires dans le cyclohexane par resonance paramagnetique electronique et double resonance electronique-nucleaire a l'aide d'une sonde radicalaire

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


    Solutions of stable free radicals of the nitroxide type have been studied as a function of temperature. In the plastic or globular state, the cyclohexane molecules have rapid rotational and diffusional movements. They transmit this movement to dissolved free radicals. Conversely, measurements by electron spin resonance of the absolute movement of the radicals, and by electron nuclear double resonance of their movement relative to the cyclohexane molecules give very precise methods for local analyses of the movement present in the cyclohexane matrix. The principle of these techniques makes up the 'radical probe method'. (author) [French] Des solutions de radicaux libres stables, du type nitroxyde dans le cyclohexane ont ete etudiees, en fonction de la temperature. Les molecules de cyclohexane, dans l'etat plastique ou globulaire, sont animees de mouvements rapides de rotation sur elles-memes et de diffusion. Elles transmettent leur mobilite aux radicaux libres dissous. Reciproquement, la mesure du mouvement absolu des radicaux, a l'aide de la resonance paramagnetique electronique, et celle du mouvement relatif des radicaux et des molecules de cyclohexane par double resonance electronique-nucleaire, constituent des methodes tres precises pour analyser localement les mouvements presents dans la matrice de cyclohexane. Ce principe et ces techniques constituent la 'methode de la sonde radicalaire'. (auteur)

  11. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.


    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  12. Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors; Etude des materiaux sacrificiels absorbants et diluants pour le controle de la reactivite dans le cas d'un accidnet hypothetique de fusion du coeur de reacteurs de quatrieme generation

    Plevacova, K.


    In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B{sub 4}C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B{sub 4}C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B{sub 4}C - UO{sub 2} system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B{sub 4}C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu{sub 2}O{sub 3} or HfO{sub 2} as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al{sub 2}O{sub 3} - HfO{sub 2} and Al{sub 2}O{sub 3} - Eu{sub 2}O{sub 3} were preselected. These systems being completely unknown to date at high temperature in association with UO{sub 2}, first points on the corresponding ternary phase diagrams were researched. Contrary to Al{sub 2}O{sub 3} - Eu{sub 2}O{sub 3} - UO{sub 2} system, the Al{sub 2}O{sub 3} - HfO{sub 2} - UO{sub 2} mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [French] Resume: Afin de limiter les consequences d'un accident grave avec la fusion du coeur dans un reacteur a neutrons rapides de generation IV refroidi au sodium, la recriticite doit

  13. Nuclear era. L'ere nucleaire

    Leclercq, J.


    This book is a guide in the space and in the time, along the nuclear energy history and through all the industrial installations such as nuclear power plants and the associated plants. The main points developed in this book are the following ones: the nuclear energy in its historic perspective, the variety of reactors, safety and environment, architecture and large engineering, installation of the reactor components and associated machines, nuclear fuels, and the nuclear energy in the electricity service.

  14. Nuclear Safety Charter; Charte Surete Nucleaire



    The AREVA 'Values Charter' reaffirmed the priority that must be given to the requirement for a very high level of safety, which applies in particular to the nuclear field. The purpose of this Nuclear Safety Charter is to set forth the group's commitments in the field of nuclear safety and radiation protection so as to ensure that this requirement is met throughout the life cycle of the facilities. It should enable each of us, in carrying out our duties, to commit to this requirement personally, for the company, and for all stakeholders. These commitments are anchored in organizational and action principles and in complete transparency. They build on a safety culture shared by all personnel and maintained by periodic refresher training. They are implemented through Safety, Health, and Environmental management systems. The purpose of these commitments, beyond strict compliance with the laws and regulations in force in countries in which we operate as a group, is to foster a continuous improvement initiative aimed at continually enhancing our overall performance as a group. Content: 1 - Organization: responsibility of the group's executive management and subsidiaries, prime responsibility of the operator, a system of clearly defined responsibilities that draws on skilled support and on independent control of operating personnel, the general inspectorate: a shared expertise and an independent control of the operating organization, an organization that can be adapted for emergency management. 2 - Action principles: nuclear safety applies to every stage in the plant life cycle, lessons learned are analyzed and capitalized through the continuous improvement initiative, analyzing risks in advance is the basis of Areva's safety culture, employees are empowered to improve nuclear Safety, the group is committed to a voluntary radiation protection initiative And a sustained effort in reducing waste and effluent from facility Operations, employees and subcontractors are treated alike, and a high level Of know-how is supported by training and skills maintenance. 3 - Transparency and reporting: Areva endeavors to provide reliable and relevant information enabling an objective assessment of the status of nuclear safety in its facilities (incident reporting, annual report of the general inspectorate, annual reporting on occupational safety during nuclear facility operations). 4 - Glossary.

  15. Nuclear wastes management; Gestion des dechets nucleaires



    This document is the proceedings of the debate that took place at the French Senate on April 13, 2005 about the long-term French policy of radioactive wastes management. The different points tackled during the debate concern: the 3 axes of research of the 1991 law, the public acceptance about the implementation of repositories, the regional economic impact, the cost and financing, the lack of experience feedback, the reversibility or irreversibility of the storage, the share of nuclear energy in the sustainable development policy, the European Pressurized Reactor (EPR) project, the privatization of Electricite de France (EdF) etc. (J.S.)

  16. Nuclear energy; Le dossier du nucleaire

    Pierret, Ch. [Ministere de l' Industrie et de l' Amenagement du Territoire, 75 - Paris (France); Bacher, P.; Tanguy, P. [and others


    11 contributions written by experts are gathered in this document dedicated to nuclear energy (N.E): 1) could we live without N.E?, 2) do we have to fear N.E?, 3) a critical point of view on N.E, 4) the sanitary consequences of Chernobyl accident, 5) the impact of low radiation doses, 6) the management of a nuclear power plant, 7) the optimisation of wastes at The Hague facility, 8) the management of radioactive wastes in France, 9) the next generation of nuclear reactors, 10) N.E and public acceptance, and 11) is N.E the energy for the future? (A.C.)

  17. The nuclear proliferation; La proliferation nucleaire

    Gere, F. [Ecole Polytechnique, 91 - Palaiseau (France)


    In this book is detailed the beginning of nuclear military power, with the first bomb of Hiroshima, the different ways of getting uranium 235 and plutonium 239, and how the first countries (Usa, Ussr, China, United kingdom, France) got nuclear weapons. Then the most important part is reviewed with the details of non-proliferation treaty and the creation of IAEA to promote civilian nuclear power in the world and to control the use of plutonium and uranium in nuclear power plants. The cases of countries who reached the atom mastery, such Israel, South Africa, Pakistan, Iraq, North Korea, Argentina, Brazil, Iran, Algeria, Taiwan and the reasons which they wanted nuclear weapon for or why they gave up, are exposed.

  18. The nuclear refugees; Les refugies du nucleaire

    Linton, M.


    The authors propose a report on the various situations of people who had to be evacuated after the Fukushima accident. Along with examples of people who left their homes with taking with them a single object, the authors describe and comment how this evacuation occurred, the problems faced by the authorities for refugee reception and accommodation. This evacuation has been either organised or spontaneous. Hospitals had to be evacuated as well. Then, local authorities faced food shortage. Some animals have been saved, other starved to death. Dead animals are covered with lime. Dead bodies are decontaminated before being given back to families. Tests are regularly performed to assess people contamination. A second article discussed the bad news concerning the different Fukushima reactors with their melted cores. The geophysical aspects of the earthquake are evoked in a last article

  19. The nuclear threat; La menace nucleaire

    Tertrais, Bruno


    For a long time, a small group of big powers has been the only holder of nuclear weapons (US, USSR, Great Britain, France and China). Since then, new weapons have come out on the geopolitical scene: Israel, India, Pakistan, and some others remain uncertain and generate a worrying atmosphere (North Korea, Iran..). But what is the real risk with nuclear proliferation? Should we dread about it? Is nuclear terrorism a real threat? What are the political stakes of nuclear weapons? Is disarmament a real solution? These are some of the questions that the author answers in a precise and clear manner in this book. Contents: 1 - from monopoly to proliferation: who owns nuclear weapons today, why is it so coveted, is it easy to make one?; 2 - the newcomers: what do we really know about the Iranian nuclear programme, Iran and North Korea: between negotiation and confrontation; 3 - international control and regulation: do we have reliable information, how do we know what we know, Iraq: was there a 'lie' somewhere, who are the states who have renounced nuclear weapons?; 4 - the future: is there still a nuclear warfare risk, what if Pakistani weapons fall into islamic hands, is nuclear terrorism a fantasy or a real risk?

  20. Nuclear Safety. 1997; Surete Nucleaire. 1997



    A quick review of the nuclear safety at EDF may be summarized as follows: - the nuclear safety at EDF maintains at a rather good standard; - none of the incidents that took place has had any direct impact upon safety; - the availability remained good; - initiation of the floor 4 reactor generation (N4 unit - 1450 MW) ensued without major difficulties (the Civaux 1 NPP has been coupled to the power network at 24 december 1997); - the analysis of the incidents interesting from the safety point of view presents many similarities with earlier ones. Significant progress has been recorded in promoting actively and directly a safe operation by making visible, evident and concrete the exertion of the nuclear operation responsibility and its control by the hierarchy. The report develops the following chapters and subjects: 1. An overview on 1997; 1.1. The technical issues of the nuclear sector; 1.2. General performances in safety; 1.3. The main incidents; 1.4. Wastes and radiation protection; 2. Nuclear safety management; 2.1. Dynamics and results; 2.2. Ameliorations to be consolidated; 3. Other important issues in safety; 3.1. Probabilistic safety studies; 3.2. Approach for safety re-evaluation; 3.3. The network safety; 3.4. Crisis management; 3.5. The Lifetime program; 3.6. PWR; 3.7. Documentation; 3.8. Competence; 4. Safety management in the future; 4.1. An open future; 4.2. The fast neutron NPP at Creys-Malville; 4.3. Stabilization of the PWR reference frame; 4.4. Implementing the EURATOM directive regarding the radiation protection standards; 4.5. Development of biomedical research and epidemiological studies; 4.6. New regulations concerning the liquid and gaseous effluents; 5. Visions of an open future; 5.1. Alternative views upon safety ay EDF; 5.2. Safety authority; 5.3. International considerations; 5.4. What happens abroad; 5.5. References from non-nuclear domain. Four appendices are added referring to policy of safety management, policy of human factors in NPPs, technical incidents, and human factor incidents.

  1. Nuclear and radioactivity; Nucleaire et radioactivite



    Among the industrial risks of nuclear facilities, the nuclear risk is often associated to the Chernobyl accident. This paper presents the nuclear major risk in a french PWR type power plant, with consequences on the personnel, the surrounding population and the environment. (A.L.B.)

  2. The Fukushima accident; Accident nucleaire a Fukushima

    Delbecq, D.


    The Fukushima accident is characterized by a sequence of natural disasters: earthquake and tsunamis that deprived simultaneously 3 reactors from cooling and electrical power for quite a long time. A series of hydrogen explosion has added to the mess. Experts agree to say that certainly nuclear fuel has melt to form corium in all 3 reactors. The accident has contaminated tens of thousand acres of land around the plant and has jeopardized local coastal fishery. The human toll is unexpectedly low: no direct casualty in the population but several suicides among the people that was forced to leave their home. 5 people from the plant staff died certainly from the consequences of the tsunami. (A.C.)

  3. Nuclear energy; L'energie nucleaire

    Reuss, Paul


    With simple and accessible explanations, this book presents the physical principles, the history and industrial developments of nuclear energy. More than 25 years after the Chernobyl accidents and few months only after the Fukushima one, it discusses the pros and cons of this energy source with its assets and its risks. (J.S.)

  4. Ex-core instrumentation; Instrumentation hors coeur des reacteurs

    Burel, J.P. [Schneider Electric S.A., 92 - Boulogne-Billancourt (France)


    The safety and the control of the reactor need to master the nuclear power from the core radiation measurement. According to the reactor dimensions and conception, the nuclear parameters monitoring is realized through two instrumentation systems: the ex-core instrumentation system which use detector placed outside of the core and the in-core instrumentation system. This paper deals with the ex-core systems based on neutronic measurements and details the detectors choice, the treatment circuits, data processing, realizations in France and also example of the Wwr reactor instrumentation. (A.L.B.)

  5. China experimental fast reactor; Le reacteur rapide experimental chinois

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)


    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

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



    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.)

  7. Mise au point d'un reacteur epitaxial CBE

    Pelletier, Hubert

    Ce projet de maîtrise consiste à l'asservissement et la mise en marche d'un réacteur d'épi-taxie par jets chimiques au Laboratoire d'Épitaxie Avancée de l'Université de Sherbrooke. Le réacteur sert à la croissance dans l'ultravide de matériaux semi-conducteurs tels que l'arséniure de gallium (GaAs) et le phosphure d'indium-gallium (GaInP). La programmation LabVIEW™ et du matériel informatique de National Instruments sont utilisés pour asservir le réacteur. Le contrôle de la température de l'échantillon et de la pression de contrôle des réactifs de croissance dans le réacteur est assuré par des boucles de rétroaction. Ainsi, la température de l'échantillon est stabilisée à ±0, 4 °C, alors que les pressions de contrôle de gaz peuvent être modulées sur un ordre de grandeur en 2 à 4 secondes, et stabilisées à ±0, 002 Torr. Le système de pompage du réacteur a été amélioré suite à des mesures de vitesse de pompage d'une pompe cryogénique. Ces mesures révèlent une dégradation sur plus d'un ordre de grandeur de son pompage d'hydrogène avec l'opération à long terme. Le remplacement de la pompe cryogénique par une pompe turbo-moléculaire comme pompe principale a permis d'améliorer la fiabilité du système de pompage du système sous vide. D'autre part, la conductance du système d'acheminement de gaz et d'injection a été augmentée afin de réduire un effet mémoire des sources le système et faciliter la croissance de matériaux ternaires. Ainsi, des croissances de GaAs (100) sur substrat de même nature ont été effectuées et ont révélé un matériau de bonne qualité. Sa rugosité moyenne de 0,17 nm, mesurée par microscopie à force atomique, est très faible selon la littérature. De plus, une mobilité élevée des porteurs est obtenue à fort dopage au silicium, au tellure et au carbone, notamment une mobilité de 42 ± 9 cm2V -1s-1 des porteurs majoritaires '(trous) lors du dopage au carbone à 1, 5 · 1019 cm-3, en accord avec la courbe théorique. La croissance du matériau ternaire GaInP a aussi été réalisée en accord de maille avec le substrat de GaAs, et avec une rugosité de 0, 96 nm. Ceci constitue un premier pas dans la croissance d'alliages ternaires au laboratoire. Finalement, la mise en marche du réacteur d'épitaxie par jets chimiques permet maintenant à cinq étudiants gradués de faire progresser des projets reliés directement à la croissance épitaxiale au Laboratoire d'Épitaxie Avancée de l'Université de Sherbrooke. Mots-clés : Épitaxie par jets chimiques; Chemical beam epitaxy; CBE; MOMBE; GaAs; GaInP; LabVIEW; Théorie du vide.

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

    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


    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.)

  9. Development of Ultrasonic Probes for VVER-1000 Type Reactor

    WANG; Hua-cai; ZHU; Xin-xin; YIN; Zhen-guo; LIANG; Zheng-qiang


    Over a long period of time,the majority of domestic power plant damaged fuel rod assemblies monitoring system,mostly uses NDT methods such as sipping test,visual inspection and eddy current inspection.These methods all have some shortcomings,such as the visual inspection method can only inspect the outside fuel rods of the assemblies and the visual inspection can only

  10. Contribution to the determination of Sb-Ag-Cu-Ga-Mo-Zn using 14 MeV neutron activation; Contribution au dosage de Sb-Ag-Cu-Ga-Mo-Zn par activation aux neutrons de 14 MeV

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


    By using, 14 MeV, neutron irradiation it is possible to extend the field of application of neutron radio-activation analysis, in particular to the case of light elements. For, many other elements it can replace in-pile irradiation thereby making it possible, thanks to portable 14 MeV neutron generators, to carry out radio-activation analyses away from nuclear-research c e n t r e s. With a view to applying this analytical technique to routine work, we have developed some rapid chemical separation methods in order to make possible the determination of several elements which after exposure to fast neutrons, produce {beta} emitting nuclides which cannot be differentiated by a simple instrumental study, the emitted radiation being of the same type and of similar half-life the two cases. (author) [French] L'irradiation au moyen de neutrons de 14 MeV permet d'etendre le domaine d'application de l'analyse par radioactivation neutronique, en particulier aux elements legers. Cependant pour de nombreux autres elements elle peut remplacer l'irradiation en reacteur nucleaire permettant ainsi grace aux ensembles portables producteurs de neutrons de 14 MeV, l'extension de l'analyse par radioactivation a l'exterieur des centres d'etudes nucleaires. Dans le but d'appliquer cette methode d'analyse a des travaux de routine, nous avons mis au point des separations chimiques rapides, afin de permettre le dosage de quelques elements qui par irradiation aux neutrons rapides, engendrent des nucleides emetteurs {beta} qu'une simple etude instrumentale ne peut differencier en raison de l'identite de leur rayonnement et de leurs periodes radioactives trop proches. (auteur)

  11. Calculation of reactivity by digital processing; Calcul de la reactivite par traitement numerique

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


    With a view to exploring the new possibilities offered by digital techniques, a description is given of the optimum theoretical conditions of a computation of the realtime reactivity using counting samples (obtained from a nuclear reactor). The degree to which these optimum conditions can be attained depends on the complexity of the processing which can be accepted. A compromise thus has to be made between the accuracy required and the simplicity of the equipment carrying out the processing. An example is given, using a relatively simple structure, which gives an idea of the accuracy of the results obtained over a wide range of reactor power. (author) [French] Dans le but d'explorer les possibilites nouvelles des techniques numeriques, on decrit les conditions theoriques optimales d'un calcul de la reactivite en temps reel a partir d'echantillons de comptage (en provenance d'un reacteur nucleaire). Ces conditions optimales peuvent etre approchees d'autant mieux que l'on accepte un traitement plus complexe. Un compromis est donc a faire entre la precision desiree et la simplicite du materiel assurant le traitement. Un exemple adoptant une structure de complexite reduite permet de juger de la precision des resultats obtenus sur une importante plage d'evolution de la puissance. (auteur)

  12. The RES reactor, a test reactor for naval propulsion; Le reacteur d'essais RES, reacteur d'essais de la propulsion navale

    Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 13 - Aix en Provence (France)


    The RES, the new test reactor for naval propulsion, will replace the RNG that nears the end of its operating life after 30 years in service. The main asset of a land-based installation is to provide an in-core instrumented reactor while the on-board system must stay as simple as possible for robustness reasons. The objective of the RES is fivefold: 1) to foresee and help solving problems likely to happen on on-board reactor, 2) to validate nuclear fuels and reactor systems for naval propulsion, 3) to validate reactor system and equipment for the Barracuda submarine program, 4) to upgrade the on-ground facility located at Cadarache, and 5) to provide the Cea with a new capacity for the storing of spent fuels from naval propulsion systems and from Cea research reactors. The RES facility is made of 2 parts: one that houses the reactor and the other that is dedicated to the handling on spent fuels, their examination through a gamma spectrometry bench and their storing in a pool. The RES facility is scheduled to open in 2009. (A.C.)

  13. Experimental needs for water cooled reactors. Reactor and nuclear fuel; Les besoins experimentaux pour les reacteurs a eau legere. Reacteur et combustible

    Waeckel, N. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Beguin, S. [Electricite de France (EDF/SEPTEN), 50 - Cherbourg (France); Assedo [AREVA Framatome ANP, 92 - Paris La Defense (France)


    In order to improve the competitiveness of nuclear reactors, the trend will be to increase the fuel burn-up, the fuel enrichment, the length of the irradiation cycle and the global thermal power of the reactor. In all cases the fuel rod will be more acted upon. Experimental programs involving research reactors able to irradiate in adequate conditions instrumented fuel rods will stay necessary for the validation of new practices or new nuclear fuel materials in normal or accidental conditions. (A.C.)

  14. Nuclear, uranium, reserves, sustainability, independence; Nucleaire, Uranium, reserves, durabilite, independance

    Acket, C


    In order to evaluate the energy independence concerning the nuclear energy, the author takes the state of the art about the uranium. He details the fuel needs, the reserves on the base of the today available techniques, the reserves on the base of the future techniques and concludes positively on the energy independence for the nuclear. (A.L.B.)

  15. Dismantling of nuclear facilities; Demantelement des installations nucleaires

    Tallec, M. [CEA Marcoule, Dept. des Projets d' Assainissement-Demantelement, 30 (France); Kus, J.P. [Electricite de Fance (EDF/CIDEN), 69 - Villeurbanne (France)


    Nuclear facilities have a long estimable lifetime but necessarily limited in time. At the end of their operation period, basic nuclear installations are the object of cleansing operations and transformations that will lead to their definitive decommissioning and then to their dismantling. Because each facility is somewhere unique, cleansing and dismantling require specific techniques. The dismantlement consists in the disassembly and disposing off of big equipments, in the elimination of radioactivity in all rooms of the facility, in the demolition of buildings and eventually in the reconversion of all or part of the facility. This article describes these different steps: 1 - dismantling strategy: main de-construction guidelines, expected final state; 2 - industries and sites: cleansing and dismantling at the CEA, EDF's sites under de-construction; 3 - de-construction: main steps, definitive shutdown, preparation of dismantling, electromechanical dismantling, cleansing/decommissioning, demolition, dismantling taken into account at the design stage, management of polluted soils; 4 - waste management: dismantlement wastes, national policy of radioactive waste management, management of dismantlement wastes; 5 - mastery of risks: risk analysis, conformability of risk management with reference documents, main risks encountered at de-construction works; 6 - regulatory procedures; 7 - international overview; 8 - conclusion. (J.S.)

  16. History of the nuclear menace; Histoire de la menace nucleaire

    Le Guelte, G.


    For a half century, the nuclear conflict represents a permanent menace for the whole humanity. Through the crisis of Cuba rockets, Israel weapons, Indian explosion, Pakistan equipment, the dismantling of South Africa weapons, the discovery of the clandestine program of Iraq or the North Korean crisis, we understand nuclear threat. 178 countries decided to extend the non proliferation treaty for an indefinite period. But some questions are still to be studied: how it is possible to avoid the clandestine making of nuclear weapons, or if it is possible to convince recalcitrant countries to adhere to a treaty and only when all the questions will find an answer, the nuclear threat will be eliminated. (N.C.).

  17. Nuclear engineering vocabulary; Vocabulaire de l'ingenierie nucleaire

    Dumont, X. [FRAMATOME, 92 - Paris-La-Defense (France); Andrieux, C. [CEA/Saclay, Direction des Technologies de l' Information (DTI), 91 - Gif-sur-Yvette (France)] [and others


    The aim of this book is to bring together the French technical terms and expressions as defined by the specialized commission of terminology and neology of nuclear engineering (CSTNIN). For each term or expression is given: a possible abbreviation, its domain of use, its definition, sometimes a synonymous, eventually some notes, the related terms, the English equivalent, and its status at the date of publication of the book. This status comprises several steps: the eventual publication in the Journal Officiel de la Republique Francaise (official journal of the French republic, with its date), the related reference list and working group, the position of the enable authorities (CSTNIN, COGETERM, French Academy, minister), the last date of review, and some possible additional details. (J.S.)

  18. French Senate debate on nuclear deterrence; Dissuasion nucleaire francaise

    Vincon, S. [UMP Cher (France); Bentegeat, H. [Cema, 75 - Paris (France); Verwaerde, D. [CEA Bruyeres le Chatel, 91 (France); Quinlan, M. [IISS, Londre (United Kingdom); Tertrais, B. [Fondation pour la Recherche Strategique (FRS), 75 - Paris (France)


    The Senate committee on foreign affairs, defence and the Armed Forces met at a round table session on 14 June 2006 to discuss French nuclear deterrence. Serge Vincon presided the discussion, which covered three aspects of the subject: first, an analysis of the current and medium-term future strategic contexts and their consequences for the role of deterrence, and thus whether or not current doctrine is matched to current and future threats; second, the assets dedicated to deterrence, how well they reflect doctrine and how they fit in with other defence priorities; and finally an examination of Britain position within NATO along with future possibilities arising from closer European defence cooperation. (author)

  19. All about nuclear sciences; Tout savoir sur le nucleaire

    Le Blanc, A. [CEA Saclay DSM-SAC/USL2TI/STI/SVI, 91 - Gif-sur-Yvette (France)


    This French/English flyer, made by the French INIS team of CEA-Saclay, presents the INIS information system, the coverage and content of the database and makes the promotion of the opening of the database to free access through the web link: (J.S.)

  20. Nuclear situation in Japan; La situation du nucleaire au Japon



    This analysis takes stock on the nuclear situation in Japan. It discusses the ambitious equipment program in collaboration with the France, the destabilization of the japanese nuclear industry following the accidents and the energy policy evolutions. It presents the projects of the japanese nuclear industry: the Monju reactor restart, the Pluthermal project, the reprocessing power plant of Rokkasho Mura, the new reactors, the russian weapons dismantling, the ITER site selection and the buy out of Westinghouse by Toshiba. (A.L.B.)

  1. Nuclear physics and stable isotopes; Physique nucleaire et isotopes stables

    Goutte, D. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. d`Astrophysique, de la Physique des Particules, de la Physique Nucleaire et de l`Instrumentation Associee


    The aim of this paper is to show that fundamental research in nuclear physics requires utilization of stable isotopes; stable isotopes are essential as target material since a large quantity of nucleus have to be studied in order to appreciate all the complexity of the nuclear structure, but also as a tool, such as beams, for the same purpose. Examples are given with samarium, tin and germanium isotopes. 7 figs.

  2. Nuclear toxicology. To detect, to clean; Toxicologie nucleaire. Dectecter, depolluer

    Garcia, D. [CEA Cadarache (IBEB), Lab. de Bioenergetique Cellulaire, 13 - Saint-Paul-lez-Durance (France); Lecomte-Pradines, C. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DEI), Lab. de Radioecologie et d' Ecotoxicologie, 13 - Saint-Paul-lez-Durance (France); Quemeneur, E. [CEA Valrho (IBEB), Lab. de Radioecologie et d' Ecotoxicologie Nucleaire, 30 - Marcoule (France); Petitot, F. [CEA Centre de Pierrelatte, Institut de Radioprotection et de Surete Nucleaire, (IRSN-DRPH-SRBE), Lab. de Radiotoxicologie Experimentale, 26 (France); Souidi, M.; Bertho, J.M. [CEA Fontenay-aux-Roses, Institut de Radioprotection et de Surete Nucleaire, (IRSN-DRPH-SRBE), Lab. de Radiotoxicologie Experimentale 92 (France); Junot, Ch. [CEA Saclay (iBiTec-S/SPI), Lab. d' Etude du Metabolisme des Medicaments, 91 - Gif-sur-Yvette (France); Malard, V. [CEA Valrho (IBEB-SBTN), Lab. de Biochimie des Systemes Perturbes, 30 - Marcoule (France); Berthomieu, C.; Chapon, V. [CEA Cadarache (IBEB), Lab. des Interactions Proteine Metal, 13 - Saint-Paul-lez-Durance (France); Gilbin, R.; Misson-Pons, J. [CEA Cadarache, Institut de Radioprotection et de Surete Nucleaire (IRSN/DEI), Lab. de Radioecologie et d' Ecotoxicologie, 13 - Saint-Paul-lez-Durance (France); Vavasseur, A. [CEA Cadarache (IBEB), Lab. des Echanges Membranaires et Signalisation, 13 - Saint-Paul-lez-Durance (France); Richaud, P. [CEA Cadarache (IBEB), Lab. de Bioenergetique et Biotechnologie des Basteries et Microalgues, 13 - Saint-Paul-lez-Durance (France); Ansoborlo, E.; Taran, F.; Benech, H.; Fattal, E.; Tsapis, N.; Menetrier, F.; Deverre, J.R.; Burgada, R. [CEA Marcoule (DEN/DRCP/CETAMA), 30 (France)


    This file shows two complementary parts: one aiming to a better detection of exposure for man and environment and and other one relative to the treatments to be used when there is a contamination. The development of biological captors is a research axis that could be very useful for nuclear toxicologists that wish to dispose of perceptible measurement tools. In the same idea biological markers could be an important help to determine the toxic quantity in organism in case of internal radioactive contamination. About remedial actions, bacteria are able to reduce, to oxide, to capture pollutants and then it is not insane to use them in efficient and low cost remediation for waters or contaminated lands, especially by trace metals or radioactive compounds. Next to them, plants can offer the same service it is the case for sunflower able to treat water loaded in uranium. This file ends with a review of the different treatments known nowadays as therapies for contamination by radioisotopes used in nuclear industry. (N.C.)

  3. A comprehensive survey of nuclear reactions; Panorama des reactions nucleaires

    Cugnon, J. [Liege Univ., IFPA, AGO Dept. (Belgium)


    The various mechanisms of nuclear reactions are surveyed and classified in different regimes, based on the notions of coherent mechanisms and hard versus soft processes. The emphasis is put on the concepts at the basis of the understanding of these regimes and on the elements of nuclear structure which are involved in these different regimes, as well as the on the possibility of extracting this information. Due to lack of space and for pedagogical reasons, the discussion is limited to nucleon-induced and light-ion-induced reactions. However, a few remarks are given concerning some specific probes, such as weakly bound projectiles or neutron-rich nuclei. (author)

  4. Nuclear, phantasm and emotions; Nucleaire, fantasmes et emotions

    Michel, A


    Nuclear energy as it appears in novels and films gives an image that can influence our judgment. If the phantasms amplified by the antinuclear groups can influence the public opinion, so the communication on the nuclear subjects must be more emotional to touch the general public. Through different images carried in comic books or in fiction novels it is the anxiety that is privileged. If nuclear industry must propose a rational and well informed approach that gives value to its experience, this must not prevent to consider a more emotional perspective. (N.C.)

  5. The control of nuclear sector; Le controle du nucleaire



    The Asn is loaded with the control of the nuclear safety and the radiation protection in France: it provides this control, in the name of the state, to protect the workers, the patients, the public and the environment of the risks in relation with nuclear activities. The control is the core business of Asn. Asn so checks the nuclear basic installations (I.N.B.), since their conception until their dismantling, the pressure equipment specially conceived for these installations, the management of the radioactive waste as well as the transport of radioactive substances. Asn also checks all the industrial and research installations as well as the hospitals where are used ionizing radiations. It is a more recent profession there, because dating the reform of the control of the nuclear power of 2002, which constitutes that of the radiation protection. The first responsibility of the activities at risks falls to the one who begins them. This principle applies to all the sectors checked by Asn: an industrialist is responsible for the safety of the nuclear installations which he exploits, a doctor is responsible for the use of the ionizing radiations which he uses. (N.C.)

  6. Which nuclear energy for tomorrow?; Quelle energie nucleaire demain?

    Huffer, E.; Nifenecker, H


    Facing the constant increase of electric power consumption, the authors wonder about the energy sources possibilities. After a synthesis of the fossil fuels and the renewable energies they present the nuclear energy and more specially the new hybrid reactor project (Carlo Rubbia), or ADS (Accelerator Driven System). (A.L.B.)

  7. EPR: the nuclear impasse; EPR: l'impasse nucleaire

    Marillier, F. [Association Ecologiste Greenpeace (France)


    The questions relative to the climatic change constitute crucial challenges for the next ten years. In this context the author aims to show how the EPR project illustrates the nuclear french ''autism''. He presents and analyzes the international and environmental impacts of this obsolete technology, as a project useless and dangerous. (A.L.B.)

  8. 2035: a no nuclear France; 2035: une France sans nucleaire

    Dupin, L.; Chandes, C.; James, O.; Moragues, M.


    The authors propose a prospective scenario: the newly elected French president decides of a 20-year program to give up nuclear energy production. First, the Fessenheim and Gravelines reactors are closed. The others are to be closed by 2035. Investments are decided for offshore wind energy production, methanation projects, housing thermal insulation. Employees of the nuclear energy sector are taken into account. The authors describe the situation in 2020: energy supply problems, 5 more years of lifetime awarded to some nuclear power stations, decision to build only positive energy buildings, mandatory housing renovation, job creation, decision to develop carbon capture and storage projects. In 2025: the dismantling of nuclear reactors is going on and its cost is assessed, always more electrical vehicles, drastic cost reduction for lithium batteries. In 2035: renewable energies represent the half of the energy mix, the dismantling activity is a success for Areva. In parallel, current figures are given for energy consumption per year and per person in France and Germany, for energy French exports and imports, for electricity cost associated with the different energy sources, for the energy mix in France, for the number of jobs in the nuclear sector. In an interview, a member of the CEA comments the Italian, German and Swiss decisions to give up nuclear energy, the possibilities of its replacement by renewable energies, and the challenges associated with such a decision in France

  9. Nuclear energy and natural risks; Nucleaire et risques naturels

    Klingler, C.; Lemarchand, F.; Muller, X.


    In 10 years France has cumulated 670 significant natural events among which 136 floods, despite that, 5 nuclear power plants are situated in seismic areas and 10 near a river. This dossier reports the seismic risks in France and in Europe and shows how the risks of earthquakes, floods and heat waves are assessed and taken into account. The Fukushima accident shows that even very unlikely events have to be taken into account to assure an adequate level of safety. It is acknowledged that the state of a nuclear installation depends strongly on humane and organisational factors and the adequate use and management of sub-contracting is fundamental. (A.C.)

  10. Nuclear power in our societies; Le nucleaire dans nos societes

    Fardeau, J.C.


    Hiroshima, Chernobyl, Fukushima Daiichi are the well known sad milestones on the path toward a broad development of nuclear energy. They are so well known that they have blurred certainly for long in a very unfair way the positive image of nuclear energy in the public eye. The impact of the media appetite for disasters favours the fear and puts aside all the achievements of nuclear sciences like nuclear medicine for instance and all the assets of nuclear power like the quasi absence of greenhouse gas emission or its massive capacity to produce electricity or heat. The unique solution to enhance nuclear acceptance is the reduction of the fear through a better understanding of nuclear sciences by the public. (A.C.)

  11. Nuclear energy: a reasonable choice?; Le nucleaire: un choix raisonnable?

    Nifenecker, H.


    While nuclear energy appears today as a powerful and carbon-free energy, it generates at the same time doubts and apprehension in the general public. Are these fears justified? Is France the most advanced country in the nuclear domain? Should we fear a Chernobyl-like accident in France? Is any irradiation dangerous? What would be the consequences of a terror attack against a reactor? Will nuclear energy be powerful enough to take up the energy reserves challenge? Will the waste management and the nuclear facilities dismantlement be extremely expensive in comparison with the electricity production costs? Do we know how to manage nuclear wastes on the long-term? This book tries to supply some relevant arguments in order to let the reader answering these questions himself and making his own opinion on this topic. (J.S.)

  12. Nuclear and sustainable development; Nucleaire et developpement durable

    Audebert, P.; Balle, St.; Barandas, Ch.; Basse-Cathalinat, B.; Bellefontaine, E.; Bernard, H.; Bouhand, M.H.; Bourg, D.; Bourgoignon, F.; Bourlat, Y.; Brunet, F.; Buclet, N.; Buquet, N.; Caron, P.; Cartier, M.; Chagneau, E.; Charles, D.; Chateau, G.; Collette, P.; Collignon, A.; Comtesse, Ch.; Crammer, B.; Dasnias, J.; Decroix, G.; Defoy, B.; Delafontaine, E.; Delcroix, V.; Delerue, X.; Demet, M.; Dimmers, G.; Dodivers, S.; Dubigeon, O.; Eimer, M.; Fadin, H.; Foos, J.; Ganiage, D.; Garraud, J.; Girod, J.P.; Gourod, A.; Goussot, D.; Guignard, C.; Heloury, J.; Hondermarck, B.; Hurel, S.; Jeandron, C.; Josse, A.; Lagon, Ch.; Lalleron, Ch.; Laurent, M.; Legrand, H.; Leveau, E


    On September 15. and 16., 2004, at Rene Delcourt invitation, President of the C.L.I. of Paluel and Penly, took place the 4. colloquium of the A.N.C.L.I.. Jean Dasnias, new President of the C.L.I., welcomed the colloquium. Hundred of persons participated. The place of the nuclear power in the energy perspectives of tomorrow, its assets and its weaknesses in front of the other energies and within the framework of a sustainable development, are so many subjects which were discussed. The different tackled subjects are: the stakes in the sustainable development; energy perspectives; the reactors of the fourth generation; nuclear power and transparency; sustainable development and I.R.S.N. (N.C.)

  13. Overview of the Russian nuclear industry; Le panorama nucleaire russe



    In 2004, President Poutine decided to replace the atomic energy ministry (Minatom) by the federal atomic energy agency (Rosatom). Several projects were launched during the next two years which aimed at bringing back Russia to the fore front of the world leaders of nuclear energy use and nuclear technology export. In 2007, Rosatom agency was changed to a public holding company and a new company, named Atomenergoprom, was created which gathers all civil nuclear companies (AtomEnergoMash for the exploitation of power plants, Technabsexport (Tenex) specialized in enrichment or Atomstryexport in charge of export activities). Thus, Rosatom is at the head of all civilian and military nuclear companies, of all research centers, and of all nuclear and radiological safety facilities. In 2006, Russian nuclear power plants supplied 15.8% of the whole power consumption. Russia wishes to develop its nuclear program with the construction of new reactors in order to reach a nuclear electricity share of 25% from now to 2020. This paper presents first the 2007 institutional reform of the Russian atomic sector, and the three sectorial federal programmes: 1 - development of the nuclear energy industrial complex for the 2007-2010 era and up to 2015 (future power plants, nuclear fuel centers and reactor prototypes), 2 - nuclear safety and radioprotection for the 2008-2015 era (waste management, remedial actions, radiation protection), 3 - military program (confidential). Then, the paper presents: the international actions (export of Russian technology, cooperation agreements, non-proliferation), the situation of the existing nuclear park (reactors in operation, stopped, under construction and in project), the fuel cycle activities (production of natural uranium, enrichment, fuel fabrication, spent fuel storage, reprocessing, waste management), the nuclear R and D in Russia, and the nuclear safety authority. (J.S.)

  14. Nuclear armament and disarmament; Armement et desarmement nucleaires



    This document discusses the objectives and the specifications of the non-proliferation treaty, in the framework of the nuclear armament and disarmament. Three chapters are proposed: State of the art; the international agreements and treaties and the United Nation Organization part; debates and forecasts on the proliferation fight, the Pugwash movement and a chronology of the situation. (A.L.B.)

  15. Energy: nuclear energy; Energies: l'energie nucleaire

    Lung, M. [Societe Generale pour les Techniques Nouvelles (SGN), 78 - Saint-Quentin-en-Yvelines (France)


    Convinced that the nuclear energy will be the cleaner, safer, more economical and more respectful of the environment energy of the future, the author preconizes to study the way it can be implemented, to continue to improve its production, to understand its virtues and to better inform the public. He develops this opinion in the presentation of the principal characteristics of the nuclear energy: technology, radioactive wastes, radiation protection, the plutonium, the nuclear accidents, the proliferation risks, the economics and nuclear energy and competitiveness, development and sustainability. (A.L.B.)

  16. Non-nuclear energies; Les energies autres que le nucleaire

    Nifenecker, H. [Laboratoire de Physique Subatomique et de Cosmologie, IN2P3-CNRS/UJF/INPG, 53 av. des Martyrs, 38026 Grenoble Cedex and Sauvons le Climat (, Grenoble (France)


    The different meanings of the word 'energy', as understood by economists, are reviewed and explained. Present rates of consumption of fossil and nuclear fuels are given as well as corresponding reserves and resources. The time left before exhaustion of these reserves is calculated for different energy consumption scenarios. On finds that coal and nuclear only allow to reach the end of this century. Without specific dispositions, the predicted massive use of coal is not compatible with any admissible value of global heating. Thus, we discuss the clean coal techniques, including carbon dioxide capture and storage. One proceeds with the discussion of availability and feasibility of renewable energies, with special attention to electricity production. One distinguishes controllable renewable energies from those which are intermittent. Among the first we find hydroelectricity, biomass, and geothermal and among the second, wind and solar. At world level, hydroelectricity will, most probably, remain the main renewable contributor to electricity production. Photovoltaic is extremely promising for providing villages remote deprived from access to a centralized network. Biomass should be an important source of bio-fuels. Geothermal energy should be an interesting source of low temperature heat. Development of wind energy will be inhibited by the lack of cheap and massive electricity storage; its contribution should not exceed 10% of electricity production. Its present development is totally dependent upon massive public support. A large part of this paper follows chapters of the monograph 'L'energie de demain: technique, environnement, economie', EDP Sciences, 2005. (author)

  17. Exit this way; Par ici la sortie du nucleaire



    This paper aims to inform the public on the nuclear power phaseout necessity. The fist part presents the hazards bound the nuclear activities, from the uranium mines to the radioactive wastes disposal and the reasons of a necessary phaseout. The second part recalls the historical aspects of the nuclear power implementation in the french energy policy and the today government attitude to value its choice. The last part presents the advantages of other energies sources and scenario of nuclear power phaseout. (A.L.B.)

  18. Nuclear molecular imaging of paragangliomas; Imagerie moleculaire nucleaire des paragangliomes

    Taieb, D.; Tessonnier, L.; Mundler, O. [Service central de biophysique et de medecine nucleaire, CHU de la Timone, 13 - Marseille (France)


    Paragangliomas (PGL) are relatively rare neural crest tumors originating in the adrenal medulla (usually called pheochromocytoma), chemoreceptors (i.e., carotid and aortic bodies) or autonomic ganglia. These tumors are highly vascular, usually benign and slow-growing. PGL may occur as sporadic or familial entities, the latter mostly in association with germline mutations of the succinate dehydrogenase (SDH) B, SDHC, SDHD, SDH5, von Hippel-Lindau (VHL), ret proto-oncogene (RET), neurofibromatosis 1 (NF1) (von Recklinghausen's disease), prolyl hydroxylase domain protein 2 (PHD2) genes and TMEM127. Molecular nuclear imaging has a central role in characterization of PGL and include: somatostatin receptor imaging ({sup 111}In, {sup 68}Ga), MIBG scintigraphy ({sup 131}I, {sup 123}I), {sup 18}F-dihydroxy-phenylalanine ({sup 18}F-DOPA) positron emission tomography (PET), and {sup 18}F-deoxyglucose ({sup 18}F-FDG) PET. The choice of the tracer is not yet fully established but the work-up of familial forms often require the combination of multiple approaches. (authors)

  19. The 9. European nuclear conference; La 9. conference nucleaire europeenne

    Maurel, V.; Lewis, D.; Smirnov, V.P.; Gutierrez, J.E.; Paulin, Ph.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Horhoianu, G.; Olteanu, G.; Van der Schaaf, B.; Gavillet, D.; Lapena, J.; Ohms, C.; Roth, A.; Van Dyck, St.; Mardon, J.P.; Thomas, A.; Cipiere, M.F.; Faidy, C.; Hedin, F.; Delnondedieu, M.; Chassignole, B.; Doudet, L.; Dupond, O.; Kang, K.; Park, K.; Kim, K.; Ha, J.; Hoon-Seok, Jung; Yong-koo, Lee; Kwang-Ho, Kim; Seungwoo, Paek; Heui-Joo, Choi; Do-Hee, Ahn; Kwang-Rag, Kim; Minsoo, Lee; Sung-Paal, Yim; Hongsuk, Chung; Detroux, P.; Meessen, O.; Defloor, J.; Lars-Erik, Holm; Barescut, J.C.; Vacquier, B.; Laurier, D.; Caer, S.; Quesne, B.; Oudalova, A.; Geras' kin, St.; Dikarev, V.; Dikareva, N.; Chernonog, E.; Yang-Geun, Chung; Gab-Bock, Lee; Sun-Young, Bang; Yong-Sun, Lee; Bolognese-Milsztajn, T.; Frank, D.; Lacoste, V.; Pihet, P.; Lacronique, J.F.; Chauliac, C.; Verwaerde, D.; Pavageau, O.; Zaetta, A.; Varaine, F.; Warin, D.; Hudelot, J.P.; Bioux, Ph.; Klann, R.; Petruzzi, A.; D' auria, F.; Yung Kwon, Jin; Chul Jin, Chol; Mihalache, M.; Radu, V.; Pavelescu, M.; Schneidesch, Ch.R.; Jinzhao, Zhang; Dalleur, J.P.; Nuttin, A.; Meplan, O.; Wilson, J.; Perdu, F.; Campioni, G.; Mounier, C.; Sigrist, J.F.; Laine, Ch.; Broc, D.; Robbe, M.F.; Cariou, Y.; Seok-Kyun, Yoon; Win, Naing; Myung-Hyun, Kim; Kyung, Hee; Fridman, E.; Shwageraus, E.; Galperin, A.; Meplan, O.; Laulan, O.; Mechel-Sendis, F.; Belgaid, M.; Kadem, F.; Amokrane, A.; Hamidouche, T.; El-Khider, Si-Ahmed


    This issue gathers the abstracts of the papers presented at the ninth European nuclear conference (ENC-2005). The main part of the conference is split into 20 sessions. These sessions cover all technical aspects of nuclear power, from reactor design to waste management, without forgetting experimental and research reactors, reactor dismantling, economy, resources, safety, radioprotection and education issues. Perspectives of a nuclear renaissance are clearly visible in the world. This renaissance, mainly due to political, economical, societal and ecological factors, is fuelled by scientific and technical progress. This conference was the opportunity to present together these aspects of nuclear power and to analyze their mutual interactions.

  20. Nuclear physics in astrophysics; La physique nucleaire en astrophysique

    Arnould, M. [Universite Libre de Bruxelles (Belgium). Inst. d' Astronomie et d' Astrophysique; Samyn, M. [Universite Libre de Bruxelles (Belgium). Inst. d' Astronomie et d' Astrophysique; Fonds National de la Recherche Scientifique, Brussels (Belgium)


    In this review the following topics are covered: nuclei in astrophysics, weak and electromagnetic interactions in nuclei, primordial nucleosynthesis, big bang, stellar evolution and stellar structure and stability, non-explosive stellar nucleosynthesis, element abundances, explosive nucleosynthesis in massive stars, supernovae (WL)

  1. Questions concerning the nuclear wastes; Les dechets nucleaires en questions

    Daures, Pierre [ed.] [Electricite de France (EDF), 75 - Paris (France)


    At present, 75% of the electricity in France is of nuclear origin. Most of French people approve this mode of energy production and agree upon the continuation of the electronuclear sector exploitation. However, as any industry, the nuclear industry produces wastes which constitute a keen preoccupation of the public opinion. The nuclear program, even at its very inception, has provided the appropriate mastering of radioactive wastes by reducing their volume, by conditioning, reprocessing and storing, expressing continually its carefulness for population protection as well as for environment defence against the radiological effects. Pursuing its policy of transparency the EDF demonstrated openness and understanding towards questions raised by anyone. This brochure gives answers to the following 17 questions: -what the nuclear wastes are, which is their origin? - what is their amount? - are the nuclear waste dangerous? - how to treat the nuclear wastes? - are the radioactive waste storage sure? - is the nuclear waste transportation sure? - are these solutions sure? - why searches for long-lived radioactive wastes? - what is transmutation? - shall we bequeath to the next generations our nuclear wastes? - are there particular problems in nuclear power plant decommissioning? - what the wastes issued from decommissioning become? - are the costs of reprocessing and decommissioning taken into account in the price of the kWh? - were the nuclear wastes taken into account since the nuclear program inception? - who manages the nuclear wastes? - why France accepted the reprocessing of nuclear wastes produced in foreign countries? - is there an international policy for nuclear wastes?.

  2. Radioactive wastes conditioning; Le conditionnement des dechets nucleaires

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Moisy, P.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Advocat, T.; Andrieux, C.; Bardez, I.; Bart, F.; Boen, R.; Bourniol, P.; Brunel, G.; Chartier, D.; Cau dit Coumes, C.; Delaye, J.M.; Deschanels, X.; Faure, S.; Ferry, C.; Fillet, C.; Fournel, B.; Frizon, F.; Galle, C.; Gin, S.; Girold, C.; Grandjean, A.; Hudry, D.; Joussot-Dubien, C.; Lambertin, D.; Ledieu, A.; Lemont, F.; Moulin, N.; Peuget, S.; Pinet, O.; Piron, J.P.; Ranc, G.; Ribet, I.; Sarrade, S.; Tribet, M.; Pradel, P.; Bonnin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Forestier, A.; Bazile, F.; Parisot, J.F.; Finot, P


    Very early in its history, nuclear industry has taken care of the future of its wastes. Cementation processes for medium-level activity wastes, vitrification processes for minor actinide solutions and fission products are now proven technologies. The conditioning of wastes is just one ink in the full chain of the waste management process. However, this link is of prime importance because the future of the waste depends on the way it is conditioned. Reciprocally, the storage and disposal largely rely on the confidence given to the behaviour of waste packages with time. The leading role of France in the domain of radioactive wastes conditioning is a strong and valorisable asset at the international industrial plan, but also in terms of social acceptance by showing to the public that technical solutions exist. This monograph takes stock of the conditioning of nuclear wastes and describes the researches in progress, the stakes and the recent results obtained by the CEA (French atomic energy commission). Content: 1 - introduction: waste volumes and fluxes, management strategy, conditioning; 2 - decontamination processes and treatment processes for effluents and technological wastes; 3 - glasses, a long-lasting conditioning of wastes: glass package making, vitrification, glass formulation, structure and properties, long-term behaviour of glasses, cold crucible vitrification; 4 - present day conditioning of low- and medium-activity wastes: cements, bitumens, conditioning of metal structure wastes; 5 - search for alternate matrices and processes for the processing-conditioning of wastes: plasma-based processes for the incineration/vitrification of wastes, the Shiva process, alternate confinement materials, confinement of wastes from pyro-chemical processes; 6 - can the spent fuel be considered as a confinement matrix?: initial characteristics of spent fuels, evolution in dry storage environment, modeling of the spent fuel long-term behaviour, spent fuel containers in long-lasting and direct disposal concepts, conclusion about the direct disposal of spent fuels; 7 - general conclusion. (J.S.)

  3. Nuclear engineering vocabulary; Vocabulaire de l'ingenierie nucleaire



    The terms, expressions and definitions presented in this booklet come from the works carried out by the French specialized commission of nuclear engineering terminology and neology. This selection of terms cannot be found, in general, in classical dictionaries, or can be found but with a different meaning than the one used in nuclear engineering. All terms and expressions contained in this booklet have been already published in different issues of the Official Journal of the French Republic. This publication makes their use mandatory in replacement of foreign language equivalents inside all government services and public buildings. (J.S.)

  4. Liability for the nuclear risk; Aansprakelijkheid voor het nucleaire risico

    Faure, M. [ed.] [Rijksuniversiteit Limburg, Maastricht (Netherlands); Govaerts, P.; Malbrain, C.; Veuchelen, L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium); Spriet, B. [Katholieke Univ. Leuven (Belgium). Inst. voor Strafrecht; Heldeweg, M.; Hertogs, M.; Van Maanen, G.; De Roos, T.; Seerden, R. [Maastrichts Europees Instituut voor Transnationaal Rechtswetenschappelijk Onderzoek METRO, Rijksuniversiteit Limburg, Maastricht (Netherlands)


    Results of a cooperative research project on the juridical aspects of nuclear risk (criminal, civil and administrative aspects), according to the Belgian and Dutch laws, are presented. In this multi-disciplinary project also attention is paid to the economic impacts and positive-scientific aspects of the nuclear risk regarding radioactive waste problems and nuclear accidents. The liability for and the decision-making regarding the site selection of nuclear power plants is dealt with as well. 9 figs., 23 tabs., 198 refs.

  5. Nuclear: how to get off?; Nucleaire: comment en sortir?



    This study aims to propose actions to get off the nuclear in less than 10 years. The first part is devoted to the energy efficiency: the energy produced by the today electric network, the energy conservation in the ternary, industrial and residential sector. The second part presents an offer of electricity without the nuclear: the renewable energies, the fossil energies, the carbon tax and the nuclear park closure. (A.L.B.)

  6. Nuclear power, society and environment; Nucleaire, societe et environnement



    2 subjects are treated: the regular public opinion poll ordered by CEA, EdF, COGEMA and FRAMATOME and the denuclearization of one of the nuclear research center belonging to CEA. Every year in december the BVA polling institute leads a public opinion poll about how nuclear activities are perceived by people. This year about 1000 people have been questioned about the French nuclear power program, radioactivity, safety in nuclear facilities, nuclear wastes, information and public debates. The most meaningful result is that now fewer people think that nuclear energy will play a major role in 10 or 20 years. More people now think that radioactivity even at very low doses is dangerous. In 1946 the ZOE reactor was built on the site of the ancient stronghold of Chatillon which became the nuclear research center of Fontenay-aux-roses in april 1957. From 1958 to 1962 ZOE and a pilot unit of spent fuel reprocessing were dismantled. The test reactor Triton whose definitive shutdown took place in 1982, underwent a complete decontamination, as for Minerve reactor it was removed to Cadarache. The hot laboratories in which methods concerning the fabrication of plutonium fuels, the reprocessing and the handling of high activity wastes have been settled and tested, are due to be dismantled. 20 hot cells, 134 glove boxes and about 100 tanks of liquid effluents are involved. CEA has budgeted 910 millions of francs for the complete denuclearization of this site, it will be over in 2010. (A.C.)

  7. Nuclear disarmament: the rebound?; Desarmement nucleaire: le rebond?

    Durand, D. [Institut de Documentation et de Recherche sur la Paix - I. D. R. P., 93 - Saint-Ouen (France)


    The elimination of nuclear weapons is very often considered as a naive and unrealistic utopia, or as an ethical ambition or requirement but with no possible implementation. In this book, the author shades light on some concrete elements of the existing debate between international actors: governments, institutions, but also non-governmental organizations and opinion movements born during the last non-proliferation treaty conference. (J.S.)

  8. The nuclear fuel cycle; Le cycle du combustible nucleaire



    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  9. The japan a nuclear power?; Le Japon puissance nucleaire?

    Cumin, D.; Joubert, J.P. [Universite Jean Moulon Lyon-3, Centre Lyonnais d' Etudes de Securite Internationale et de Defense, 69 - Lyon (France)


    This work analyzes the Japan nuclear policy, in the frame of its foreign and safety policy in Pacific Asia, since the end of the cold war, especially the relations with the Usa and China. The Japan is a civil power because it has submitted the military institution to juridical restrictions and because it does not rely on the armed force to promote its national interests. The anti nuclear speech is joined with the acknowledgement of the dissuasion necessity, of the control of industrial processes and energy channels susceptible of military applications. Cultivating the ambiguity, the Japanese government can send a dissuasive message, perfectly legible, kind of communication of latent intimidation constituted by the virtual nuclear power of a state that takes part to the non proliferation treaty. (N.C.)

  10. Nuclear law 2006-2009; Droit nucleaire 2006-2009

    Bringuier, P.


    The author proposes an overview and comments of the evolution of the French legal context related to nuclear activities. More precisely, he addresses different issues and aspects: the institutional organisation, the transparency and public information, safety and radioprotection, nuclear materials (in terms of control and physical protection), transportation, trade and non proliferation issues, radioactive wastes, radiological accident, and responsibility and insurance

  11. Nuclear law; Le droit nucleaire 2006-2008

    Bringuier, P. [Montpellier-1 Univ., Droit International Public, UMR 5815, 34 (France)


    The object of this report is to present the evolution of the nuclear law during the period from 2006 to 2008, period that was characterized in France by a real rewriting from the implementation of a control authority. The prescriptive backing of nuclear activities has been deeply changed by numerous texts. In this first part are presented: (1) the institutional aspects, (2) openness and public information, (7) radioactive wastes and (9) liability and insurance. In a next publication will be treated: (3) safety and radiation protection; (4) nuclear matter, inspection, physical protection; (5) transports; (6) trade, non-proliferation; (8) radiological accidents. (N.C.)

  12. Climatic change and nuclear energy; Changement climatique et energie nucleaire

    Schneider, M


    The data presented in the different chapters lead to show that nuclear energy ids not a sustainable energy sources for the following reasons: investments in nuclear energy account financing that lacks to energy efficiency programmes. The nuclear programmes have negative effects such the need of great electric network, the need of highly qualified personnel, the freezing of innovation in the fields of supply and demand, development of small performing units. The countries resort to nuclear energy are among the biggest carbon dioxide emitters, because big size nuclear power plants lead to stimulate electric power consumption instead of inducing its rational use. Nuclear energy produces only electric power then a part of needs concerns heat (or cold) and when it is taken into account nuclear energy loses its advantages to the profit of cogeneration installations. Finally nuclear energy is a dangerous energy source, difficult to control as the accident occurring at Tokai MURA showed it in 1998. The problem of radioactive wastes is not still solved and the nuclear proliferation constitutes one of the most important threat at the international level. (N.C.)

  13. Climatic change and nuclear; Changement climatique et nucleaire

    Schneider, M


    One of the main priorities of the WWF is to increase the implementing of solutions relative to the greenhouse effect fight. In this framework the foundation published a study on the nuclear facing the climatic change problem. The following chapters are detailed: the nuclear and the negotiations on the climatic change; the nuclear close; the unrealistic hypothesis of the nuclear forecast; the nuclear facing other energy supplying options; supplying efficiency for heating, electric power, gas and renewable energies; the consumption efficiency facing the nuclear; the economical aspects; the deregulation effect; the political aspects; the nuclear AND the greenhouse effect. (A.L.B.)

  14. Edema: is there a role for nuclear medicine?; Oedemes et medecine nucleaire: pathologies et explorations diagnostiques en medecine nucleaire

    Lambert, M.; Perez, M.; Lamotte, C.; Hatron, P.Y. [CHRU de Lille, Service de Medecine Interne et d' Exploration Vasculaire, Hopital Huriez, 59 - Lille (France); Segard, M. [CHRU de Lille, Service de Dermatologie, Consultation de Cicatrisation, Hopital Huriez, 59 - Lille (France); Huglo, D. [CHRU de Lille, Service de Medecine Nucleaire, Hopital Huriez, 59 - Lille (France); Tiffreau, V. [CHRU de Lille, Service de Reeducation Fonctionnelle, Hopital Swingedauw, 59 - Lille (France); Lambert, M.; Perez, M.; Lamotte, C.; Hatron, P.Y.; Segard, M.; Huglo, D.; Tiffreau, V. [Lille Univ. Nord-de-France, 59 - Lille (France)


    The nuclear doctors are rarely requested face to a table of segment or diffuse edema. The lymphedema is certainly the only pattern of using radioisotopes. however, other rare pathologies can find benefit of the expertise of nuclear doctors. The use of functional nuclear imaging examinations can help the clinician in a diagnosis approach sometimes difficult face to pathologies for which the vital prognosis can be engaged at short term. (N.C.)

  15. Safety of the French reactors in operation; Surete des reacteurs francais en service

    Libmann, J. [CEA/Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire (IPSN), 92 (France)]|[Agence Internationale pour l' Energie Atomique, AIEA, Vienne (Austria)


    The French nuclear reactors still in operation at the end of the 1990's are all of PWR type. This paper focusses on the technical aspects of the safety of these reactors which depends on the original design and on the quality of the realization, on the ageing of the facilities, on the improvements added with time, and on the conditions of operation (incidents, periodical inspections, maintenance). The experience feedback, the reexamination of safety rules and the use of probabilistic evaluations have permitted to reach a satisfactory level of safety so far. The following aspects are presented successively: 1 - design and expected safety: design basis, defense-in-depth concept, postulated accidents and methods of accidents analysis, safety systems and principles of materials classification, complementary accidental conditions, preparation to the management of serious accidents, relative safety differences between the different units; 2 - expected safety during operation: general operation rules, periodical safety tests, preventive maintenance, training of personnel, safety culture; 3 - probabilistic evaluation of safety: interest of probabilistic safety studies, main results, evolutions; 4 - safety verifications: detection and analysis of incidents, global behaviour of the electronuclear park, presentation of some serious French incidents, importance of human factors, monitoring of the ageing of installations, the international nuclear events scale (INES); 5 - the periodical reexamination of safety: principles and practice, main results. (J.S.)

  16. Utilization of stable isotopes in power reactor; Utilisation des isotopes stables dans les reacteurs de puissance

    Desmoulins, P. [Electricite de France (EDF), 75 - Paris (France)


    The stable isotopes, besides uranium, used in EDF power nuclear reactors are mainly the boron 10 and the lithium 7. Boron is used in reactors as a neutrophagous agent for core reactivity control, and lithium, and more especially lithium 7, is extensively used as a solution in PWR moderators for primary fluid pH control. Boron and lithium ore reserves and producers are presented; industrial isotopic separation techniques are described: for the boron 10, they include dissociative distillation (Sulzer process) and separation on anionic resins, and for lithium 7, ion exchange columns (Cogema). 1 tab.

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

    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


    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.)

  18. Presentation of the Jules Horowitz reactor; Presentation du reacteur Jules Horowitz

    Dupuy, J.P. [AREVA-Technicatome, 13 - Aix en Provence (France); Perotto, G. [AREVA Framatome ANP, 92 - Paris La Defense (France); Ithurrald, G. [Electricite de France (EDF), 13 - Aix en Provence (France)


    The concepts on which the RJH reactor has been designed are: -) flexibility: the reactor must be able to give experimental support to any type of reactor (light water- heavy water- gas- or molten metal-cooled power reactor); -) adaptability: RJH's operating life will be over 50 years so the reactor will have to adjust to the experimental needs that may change a lot over a so long period; -) reactor operation: RJH is optimized to produce an availability time of 275 full charge equivalent days a year; and -) costs: the minimization of experiment costs is obtained through the integration in the same facility (or located nearby on the site) of the technical means required to perform a complete experimental program from the design of the irradiation device to its recovery after irradiation and to the examination or the possible testing of the samples. Transport costs and time delays are spared. This article gives the main features of the RJH facility from the description of the building to the specificities of the reactor core. (A.C.)

  19. BWR: Development and Validation of KERENA reactor; Les REB: Developpement et validation du reacteur KERENA

    Diercks, F.; Fuchs, M. [E.ON Kernkraft GmbH (Germany); Erve, M.; Pasler, D. [AREVA (Germany)


    KERENA is an advanced boiling water reactor, combining AREVA's and E.ON's expertise. A project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1, 250 MW, a net efficiency of about 37% and a design service life of 60 years. The development takes into account the technical and accumulated operating experience of the project partners. The plant safety concept is based on an optimized combination of a reduced number of proven active safety systems and passive safety systems, utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. All passive safety systems are validated in an experimental test program at AREVA, using 1:1 scale test facilities (INKA test facility Karlstein). The KERENA boiling water reactor is compliant with international nuclear codes and standards, and is also designed to withstand the effects of an aircraft crash involving a military aircraft or a large passenger airline. The safety level of the KERENA reactor has been able to be significantly increased compared to existing BWR plants. The advantages of the new safety concept are: -) Reduced susceptibility of safety systems to failures; -) Larger safety margins; -) Good plant behavior in the event of accidents due to the fact that conditions change at a slower rate; -) Grace periods of several days after an accident before operator intervention is required; -) Significantly reduced impact of operator error on reactor safety; -) No need for large-scale emergency response actions such as temporary evacuation or relocation of the neighboring population following a core melt accident. (A.C.)

  20. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)


    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  1. Public debate on the EPR reactor; Debat public sur le reacteur EPR



    In the framework of the new EPR European Pressurized Reactor implementation in France, the public asked the first Ministry on the protection of nuclear matters, transports and installations against the terrorism and the spiteful actions. This document provides information on the subject and shows the safety of the new reactor. (A.L.B.)

  2. The JHR reactor: a multipurpose asset for materials; Le reacteur RHJ: polyvalence au service des materiaux

    Iracane, D.; Yvon, P. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)


    The creeping obsolescence of experimental irradiation reactors around the world, in the European Union in particular, means that the Jules Horowitz Reactor (JHR), due to be commissioned by CEA in 2014, at Cadarache, is an essential addition, whether it be for the investigation of materials, or of novel fuels. Its capabilities extend beyond the requirements of investigations relating to fourth-generation systems. (authors)

  3. Phenix, a unique research reactor in Europe; Phenix, un reacteur de recherche unique en europe

    Guidez, J.; Goux, D. [CEA Valrho, Dir. de l' Energie Nucleaire (DEN), 30 - Marcoule (France); Dupraz, R. [AREVA Framatome ANP, 92 - Paris La Defense (France)


    The Phenix reactor located at Marcoule (France) went critical in 1973 for the first time. Between 1974 and 1990 it contributed to the demonstration of the feasibility of an industrial fast reactor, this demonstration was based on the following conclusions: -) an adequate handling of the sodium coolant, -) a low staff dosimetry, -) very few clad failures, -) a breeding ratio of 1.16 while 1.13 was expected, and -) a correct behaviour of the fuel assemblies. In 1990, 2 new experimental programs were launched: Capra for assessing the role of fast reactors in the management of plutonium stocks and Spin for the investigation of the transmutation of minor actinides and long life fission products. An important upgrading took place between 1993 and 2003, Phenix was then allowed to operate for 6 cycles more. The last experiments performed in Phenix are dedicated to the study of actinide transmutation (within the framework of the axis 1 of the Bataille's law) and the investigation of new nuclear fuels and materials for advanced reactors. (A.C.)

  4. The 900 MWe reactor check-up; Le check-up des reacteurs de 900 MWe



    A set of articles presents the third decennial visit of the thirty-four 900 MW French nuclear reactors. This process started in 2009 in Tricastin and Fessenheim. For this visit, reactors are stopped, one after the other, for three months, in order to re-examine safety issues. The different involved actors and their roles are presented: EDF, the IRSN, the ASN and the GPR (the expert group for nuclear reactors). Some specific issues are evoked by some experts and interveners. The case of Fessenheim, the visit of which lasted five months, is somehow more widely commented

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

    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 (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.

  6. The experimental and technological developments reactor; Le reacteur d'etudes et de developpements technologiques

    Carbonnier, J.L. [CEA Cadarache, Dept. d' Etudes des Reacteurs (DEN/DER), 13 - Saint-Paul-lez-Durance (France)


    THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

  7. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others


    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  8. Ship propulsion reactors technology; La technologie des reacteurs de propulsion navale

    Fribourg, Ch. [Technicatome, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)


    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  9. Optimisation de la gestion du combustible dans les reacteurs CANDU refroidis a l'eau legere

    Chambon, Richard

    This research has two main goals. First, we wanted to introduce optimization tools in the diffusion code DONJON, mostly for fuel management. The second objective is more practical. The optimization capabilities are applied to the fuel management problem for different CANDU reactors at refueling equilibrium state. Two kinds of approaches are considered and implemented in this study to solve optimization problems in the code DONJON. The first methods are based on gradients and on the quasi-linear mathematical programming. The method initially developed in the code OPTEX is implemented as a reference approach for the gradient based methods. However, this approach has a major drawback. Indeed, the starting point has to be a feasible point. Then, several approaches have been developed to be more general and not limited by the initial point choice. Among the different methods we developed, two were found to be very efficient: the multi-step method and the mixte method. The second kind of approach are the meta-heuristic methods. We implemented the tabu search method. Initially, it was designed to optimize combinatory variable problems. However, we successfully used it for continuous variables. The major advantage of the tabu method over the gradient methods is the capability to exit from local minima. Optimisation of the average exit burnup has been performed for CANDU-6 and ACR-700 reactors. The fresh fuel enrichment has also been optimized for ACR-700. Results match very well what the reactor physics can predict. Moreover, a comparison of the two totally different types of optimization methods validated the results we obtained.

  10. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others


    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

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

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


    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)

  12. Advanced modeling of the size poly-dispersion of boiling flows; Modelisation avancee de la polydispersion en taille des ecoulements bouillants

    Ruyer, Pierre; Seiler, Nathalie [Cadarache Batiment 700, IRSN/DPAM/SEMCA/LEIDC, BP 3, 13 115 Saint Paul lez Durance cedex (France)


    Full text of publication follows: This work has been performed within the Institut de Radioprotection et de Surete Nucleaire that leads research programs concerning safety analysis of nuclear power plants. During a LOCA (Loss Of Coolant Accident), in-vessel pressure decreases and temperature increases, leading to the onset of nucleate boiling. The present study focuses on the numerical simulation of the local topology of the boiling flow. There is experimental evidence of a local and statistical large spectra of possible bubble sizes. The relative importance of the correct description of this poly-dispersion in size is due to the dependency of (i) main hydrodynamic forces, like lift, as well as of (ii) transfer area with respect to the individual bubble size. We study the corresponding CFD model in the framework of an ensemble averaged description of the dispersed two-phase flow. The transport equations of the main statistical moment densities of the population size distribution are derived and models for the mass, momentum and heat transfers at the bubble scale as well as for bubble coalescence are achieved. This model introduced within NEPTUNE-CFD code of the NEPTUNE thermal-hydraulic platform, a joint project of CEA, EDF, IRSN and AREVA, has been tested on boiling flows obtained on the DEBORA facility of the CEA at Grenoble. These numerical simulations provide a validation and attest the impact of the proposed model. (authors) [French] Un programme de Recherche et Developpement dans le domaine de l'Accident de Perte du Refrigerent Primaire, sur un Reacteur a Eau sous Pression, a ete lance par l'Institut de Radioprotection et de Surete Nucleaire, a la Direction de Prevention des Accidents Majeurs. Au cours d'un transitoire accidentel de type APRP, le denoyage du coeur du reacteur conduit a une ebullition locale du fluide sur la paroi des crayons accompagnee d'une montee en temperature des gaines des crayons, terminee par une phase de renoyage

  13. Harmonization of welding qualification provisions in RCC-M and European standards; Rapprochement des regles relatives aux qualifications de soudage dans les normes europeennes et le RCC-M

    Lemoine, M. [Areva-NP, Tour AREVA, 92084 - Paris La Defense cedex (France); Lainez, B. [Areva-NP, usine de Chalon sur Saone 71380 Saint Marcel (France); Anastassiades, P. [EDF/Ceidre - 2 rue Ampere - 93206 SAINT-DENIS Cedex (France)


    Since a long time, numerous precautions for welding have been integrated in the nuclear codes, in particular in the RCC-M applicable to pressurized water reactors, in order to guarantee a high quality level of permanent assemblies. In parallel, European and ISO standardization works have led to a harmonisation of practices on qualification of welding processes, welders and operators. In the context of the regulatory evolutions presented during this conference, it was judged appropriate to bring closer the RCC-M practices and those of EN and ISO standards, while taking the precaution of specifying, if necessary, the complementary provisions allowing maintaining guarantees of quality consistent with the prior experience feedback. This paper presents the amendments brought to the RCC-M Code by the 2005 and 2007 addenda, in order to respond to this objective, and develops their motivations. (authors) [French] De nombreuses precautions ont ete integrees de longue date dans les codes nucleaires, et en particulier dans le RCC-M applicable aux reacteurs a eau sous pression, en matiere de soudage afin de garantir un haut niveau de qualite de ces assemblages permanents. Par ailleurs, des travaux europeens, puis ISO, ont conduit a une harmonisation des pratiques en matiere de qualification des modes operatoires de soudage et des soudeurs et operateurs. Dans le contexte des evolutions reglementaires presentees par ailleurs au cours de cette conference, il a ete juge opportun de rapprocher les pratiques du RCC-M et celles des normes EN et ISO, en prenant la precaution de specifier au besoin les dispositions complementaires permettant de maintenir des garanties de qualite coherentes avec le retour d'experience anterieur. La communication presente les amenagements apportees au code RCC-M par les modificatifs 2005 et 2007, afin de repondre a cet objectif, et en developpe les motivations. (auteurs)

  14. Use of ethanolamine for alkalization of secondary coolant. First experience at VVER reactor

    Smiesko, I. [NPP Jaslovske Bohunice (Slovakia); Bystriansky, J. [TEDIS-KOR, Dobra (Czech Republic); Szalo, A. [NPPRI Trnava (Slovakia)


    The paper summarises preparatory work and results of six-week plant trial aimed at use of ethanolamine for alkalization of secondary coolant. Operational data in pre-test and test period are given and outage inspection results are commented. Future plans are outlined. (authors)

  15. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N


    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  16. Uncertainties in the Fluence Determination in the Surveillance Samples of VVER-440

    Konheiser Joerg


    Full Text Available The reactor pressure vessel (RPV represents one of the most important safety components in a nuclear power plant. Therefore, surveillance specimen (SS programs for the RPV material exist to deliver a reliable assessment of RPV residual lifetime. This report will present neutron fluence calculations for SS. These calculations were carried out by the codes TRAMO [1] and DORT [2]. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP. The main neutron activation monitoring reactions were 54Fe(n,p54Mn and 58Ni(n,p58Co. Good agreement was found between the deterministic and stochastic calculation results and between the calculations and the ex-vessel measurements. The different influences on the monitors were studied. In order to exclude the possible healing effects of the samples due to excessive temperatures, the heat release in the surveillance specimens was determined based on the calculated gamma fluences. Under comparatively realistic conditions, the heat increased by 6 K.

  17. The analysis of normative requirements to materials of VVER components, basing on LBB concepts

    Anikovsky, V.V.; Karzov, G.P.; Timofeev, B.T. [CRISM Prometey, St. Petersburg (Russian Federation)


    The paper demonstrates an insufficiency of some requirements native Norms (when comparing them with the foreign requirements for the consideration of calculating situations): (1) leak before break (LBB); (2) short cracks; (3) preliminary loading (warm prestressing). In particular, the paper presents (1) Comparison of native and foreign normative requirements (PNAE G-7-002-86, Code ASME, BS 1515, KTA) on permissible stress levels and specifically on the estimation of crack initiation and propagation; (2) comparison of RF and USA Norms of pressure vessel material acceptance and also data of pressure vessel hydrotests; (3) comparison of Norms on the presence of defects (RF and USA) in NPP vessels, developments of defect schematization rules; foundation of a calculated defect (semi-axis correlation a/b) for pressure vessel and piping components: (4) sequence of defect estimation (growth of initial defects and critical crack sizes) proceeding from the concept LBB; (5) analysis of crack initiation and propagation conditions according to the acting Norms (including crack jumps); (6) necessity to correct estimation methods of ultimate states of brittle an ductile fracture and elastic-plastic region as applied to calculating situation: (a) LBB and (b) short cracks; (7) necessity to correct estimation methods of ultimate states with the consideration of static and cyclic loading (warm prestressing effect) of pressure vessel; estimation of the effect stability; (8) proposals on PNAE G-7-002-86 Norm corrections.

  18. TACIS 91: Application of leak-before-break concept in VVER 440-230

    Bartholome, G.; Faidy, C.; Franco, C. [and others


    The applicability of the leak-before-break (LBB) concept for primary piping in the first generation of WWER type plants in Russia is investigated. The procedures for LBB behavior used in France and Germany are applied, and the evaluation is discussed within the framework of the European Technical Assistance for the Community of Independent States (TACIS) project. Emphasis is placed on experimental validation of national and international engineering practice for evaluating and optimizing existing installations. Design criteria of WWER plants are compared to western standard design.

  19. Uncertainties in the Fluence Determination in the Surveillance Samples of VVER-440

    Konheiser, Joerg; Grahn, Alexander; Borodkin, Pavel; Borodkin, Gennady


    The reactor pressure vessel (RPV) represents one of the most important safety components in a nuclear power plant. Therefore, surveillance specimen (SS) programs for the RPV material exist to deliver a reliable assessment of RPV residual lifetime. This report will present neutron fluence calculations for SS. These calculations were carried out by the codes TRAMO [1] and DORT [2]. This study was accompanied by ex-vessel neutron dosimetry experiments at Kola NPP. The main neutron activation monitoring reactions were 54Fe(n,p)54Mn and 58Ni(n,p)58Co. Good agreement was found between the deterministic and stochastic calculation results and between the calculations and the ex-vessel measurements. The different influences on the monitors were studied. In order to exclude the possible healing effects of the samples due to excessive temperatures, the heat release in the surveillance specimens was determined based on the calculated gamma fluences. Under comparatively realistic conditions, the heat increased by 6 K.

  20. Dynamic Response of VVER 1000 Type Reactor Excited by Pressure Pulsations

    Zeman, Vladimír; Hlaváč, Zdeněk


    The paper deals with the modelling of forced vibrations of reactor components excited by pressure pulsations generated by main circulation pumps. For the vibration analysis a new generalised model of the reactor with spatial localization of the nuclear fuel assemblies and protection tubes, continuously mass distribution of beam type components and more accurate model of the linear stepper drives for actuation of control cassettes was applied. Slightly different pump revolutions are sources of...

  1. Nuclear safety and radiation protection report of the Fessenheim nuclear facilities - 2010; Rapport sur la surete nucleaire et la radioprotection des installations nucleaires de Fessenheim - 2010



    This safety report was established under the article 21 of the French law no. 2006-686 of June 13, 2006 relative to nuclear safety and information transparency. It presents, first, the facilities of the Fessenheim nuclear power plant (INB 75, Haut-Rhin, 68 (FR)). Then, the nuclear safety and radiation protection measures taken regarding the facilities are reviewed: nuclear safety definition, radiation protection of intervening parties, safety and radiation protection improvement paths, crisis management, external and internal controls, technical situation of facilities, administrative procedures in progress. The incidents and accidents which occurred in 2010, are reported as well as the radioactive and non-radioactive (chemical, thermal) effluents discharge in the environment. Finally, The radioactive materials and wastes generated by the facilities are presented and sorted by type of waste, quantities and type of conditioning. Other environmental impacts (noise) are presented with their mitigation measures. Actions in favour of transparency and public information are presented as well. The document concludes with a glossary and a list of recommendations from the Committees for health, safety and working conditions. (J.S.)

  2. A future for nuclear sites beyond their service life. Nuclear site value development; Un avenir pour les sites nucleaires en fin de cycle. Valorisation des sites nucleaires



    As the nuclear industry moves into a new development phase, many facilities built in the fifties and sixties are reaching the end of their service life. Dismantling them and rehabilitating the sites on which they stand is a major industrial challenge which will give rise to a number of new projects. AREVA has more than 20 years' experience in these highly technical fields. As more and more sites reach the end of their service life, AREVA considers nuclear site value development as a fully-fledged industrial activity. The group's competencies in this field have been grouped together to form a dedicated entity: the Nuclear Site Value Development Business Unit, created in 2008. Several billion euros are invested in site value development projects which are far-reaching and complex, and often last for several decades. Long before work actually begins, lengthy studies and preparations are required to schedule operations, select the techniques to be used and optimize costs and deadlines. The Nuclear Site Value Development BU is currently working on four major projects involving its own facilities and those of the CEA: - La Hague: dismantling of the first generation of used fuel recycling facilities. Between 1966 and 1998, almost 5,000 tons of used fuel from graphite-moderated gas-cooled reactors, 4,500 tons of light water reactor fuel, as well as fuel from fast reactors and research reactors, were treated at UP2 400, the very first industrial recycling plant on the La Hague site. - Marcoule: first-time dismantling of a recycling plant. 1,000 rooms to be cleaned up, 30,000 tons of waste to be treated, 30 years of work. - Cadarache: first-time dismantling of a Mox fuel fabrication plant. The Cadarache plant was commissioned in 1962 to fabricate fuel for fast reactors; this was followed by MOX fuel for light water reactors, an activity which continued until the plant was shut down in 2003. - Annecy and Veurey: giving a new lease of life to former industrial sites in built-up areas. AREVA is currently working on a site value development program on two industrial sites in Annecy and Veurey, near Grenoble. Dating back to 1955 and 1957, the sites were created for the manufacture and precision machining of uranium metal. The number of clean-up and dismantling projects is set to rise steadily in the years ahead and site value development will unquestionably become a new nuclear activity in its own right. AREVA is developing the corresponding competencies and expertise: new professions are springing up (value development project manager, scenario and feedback manager, operational dismantling methods manager, flushing operator, etc.) and a dedicated theoretical and practical training program is being put together. Once a nuclear site has been shut down, the long-term dismantling and value development operations make a considerable contribution to the local economy. Site rehabilitation is an environmental necessity and improves nuclear's image in the eyes of the public. As new nuclear projects get off the ground, developing the value of disused sites will free up space for future projects and make a valuable contribution to the nuclear revival.

  3. 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

    Moussiere, S


    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)

  4. Future nuclear reactors. Evolution of the safety approach; Reacteurs du futur. Evolution de l`approche de surete

    Vidard, M. [Electricite de France, 75 - Paris (France). Service Etudes et Projets Thermiques et Nucleaires


    The cumulated experience on operating nuclear power plants and the progress of safety knowledge allow to conceive reactors with high standard of safety and with optimized building and operating costs. In this article we present the proceedings followed by electricity producers, the main western nuclear plant builders and safety authorities to warrant the required level of safety. The main safety characteristics of new PWR projects are reviewed. A comparison is made between EPR, SYSTEM 80+, APWR and AP600 projects. Safety is also based on the quality of building and the excellence of operating staff. (A.C.) 16 refs.

  5. Command control of reactors and factories: general architecture; Controle-commande des reacteurs et des usines: architecture generale

    Appell, B.; Guesnier, G. [Electricite de France, 75 - Paris (France). Service Etudes et Projets Thermiques et Nucleaires; Chabert, J. [Cogema, 78 - Velizy-Villacoublay (France)


    As any industrial installation, the nuclear power plants and the fuel reprocessing plants require means to survey and to command the physical process and the associated equipment. These means are grouped under the designation of `command-control`. The command control is constituted by captors allowing to transform the physical quantities in electric signals, by automates allowing to treat these signals, by surveillance and control means at operators disposal and finally activators allowing to transform the command electric signals in mechanical actions on the process. The equipment has to answer to specifications imposed by nuclear safety. (N.C.)

  6. Representativeness elements of an hybrid reactor demonstrator; Elements de representativite d'un demonstrateur de reacteur hybride

    Kerdraon, D.; Billebaud, A.; Brissot, R.; David, S.; Giorni, A.; Heuer, D.; Loiseaux, J.M.; Meplan, O


    This document deals with the quantification of the minimum thermal power level for a demonstrator and the definition of the physical criteria which define the representative character of a demonstrator towards a power reactor. Solutions allowing to keep an acceptable flow in an industrial core, have also been studied. The document is divided in three parts: the representativeness elements, the considered solutions and the characterization of the neutrons flows at the interfaces and the dose rates at the outer surface of the vessel. (A.L.B.)

  7. Phenix reactor: a review of 35 year long operating life; Le reacteur Phenix: bilan de 35 ans de fonctionnement

    Martin, L.; Dall' Ava, D.; Rochwerger, D.; Goux, D. [CEA Marcoule 30 (France); Guidez, J.; Martin, Ph.; Seran, J.L. [CEA Saclay 91 - Gif sur Yvette (France); Sauvage, J.F.; Prele, G.; Guihard, J. [Electricite de France (EDF), 75 - Paris (France); Bernardin, B.; Vanier, M.; Zaetta, A.; Latge, Ch. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Fontaine, B.; Jolly, J.A.; Gros, J.; Pepe, D. [CEA Marcoule, Centrale Phenix, 30 (France); Pelletier, M.; Pillon, S. [CEA Cadarache, Dept. d' Etudes des Combustibles, 13 - Saint Paul lez Durance (France); Escaravage, C.; Gelineau, O.; Dupraz, R.; Dirat, J.F.; Giraud, M. [AREVA NP, 92 - Paris la Defense (France); Michaille, P. [CEA Dam, DP2I, Mar (France)


    Phenix reactor that was commissioned in 1973, had its final shutdown during the beginning of 2009. This series of articles presents the main contributions of Phenix over its 35 years of operating life in material sciences, the handling of sodium, the design of fast reactors, core physics and reactor safety. Other articles recall the feedback experience on particular components like sodium pumps, steam generators or intermediate heat exchangers and about reactor maintenance. This power plant was first an experimental reactor that, with its hot cells, has performed important irradiation programs concerning mainly fast reactor technology and transmutation as a tool for burning actinides. One article reviews the environmental impact of this reactor over its operating life in terms of waste production and dosimetry. (A.C.)

  8. RJH, a new test reactor in Europe; Le RJH - un nouveau reacteur d'essai en europe

    Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France)


    Material test reactors (MTR) are now ageing in Europe and they cannot secure the experimental needs over next decades. In this context, a new MTR, named Jules Horowitz reactor (RJH), operated as an international user-facility, is under development on the Cea's site of Cadarache (France). The design studies will end in 2007, the construction stage will follow and RJH commissioning is scheduled in 2014. Its construction costs are estimated to 500 million euros. RJH is a pool reactor of 100 MWth, its core will be inserted in a pressurized vessel with a primary circuit assuring water flow through forced convection. The core inlet-outlet temperature is about 25-40 Celsius degrees. RJH core is designed to use a high density - low enrichment UMo nuclear fuel (8 gU/cm{sup 3}, enrichment rate: 19.75%). Experimental devices located in the core will benefit from neutron fluxes ranging from 2.5 10{sup 14} n/cm{sup 2}.s to 5.10{sup 14} n/cm{sup 2}.s (E > 1 MeV). RJH is designed to manage simultaneously 10 experiments in the core and as many in the reflector. (A.C.)

  9. The Halden reactor, a facility open to the international nuclear community; Halden, un reacteur ouvert a la communaute internationale

    Vitanza, C. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (OECD/NEA), 75 - Paris (France)


    The Halden test reactor is a boiling-type reactor moderated and cooled by heavy water, it yields a thermal power of 20 MW. The reactor operates for 2 periods of about 100 days each year. The Halden reactor has been in operation for more than 45 years and is the largest OECD-NEA project, it carries out the OECD joint program and bilateral contract work. Its experimental programs are supported by about 100 organisations in 20 countries. The fuel and materials programs for the years to come focus on the following main areas: -) fuel high burn-up capabilities in normal operating conditions, -) fuel response to transients aiming at generating experimental data on the behaviour of high burn-up fuels in short duration transients and on phenomena occurring during loss of coolant accident and coolant flow oscillations, -) cladding corrosion and water chemistry issues, and -) pressure vessel embrittlement and irradiation assisted stress corrosion cracking of reactor internals. (A.C.)

  10. What availability rate for a new fast sodium reactor?; Quel taux de disponibilite pour un nouveau reacteur rapide sodium?

    Guidez, J. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)


    This article points out that 18 sodium reactors have operated in the world, prototypes to nuclear power reactors, accumulating 388 years of operation. If one discounts the prototype, only three reactors had a significant and electric power generation suitable for the analysis of availability. An analysis of availability rates for Phoenix and Superphenix is made. A comparison of availability rates of BN 600 reactor and Tricastin 1 reactor (both started in 1980) is also performed. We conclude that, since the R.E.X. (return of operating experience) of previous reactors is taken into account (mainly in material) and lack of political disturbance, can be expected for a sodium cooled fast reactor availability rates comparable to those of other existing reactors. (N.C.)

  11. Osiris, an irradiation reactor for material and nuclear fuel testing; Osiris, reacteur d'irradiation pour materiaux et combustibles

    Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Div. Nucleaire Energie, Dept. Reacteurs et Nucleaire Service, 91 - Gif-sur-Yvette (France)


    Since 1966 the Osiris reactor located at Saclay has been participating in French and international irradiation programs for research and development in the field of nuclear fuel and materials. Today the French atomic commission (Cea) pursues irradiation programs in support of existing reactors, mainly PWR, strengthening its own knowledge and the one of its clients on fuel and material behaviour under irradiation, pertaining to plant life-time issues and high burn-up. For instance important programs have been performed on pressure vessel steel aging, pellet-clad interaction, internal component aging and mox fuel qualification. With the arising of the Generation 4 research and development programs, the Osiris reactor has developed capacities to undertake material and fuel irradiation under high temperature conditions. Routine irradiations such as the doping of silicon or the production of radio-nuclides for medical or imaging purposes are made on a daily basis. The specificities of the Osiris reactor are presented in the first part of this paper while the second part focuses on the experimental devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on their associated programs.

  12. The launching of the construction of the Jules Horowitz Reactor; Lancement de la construction du reacteur de recherche Jules Horowitz



    In March 2007 the French deputy minister of industry has officially launched the construction of the new research reactor called Jules Horowitz (RJH) on the Cea site of Cadarache. RJH, that is due to operate in 2014, will be used to study the aging process of irradiated materials in any type of reactors, the behaviour of new nuclear fuels irradiated in different configurations and scenarios, and to produce radionuclides for nuclear medicine and high-quality doped silicon for the electronics industry. The investment that reaches 500 million euros is dispatched between Cea (50%), EDF (20%), Areva (10%) and foreign contributors (20%). (A.C.)

  13. Developpement de mesures non destructives, par ondes ultrasonores, d'epaisseurs de fronts de solidification dans les reacteurs metallurgiques

    Floquet, Jimmy

    Dans les cuves d'electrolyse d'aluminium, le milieu de reaction tres corrosif attaque les parois de la cuve, ce qui diminue leur duree de vie et augmente les couts de production. Le talus, qui se forme sous l'effet des pertes de chaleur qui maintiennent un equilibre thermique dans la cuve, sert de protection naturelle a la cuve. Son epaisseur doit etre controlee pour maximiser cet effet. Advenant la resorption non voulue de ce talus, les degats generes peuvent s'evaluer a plusieurs centaines de milliers de dollars par cuve. Aussi, l'objectif est de developper une mesure ultrasonore de l'epaisseur du talus, car elle serait non intrusive et non destructive. La precision attendue est de l'ordre du centimetre pour des mesures d'epaisseurs comprenant 2 materiaux, allant de 5 a 20 cm. Cette precision est le facteur cle permettant aux industriels de controler l'epaisseur du talus de maniere efficace (maximiser la protection des parois tout en maximisant l'efficacite energetique du procede), par l'ajout d'un flux thermique. Cependant, l'efficacite d'une mesure ultrasonore dans cet environnement hostile reste a demontrer. Les travaux preliminaires ont permis de selectionner un transducteur ultrasonore a contact ayant la capacite a resister aux conditions de mesure (hautes temperatures, materiaux non caracterises...). Differentes mesures a froid (traite par analyse temps-frequence) ont permis d'evaluer la vitesse de propagation des ondes dans le materiau de la cuve en graphite et de la cryolite, demontrant la possibilite d'extraire l'information pertinente d'epaisseur du talus in fine. Fort de cette phase de caracterisation des materiaux sur la reponse acoustique des materiaux, les travaux a venir ont ete realises sur un modele reduit de la cuve. Le montage experimental, un four evoluant a 1050 °C, instrumente d'une multitude de capteurs thermique, permettra une comparaison de la mesure intrusive LVDT a celle du transducteur, dans des conditions proches de la mesure industrielle. Mots-cles : Ultrasons, CND, Haute temperature, Aluminium, Cuve d'electrolyse.

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

    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


    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.)

  15. Reactors. Industrial processes in chemistry/petrochemistry. Catalytic reforming; Reacteurs. Procedes industriels chimie/petrochimie. Reformage catalytique

    Fournier, G. [AXENS-IFP, Institut Francais du Petrole (IFP), 92 - Rueil-Malmaison (France); Joly, J.F. [Institut Francais du Petrole (IFP), 92 - Rueil-Malmaison (France)


    The new regulatory constraints (pollution regulations: decrease of lead amount in gasolines, and energy savings) have led the industrialists to better valorize their petroleum cuts. More specific and adapted processes have then been perfected, for instance for the refining. After having recalled what types of reactions are to be promoted, how to obtain them with a good yield (thermodynamic and kinetic considerations), the authors describe 1)what types of catalysts are used for these reactions, what their activities are and how they are prepared 2)the industrial implementation of the catalytic reforming (main types of units, experimental conditions, equipments, main processes of reforming and pretreatment) and then give the performances of the catalytic reforming. (O.M.)

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

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


    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.)

  17. Inventory of wastes coming from EDF reactors exploitation; Inventaire des dechets issus de l'exploitation des reacteurs EDF

    Errera, J. [Electricite de France (EDF-DPN), Groupe Environnement, 93 - Saint-Denis (France)


    The present document shows the situation in the radioactive waste management of the nuclear power plants in operation. This document pays particular attention to make an inventory by waste nature the flows produced and the waste parcels delivered to the Aube plant on the period 96-99 and presents the current actions or project in order to improve the management notably for the waste without outlet. (N.C.)

  18. Caract\\'erisation de r\\'eacteurs photocatalytiques utilis\\'es pour le traitement de l'air

    Faure, Marie; Corbel, Serge; Carré, Marie-Christiane; Gérardin, Fabien; Zahraa, Orfan


    Photocatalysis is often used in air purification. Photocatalytic degradation od VOC's has been investigated in a number of devices. An optimisation of this process requires a chracterisation of the reactor : in a first step the gas flow behaviour has to be studied (plug flow, CSTR,...) by RTD analysis; then, chemical kinetics have to be determined as a function ot the relevant parameters (space time, relative humidity, concentrations, etc). All the data collected lead to a reliable design of the equipment.

  19. Mitigation of the hydrogen risk in fusion reactors; Mitigation du risque hydrogene dans les reacteurs de fusion

    Maruejouls, C.; Robin, J.C. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs; Arnould, F.; Bachellerie, E. [Technicatome DI SEPS, 13 - Aix en Provence (France); Latge, C. [CEA Cadarache, Dept. d' Etudes des Dechets DED, 13 - Saint Paul lez Durance (France); Laurent, A. [Institut National Polytechnique de Lorraine, Lab. des Sciences du Genie Chimique, 54 - Nancy (France)


    The rupture of the first wall and the intrusion of water vapor inside the torus, is one of the major accident that can occur in a thermonuclear fusion reactor. In this situation, water oxidizes the beryllium of the wall and the reaction produces hydrogen with a strong risk of explosion inside the reactor. In order to mitigate this risk, a process based on the reduction of metal oxides (MnO{sub 2}, Ag{sub 2}O) has been developed. The aim of this study is the determination of the kinetics of this reduction reaction. A mixture of both oxides has been deposited on the surface of porous balls for an experiment on fixed beds. The modeling of the phenomenon is based on the equations used in heterogenous catalysis and the experimental determination of the kinetics of the reaction is performed with the CIGNE test-facility. The velocity of the reduction reaction is deduced from the remaining amount of hydrogen in the test-gas (N{sub 2} with 1 to 2% of H{sub 2}) after it has been flowed on the oxides coated balls of the fixed bed. (J.S.)

  20. ESCRIME: Advanced control for a better safety of reactors; ESCRIME: conduite avancee pour une meilleure surete des reacteurs

    Papin, B. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires


    Activities related to the operation and exploitation of nuclear power plants can influence their safety. Probabilistic estimations show the impact of human factors on safety during operation and maintenance works. From this observation, it is tempting to minimize the human intervention during operation and on the contrary to give more weight to automation and robotization. However, such changes would not suppress the human errors but would reduce the human intervention possibilities for the treatment of unpredictable and critic situations due to the increasing complexity of automatisms. The aim of the CEA/DRN ESCRIME research program is to determine the optimum automation level of nuclear reactors which would lead to the safest operating conditions. The hardest problem to solve is the dialogue between operators and robots in the resolution of complex situations in which part of the solution is held by the operator and the other part is held by the machine. Distributed artificial intelligence techniques is one of the explored ways to solve this man-machine cooperation problem. (J.S.). 3 refs., 4 figs.

  1. Japan: the decision to build five new reactors is confirmed; Japon: la decision de construire cinq nouveaux reacteurs est confirmee



    The construction of five new units has been decided: Shimane 3 (Chugoku EPCo), Tomari 3 (Hokkaido EPCo), Fukushima I -units 7 and 8 (TEPCo), Tsuruga 3 (JAPCo). In spite of the criticality accident of Tokai-mura, the confidence in nuclear energy has not been moved. The situation is all far from easy for the Japanese nuclear. In 1998, the number of new units was varying between 16 and 20 for 2010, it is now reduced to 13 units. (N.C.)

  2. Researches on nuclear wastes: results and outlooks; Recherches sur les dechets nucleaires: resultats et perspectives



    In the framework of a sustainable development, the CEA is engaged in the design, evaluation and development of new fuels and in the research process on radioactive wastes management. This paper gathers the CEA actions in the domain giving general information on the radioactive wastes nature and management, the wastes sorting to reduce the toxicity, the wastes conditioning and the packages long-dated behavior and the wastes storage and disposal. (A.L.B.)

  3. Prospective economical study of the nuclear power file; Etude economique prospective de la filiere electrique nucleaire

    Charpin, J.M. [Commissariat General du Plan, 75 - Paris (France); Dessus, B. [Ecodev-CNRS, 92 - Meudon (France); Pellat, R. [CEA, 75 - Paris (France)


    On May 7, 1999 an economical study of the overall nuclear file, and in particular, of the back-end part of the fuel cycle and including the reprocessing, was requested by the French Prime Minister. This study includes the cost comparisons with the other means of power production and takes into consideration the environmental costs. The study is shared into five chapters dealing with: 1 - the legacy of the past: todays park of nuclear plants, economical and material status; 2 - the international evolution: the dynamics of nuclear policies worldwide (existing parks and R and D programs), the rise of environmental problems worldwide (CO{sub 2} and the climate convention, nuclear risks, attempts of including environment in the power costs), the choices made for the management of spent fuels in the main countries; 3 - the technological prospects for the power production and use: technologies for the mastery of power demand (residential, industrial and tertiary sectors, power transportation), technologies of power production (production from nuclear, fossil and renewable energies); 4 - prospective scenarios for France: two demand scenarios at the year 2050 vista (energy, electric power), power supply (supply structure with respect to scenarios, nuclear parks, power capacities), environmental aspects (CO{sub 2} emissions, plutonium and minor actinides production); 5 - the economical status of the different scenarios: data preparation, fossil fuel price scenarios, investment and operation costs of the different power production means (nuclear, fossil and renewable energies, natural gas and power distribution networks), comparison between fluxes and cumulated economic costs linked with the different scenarios (investments, exploitation, fuels, R and D, status for 2000 to 2050), time structure of expenditures with respect to the different scenarios (chronology, statuses, kWh costs, sensitivity with respect to the rate of discount, valorization of existing parks in 2050), cost overruns with respect to pollutant emissions. (J.S.)

  4. Nuclear non proliferation and disarmament; Non-proliferation nucleaire et desarmement



    In the framework of the publication of a document on the ''weapons mastership, disarmament and non proliferation: the french action'', by the ministry of Foreign Affairs and the ministry of Defense, the French Documentation organization presents a whole document. This document describes and details the following topics: the conference on the treaty of non proliferation of nuclear weapons, the France, Usa and Non Governmental Organizations position, the threats of the proliferation, the french actions towards the disarmament, the disarmament in the world, a chronology and some bibliographic resources. (A.L.B.)

  5. The nuclear cleanliness: a priority for EDF; La proprete nucleaire: une priorite pour EDF



    Since autumn 1998 EDF implemented an action plan to improve the nuclear cleanliness on its sites. The first objective is to eliminate the contamination risk, by a better professional training and installation of new detection gantries. (A.L.B.)

  6. The nuclear debate: ethics versus effectiveness; Morale et efficacite dans le debat nucleaire

    Lambert, D


    Following some political maneuvering, a new debate on the future of nuclear deterrence is about to resurface. And a first deadline has been set by the need to restore the strategic balance between the United States and Russia before the START Treaty ends on 5 December 2009, as well as by preparation for the next NPT Review Conference. Perception of the main threat has changed, but so have concepts of deterrence. Far from outmoded, deterrence forms part of a broader vision in which realism has the edge over idealism. (author)

  7. The nuclear energy in the United Kingdom; L'energie nucleaire au Royaume-Uni



    With challenges like the climatic change, the hydrocarbons prices increase and the energy supply security, the nuclear park is becoming a decisive and an urgent question in the United Kingdom. The author proposes an historical aspect of the nuclear energy in UK, the actors of the today nuclear industry and the technologies used in 2006, the radioactive wastes management, the programs of the future and the british opinion on the nuclear. (A.L.B.)

  8. Nuclear or radiology: which term to use?; Nucleaire ou radiologique: quel terme utiliser?



    This document brings information and definition to help the public in the distinction between the two terms: nuclear and radiologic. What means the words nuclear and radiologic in the physics and common languages? In which situation an accidental or malevolent event can be called nuclear or radiologic? By which technic and for which use is concerned the radiology? It concludes by recommendation for the choice of one or the other term. (A.L.B.)

  9. Nuclear energy: French government policy; Politique nucleaire et diversification energetique: orientations gouvernementales



    This report presents the French government policy concerning nuclear energy and alternative sources of energy for the next 10 years. This report overviews the situation of Super-Phenix fast reactor and presents the implications of the 30/12/1999 decree (Bataille's law) about the management of radioactive wastes.

  10. Elecnuc. Nuclear power plants in the world; Elecnuc. Les centrales nucleaires dans le monde



    This small booklet summarizes in tables all the numerical data relative to the nuclear power plants worldwide. These data come from the French CEA/DSE/SEE Elecnuc database. The following aspects are reviewed: 1997 highlights; main characteristics of the reactor types in operation, under construction or on order; map of the French nuclear power plants; worldwide status of nuclear power plants at the end of 1997; nuclear power plants in operation, under construction and on order; capacity of nuclear power plants in operation; net and gross capacity of nuclear power plants on the grid and in commercial operation; forecasts; first power generation of nuclear origin per country, achieved or expected; performance indicator of PWR units in France; worldwide trend of the power generation indicator; nuclear power plants in operation, under construction, on order, planned, cancelled, shutdown, and exported; planning of steam generators replacement; MOX fuel program for plutonium recycling. (J.S.)

  11. Regularity in time of the time dependent maxwell equations; Modeles nucleaires, modeles stellaires et hydrodynamique autogravitante

    Ducomet, B


    We review some models of self-gravitating fluids, used to described in a unified frame work collective vibration modes of heavy nuclei, and large time evolution of radiation and reacting stars. (authors)

  12. Energy policy: the nuclear in debate; Politique energetique: le nucleaire en debat

    Deleuze, O. [Secretariat d' Etat Belge a l' Energie et au Developpement Durable (Belgium); Roussely, F. [Electricite de France, EDF, 75 - Paris (France); Dessus, B. [Centre National de la Recherche Scientifique, CNRS, 75 - Paris (France); Lauvergeon, A. [Cogema, 78 - Velizy Villacoublay (France); Colombani, P. [CEA, 75 - Paris (France); Revol, H.; Jedliczka, M.; Baupin, D.; Arditi, M.; Fell, H.J


    Debates about energy policy for the France: Members of Parliament, the presidents of Electricite de France and Cogema, spokespersons for the ecologists political group, scientists, general administrator of the Cea, member of the Deutsche Bundestag set out their different points of view about the future of nuclear energy. Energy conservation and efficiency, renewable energy sources, carbon dioxide emissions and Kyoto protocol, energy supplies safety, economic points of view are as many points tackled during these debates. If everyone agrees to think that it is necessary to diversify the energy sources, no agreement comes to light about the place of nuclear energy in this diversification. (N.C.)

  13. Information report nuclear energy in Europe; Rapport d'information energie nucleaire en Europe

    Montesquiou, A. de


    This report takes stock on the nuclear energy situation in Europe. The European Union with more than 40% of the nuclear power capacity in the world, is already confronted with the nuclear energy place and stakes in the future energy policy. The report si presented in two main parts. The first part, ''the assets and the weaknesses of the nuclear energy'', deals with the economical aspects which historically based the choice of the nuclear energy and the induced impacts on the environment. The competitiveness of the nuclear energy but also the wastes management problem are discussed. The second part, ''the diplomatic and juridical framework of the nuclear energy development'', details and presents the limits of the EURATOM treaty. (A.L.B.)

  14. Which future for the nuclear counter-proliferation?; Quel avenir pour la contre-proliferation nucleaire?

    Duval, M


    After a recall of the permanent data about proliferation and of the safeguards implemented by the international community, the author demonstrates that proliferation has moved towards Asia where a real 'black market' has been created. Then he analyzes the consequences of this change on the future of nuclear deterrent. Finally, he expresses his nostalgia in front of this drift and worries about the future uselessness of the means devoted to this 'pacifying' strategy. (J.S.)

  15. Nuclear weapon race does not stop; Le nucleaire ne desarme pas

    Vahe Ter, Minassian


    60 years after Hiroshima, the race for nuclear weaponry keeps on. The comprehensive test ban treaty (CTBT), signed in 1996 by the 5 official nuclear-weapon-owning states (Usa, Russia, China, U.K. and France), has not yet been implemented because its implementation requires the ratification of 44 states that harbour on their territories industrial or research nuclear reactors. Till now only 33 such states have ratified CTBT. CTBT aims at prohibiting any nuclear test whatever the amount of energy released in it. Countries like Usa, North-Korea, Russia, soon Iran... are suspected to develop new types of nuclear warfare. For 2005 the American Congress have decided to freeze the funding of programmes dedicated to the development of 'mini-nukes' like the bunker-burster. The international network of monitoring stations will soon cover all the world and will be able to detect and locate, in an almost automated way, any test involving an energy greater than 1 kiloton. 321 stations have been settled and their efficient detection systems are based on seismic or infra-sound or radioactivity or hydro-acoustic analysis. (A.C.)

  16. Elecnuc. Nuclear power plants in the world; Elecnuc. Les centrales nucleaires dans le monde



    This small booklet summarizes in tables all data relative to the nuclear power plants worldwide. These data come from the IAEA's PRIS and AREVA-CEA's GAIA databases. The following aspects are reviewed: 2007 highlights; Main characteristics of reactor types; Map of the French nuclear power plants on 2007/01/01; Worldwide status of nuclear power plants (12/31/2007); Units distributed by countries; Nuclear power plants connected to the Grid- by reactor type groups; Nuclear power plants under construction on 2007; Evolution of nuclear power plants capacities connected to the grid; First electric generations supplied by a nuclear unit in each country; Electrical generation from nuclear power plants by country at the end 2007; Performance indicator of French PWR units; Evolution of the generation indicators worldwide by type; Nuclear operator ranking according to their installed capacity; Units connected to the grid by countries at 12/31/2007; Status of licence renewal applications in USA; Nuclear power plants under construction at 12/31/2007; Shutdown reactors; Exported nuclear capacity in net MWe; Exported and national nuclear capacity connected to the grid; Exported nuclear power plants under construction; Exported and national nuclear capacity under construction; Nuclear power plants ordered at 12/31/2007; Long term shutdown units at 12/31/2007; COL (combined licences) applications in the USA; Recycling of Plutonium in reactors and experiences; Mox licence plants projects; Appendix - historical development; Meaning of the used acronyms; Glossary.

  17. Nuclear bad year in perspective in 2007; Nucleaire mauvaise annee en perspective en 2007?

    Melquiot, P


    The author presents a mixed evaluation of the nuclear after the announcement of the shutdown of 7 reactors in the world. The report factor 4 precises that the nuclear energy represents only 2% of the final energy in the world and shows the poor part of the nuclear in the fight against the greenhouse effect. (A.L.B.)

  18. Nuclear energy and its synergies with renewable energies; Le nucleaire dans ses synergies avec les renouvelables

    Carre, F. [CEA Saclay, DEN, 91 - Gif-sur-Yvette (France); Mermilliod, N. [CEA Grenoble, Dir. de la Recherche Technologique, 38 (France); Devezeaux De Lavergne, J.G. [CEA Saclay, Dir. de l' Institut de tecchnico-economie des systemes energetiques I-tese, 91 - Gif-sur-Yvette (France); Durand, S. [CEA Grenoble, European Institute of Technology -KIC InnoEnergy, 38 (France)


    France has the ambition to become a world leader in both nuclear industry and in renewable energies. 3 types of synergies between nuclear power and renewable energies are highlighted. First, nuclear power can be used as a low-carbon energy to produce the equipment required to renewable energy production for instance photovoltaic cells. Secondly, to benefit from the complementary features of both energies: continuous/intermittency of the production, centralized/local production. The future development of smart grids will help to do that. Thirdly, to use nuclear energy to produce massively hydrogen from water and synthetic fuels from biomass. (A.C.)

  19. Nuclear energy: what scenarios for the future?; Nucleaire: quels scenarios pour le futur?

    Chatelier, Michel; Criqui, Patrick; Heuer, Daniel; Huet, Sylvestre


    Because of its energetic, environmental, economical, social, safety, political, and even ideological aspects, the nuclear energy is a major society stake. It is a technical and complex topic as well which merits a democratic, well argued and transparent debate. This book supplies to the reader all the necessary information for a thorough analysis of this much disputed energy source without any bias: why France is one of the most nuclearized country in the world? Can we get out of nuclear energy? How and at what price? Wastes and safety: what can we expect (or not) from the next generations of reactors? Should we have a referendum? In this book, two researchers of the nuclear domain, an economist and a journalist invite us to consider the problem from all angles. (J.S.)

  20. Joliot-Curie School of Nuclear Physics, 1997; Ecole Joliot-Curie de Physique Nucleaire, 1997

    Abgrall, Y. [L`Institut National de Physique Nucleaire et de Physique des Particules du CNRS (India2P3), 75 - Paris (France); Collaboration: La Direction des Sciences de la Matiere du CEA (FR); Le Fonds National de la Recherche Scientifique de Belgique (BE)


    This document contains the lectures of the Joliot-Curie International School of Nuclear Physics held at Maubuisson, France on 8-13 September 1997. The following lectures of nuclear interest were given: The N-body problem (relativistic and non-relativistic approaches); The shell model (towards a unified of the nuclear structure); Pairing correlations in extreme conditions; Collective excitations in nuclei; Exotic nuclei (production, properties and specificities); Exotic nuclei (halos); Super and hyper deformation (from discrete to continuum, from EUROGAM to EUROBALL); and The spectroscopy of fission fragments. Important new facts are reported and discussed theoretically, concerning the nuclei in high excitation and high states and of the nuclei far off stability. Important technical achievements are reported among which the production of radioactive beams, sophisticated multi-detectors as well as significant advances in the nuclear theoretical methods. The double goal of training of young researchers and of permanent formation and information of the older ones seems to have been reached

  1. The nuclear in Italy - state of the art; Le nucleaire en Italie - etat des lieux

    Schifano, F.; Ziller, T


    This report aims to evaluate the italian situation in matter of the nuclear, following the referendum of 1987 which decided to stop the nuclear power plants in the country. The first part is devoted to the historical aspects of the nuclear sector in Italy. The second chapter presents the institutional and legislative framework. The third chapter discusses the today situation and the italian actors of the nuclear, from the radioactive wastes management and the dismantling of nuclear installations to the engineering service realized in other countries. It discusses also the research and development programs. The last chapter proposes perspectives of the debate around a possible restart of the nuclear activity in Italy.

  2. Nuclear energy for the 21. century; Energie nucleaire pour le 21. siecle



    This document gathers 5 introductory papers to this conference about nuclear energy for the 21. century: the French energy policy during the last 30 years (situation of France with respect to the energy supply and demand, main trends of the French energy policy, future of the French nuclear policy); presentation of IAEA (technology transfer, nuclear safety, non-proliferation policy, structure and financial resources, council of governors, general conference, secretariat); nuclear power and sustainable development; promoting safety at nuclear facilities (promoting safety, basics of safety, safety at the design stage, risk management, regulatory control and efficiency of the regulation organization, role of IAEA); nuclear energy today (contribution to sustainable development, safety, best solution for the management of radioactive wastes, future of nuclear energy). (J.S.)

  3. The truth about nuclear energy - the forbidden choice; La verite sur le nucleaire - Le choix interdit

    Lepage, Corinne


    In France, nuclear energy is a taboo topic. From the safety of nuclear power plants, to the real cost of this energy source and its constraints on our democracy, the lack of transparency is the key word. Since the Fukushima catastrophe, everything has changed: what would happen in France if such an accident would occur? Are we really prepare to this type of event? What is the weight of the nuclear lobby? In this book, the author, who is a former French Minister of environment and today a member of the European parliament, answers these legitimate questions coming from the public opinion

  4. Links between nuclear medicine and radiopharmacy; Structuration des liens entre medecine nucleaire et radiopharmacie

    Pelegrin, M. [Inserm, U896, CRLC Val-d' Aurelle-Paul-Lamarque, institut de recherche en cancerologie de Montpellier (IRCM), universite Montpellier 1, 34 - Montpellier (France); Francois-Joubert, A. [Service de medecine nucleaire, centre hospitalier de Chambery, 73 - Chambery (France); Chassel, M.L. [Radiopharmacie, service de pharmacie, centre hospitalier de Chambery, 73 - Chambrry (France); Desruet, M.D. [Service de radiopharmacie et service pharmaceutique, clinique universitaire de medecine nucleaire, CHU de Grenoble, 38 - Grenoble (France); Bolot, C. [Service de radiopharmacie, service pharmaceutique, centre de medecine nucleaire, groupement hospitalier Est, 69 - Bron (France); Lao, S. [Service de radiopharmacie, medecine nucleaire, hopital de l' Archet, 06 - Nice (France)


    Radiopharmaceuticals are nowadays under the responsibility of the radio-pharmacist because of their medicinal product status. Radiopharmacy belongs to the hospital pharmacy department, nevertheless, interactions with nuclear medicine department are important: rooms are included or located near nuclear medicine departments in order to respect radiation protection rules, more over staff, a part of the material and some activities are shared between the two departments. Consequently, it seems essential to formalize links between the radiopharmacy and the nuclear medicine department, setting the goals to avoid conflicts and to ensure patients' security. Modalities chosen for this formalization will depend on the establishment's organization. (authors)

  5. French people and nuclear wastes; Les francais et les dechets nucleaires

    D' Iribarne, Ph. [Centre National de la Recherche Scientifique (CNRS), 75 - Paris (France)


    On March 21, 2005, the French minister of industry gave to the author of this document, the mission to shade a sociological light on the radioactive wastes perception by French people. The objective of this study was to supply an additional information before the laying down in 2006 of the decisions about the management of high-level and long-lived radioactive wastes. This inquiry, carried out between April 2004 and March 2005, stresses on the knowledge and doubts of the questioned people, on the vision they have of radioactive wastes and of their hazards, and on their opinion about the actors in concern (experts, nuclear companies, government, anti-nuclear groups, public). The last two parts of the report consider the different ways of waste management under study today, and the differences between the opinion of people living close to the Bure site and the opinion of people living in other regions. (J.S.)

  6. Particles identification using nuclear emulsion in OPERA; Identification des particules par les emulsions nucleaires dans OPERA

    Manai, K


    The Opera experiment will try to confirm the {nu}{sub {mu}} {yields} {nu}{sub {tau}} oscillations by the appearance of the {nu}{sub {tau}} in a pure {nu}{sub {mu}} beam. Indeed, a neutrino beam almost pure is produced at CERN (CNGS Beam) and sent to the Opera detector. The detector is composed of two muons spectrometers and a target formed by walls of bricks. Each brick is an alternation of lead plates and emulsions. This modular structure allows to reconstruct the kink topology of the {tau} lepton decay with a high spatial resolution. The great challenge of the Opera experiment is to detect the {nu}{sub {tau}} interactions with the less uncertainty. To reduce this uncertainty it is essential to identify with the greatest efficiency any background event not including a tau particle. My work permits to reduce background. My principal contribution concerns the selection development, the reconstruction and the muons identification at low energy. This work is based on the setting of variables related to the deposit energy and the multiple scattering. Previously, only deposit energy was used in the analyses of pion/muon separation. This study allows doubling the muon identification efficiency at low energy. This leads to increase the background events rejection in Opera and to decrease the contamination by 30%. I also studied the nuclear emulsions capacity to identify charged particles through the analysis of a test beam carried out by the Nagoya group. This test contains protons and pions with different energies. My work proves that the European scan system gives comparable results with those obtained by the Japanese scan system. (author)

  7. Situation of nuclear industry in Japan; La situation du nucleaire au Japon



    This document presents the situation of nuclear industry in Japan: cooperation with France in the domain of the fuel cycle (in particular the back-end) and of for the industrial R and D about fast reactors and nuclear safety; present day situation characterized by a series of incidents in the domain of nuclear safety and by an administrative reorganization of the research and safety organizations; power of local representatives, results of April 2003 elections, liberalization of the electric power sector, impact of the TEPCO affair (falsification of safety reports) on the nuclear credibility, re-start up of the Monju reactor delayed by judicial procedures, stopping of the program of MOX fuel loading in Tepco's reactors, discovery of weld defects in the newly built Rokkasho-mura reprocessing plant, an ambitious program of reactors construction, the opportunity of Russian weapons dismantling for the re-launching of sodium-cooled fast reactors; the competition between France and Japan for the setting up of ITER reactor and its impact of the French/Japanese partnership. (J.S.)

  8. Back end of the nuclear fuel cycle; Aval du cycle du combustible nucleaire

    Dognon, J.P.; Rabbe, C.; Beudaert, Ph.; Lamare, V.; Wipff, G.; Moisy, Ph.; Charrin, N.; Blanc, P.; Den Auwer, Ch.; Revel, R.; Charbonnel, M.C.; Presson, M.T.; Cau Dit Coumes, C.; Chopin-Dumas, J.; Devisme, F.; Rat, B.; Hill, C.; Guillaneux, D.; Madic, C.; Carrera, A.; Dozol, J.F.; Rouquette, H.; Allain, F.; Virelizier, H.; Moulin, Ch.; Lemort, F.; Orlhac, X.; Fillet, C.; Carpena, J.; Advocat, T.; Leturcq, G.; Lacombe, J.; Bonnetier, A.; Ribet, I.; Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H.; Meyer, D.; Conocar, O.; Pettier, J.L.; Raphael, T.; Bouniol, P.; Sercombe, J.; Badouix, P.; Adenot, F.; Le Bescop, P.; Mazoin, C.; Motellier, S.; Charles, Y.; Richet, C.; Ayache, R.; Pitsch, H.; Ly, J.; Beaucaire, C.; Devol-Brown, I.; Libert, M.F.; Besnainou, B


    In this chapter of the DCC 1999 scientific report, the following theoretical studies are detailed: electronic structure of lanthanides or actinides complexes, forecasting of the stoichiometry of europium nitrate complexes, actinides aqueous solutions analytical and thermodynamical chemistry, actinides complexes structural determination. It also provides experimental studies: actinides and lanthanides separation, radioactive wastes processing and conditioning, plasma torch vitrification process, simulation of the wastes packages characterization, wastes storage with concrete behaviour and biodegradation. (A.L.B.)

  9. Nuclear material management: challenges and prospects; La gestion des matieres nucleaires: defis et perspectives

    Rieu, J. [Autorite de Surete Nucleaire, 75 - Paris (France); Besnainou, J. [Areva, Dir. du Secteur Aval, 75 - Paris (France); Leboucher, I.; Chiguer, M. [Areva, Dir. International et Marketing, 75 - Paris (France); Capus, G.; Greneche, D. [Areva, 75 - Paris (France); Durret, L.F. [Areva, Dir. de la Business Unit Enrichissement, 75 - Paris (France); Carbonnier, J.L.; Delpech, M. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France); Loaec, Ch. [CEA Saclay (DEN/DANS/ITESE), 91 - Gif-sur-Yvette (France); Devezeaux de Lavergne, J.G. [AREVA NC, 78 - Velizy Villacoublay (France); Granger, S. [Electricite de France (EDF), 75 - Paris (France); Devid, S.; Bidaud, A. [Centre National de la Recherche Scientifique (CNRS), 75 - Paris (France); Jalouneix, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DEND), 92 - Clamart (France); Toubon, H. [Areva/Canberra, Dir. du Developpement, 75 - Paris (France); Pochon, E. [CEA Bruyeres-le-Chatel (DSNP), 91 (France); Bariteau, J.P.; Bernard, P.; Krellmann, J. [Areva NC, Business Unit Recyclage, 78 - Velizy Villacoublay (France); Sicard, B. [CEA Cadarache, 13 - Saint Paul lez Durance (France)


    The articles in this dossier were derived from the papers of the yearly S.F.E.N. convention, which took place in Paris, 12-13 March 2008. They deal with the new challenges and prospects in the field of nuclear material management, throughout the nuclear whole fuel cycle, namely: the institutional frame of nuclear materials management, the recycling, the uranium market, the enrichment market, the different scenarios for the management of civil nuclear materials, the technical possibilities of spent fuels utilization, the option of thorium, the convention on the physical protection of nuclear materials and installations, the characterisation of nuclear materials by nondestructive nuclear measurements, the proliferation from civil installations, the use of plutonium ( from military origin) and the international agreements. (N.C.)

  10. Report transparency and nuclear safety 2007 - CISBIO; Rapport transparence et securite nucleaire 2007 - CISBIO



    This report presents the activities of CISBIO, nuclear base installation, for the year 2007. CISBIO realizes at Saclay most of the radiopharmaceuticals and drugs distributed in France for the nuclear medicine. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. (A.L.B.)

  11. Report transparency and nuclear safety 2007 CEA Marcoule; Rapport transparence et securite nucleaire 2007 CEA Marcoule



    This report presents the activities of the CEA Center of Marcoule for the year 2007. Since its creation in 1955 the center realizes industrial and scientific activities relative to the civil and military applications of the radioactivity. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the following two base activities are detailed: Atalante and Phenix. (A.L.B.)

  12. Report transparency and nuclear safety 2007 CEA Grenoble; Rapport transparence et securite nucleaire 2007 CEA Grenoble



    This report presents the activities of the CEA Center of Grenoble for the year 2007. Since 2002 the Passage project aims to realize the decontamination and the dismantling of old nuclear installations of the CEA Grenoble. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the year 2007 saw two main steps of the Passage project: the decommissioning of the Siloette reactor, a public consultation about the Lama laboratory dismantling. (A.L.B.)

  13. Report transparency and nuclear safety 2007 CEA Cadarache; Rapport transparence et securite nucleaire 2007 CEA Cadarache



    This report presents the activities of the CEA Center of Cadarache for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the report discusses the beginning of the RJH reactor construction, the fourth generation reactors research programs, the implementing of la Rotonde the new radioactive wastes management installation, the renovation of the LECA. (A.L.B.)

  14. Report transparency and nuclear safety 2007 CEA Saclay; Rapport transparence et securite nucleaire 2007 CEA Saclay



    This report presents the activities of the CEA Center of Saclay for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially two public consultation on release authorizations and the Neurospin installations, the dismantling of the 49 nuclear installation, the shutdown of the learning reactor ULYSSE are detailed. (A.L.B.)

  15. Is there any future for nuclear weapons?; Les armes nucleaires ont-elles un avenir?

    Heisbourg, F.


    Nuclear weapons occupy a paradoxal place both in the collective imagination and in the historical reality: on the one hand everybody dreads the apocalypse horror, and on the other hand, dissuasion appears as an unchanging and quite comfortable situation. However, the world has become multipolar in this domain as well. The geopolitical map is reconstructing. Doctrinal revisions, initiatives against nuclear weapons proliferation, and nuclear disarmament measures are now on the agenda. The best foreign and French experts examine for the first time the consequences of these evolutions. They analyse in particular the split up risks and the potential consequences of a nuclear conflict in regions where atomic arms have become a key-component of the strategic landscape: Middle-Est, Far-East, Southern Asia. The choices France and its allies will have to face are examined as well. (J.S.)

  16. To control the nuclear safety and the radiation protection; Controler la surete nucleaire et la radioprotection

    Lacoste, A.C. [Direction Generale de la Surete Nucleaire et de la Radioprotection, 75 - Paris (France); Bordarier, Ph. [Autorite de Surete Nucleaire (ASN), 75 - Paris (France); Saint-Raymond, Ph. [Conseil General des Mines, 75 - Paris (France); Repussard, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Gouze, J.R. [Apave Groupe, 75 - Paris (France); Degos, L. [Haute Autorite de Sante, 93 - Saint-Denis La Plaine (France); Massart, S.; Wiroth, P.; Thezee, Ch.; Petit, G. [Electricite de France (EDF), 75 - Paris (France); Cahen, B.; Hubert, I.; Wiroth, P.; Thezee, Ch.; Petit, G. [Ministere de l' Ecologie et du Developpement Durable, Direction de la Prevention des Pollutions et des Risques (DPPR), 75 - Paris (France); Kaufer, B. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (AEN/OCDE), 75 - Paris (France); Taniguchi, T. [Agence Internationale de l' Energie Atomique (AIEA), Vienna (Austria); Revol, H. [Office Parlementaire d' Evaluation des Choix Scientifiques et Technologiques (OPECST), 75 - Paris (France)


    Publishing this dossier, the aim is to present the principles and the variety of issues linked to nuclear safety and radiation protection supervision, and the main strategic choices made to use efficiently and effectively A.S.N. supervision means. A.S.N. is responsible for nuclear safety and radiation protection supervision. A.S.N. has to be itself evaluated and supervised by external bodies. The Parliament Office for Evaluation of Scientific and Technological Options (O.P.E.C.S.T.) supervises it; the foreign peers watch and A.S.N. has to be the object of an international audit conducted by its peers under the leadership of I.A.E.A. by the beginning of 2007. (N.C.)

  17. The nuclear future in France; Le futur du nucleaire en France

    Lemperiere, F.


    After having outlined some important facts in the world concerning energy production (phasing out nuclear by many countries, share of nuclear energy, development of wind and solar energy, possibility of meeting electricity demand with 80% of renewable energies) and in France (nuclear plants, debate about nuclear, slow development of renewable energies), the author briefly assesses the cost of two scenarios: a reduction of nuclear energy from 450 to 300 TWh/year by 2025, and a reduction of nuclear energy from 300 to 0 TWh/year between 2025 and 2040

  18. The technological prospective of non nuclear channels; La prospective technologique des filieres non nucleaires

    Claverie, M.; Clement, D.; Girard, C


    This prospective study concerns the electric power demand in 2050. It examines the three non nuclear sectors of production: the natural gas combined cycle power plants, the wind turbines among the renewable energies and the cogeneration electric power - heat in the ternary and building sector. The necessity of the network adaptation to the european competition and the decentralized production of electric power will suppose new investments of transport and storage. (A.L.B.)

  19. Thermodynamic cycles of nuclear power plants; Les cycles thermodynamiques des centrales nucleaires

    Normand, T.; Andreani, J.; Tejedor, V.


    Above all a nuclear power plant is a thermal system whose efficiency relies on the thermodynamic cycle that turns the heat produced by the fission of uranium nuclei into electricity. The thermodynamic yield is an essential parameter to dimension a power plant. This book is dedicated to the presentation of the thermodynamic cycles that have been chosen in the different types of French PWR: CP0, P4, N4 and EPR. These cycles are classical steam cycles that have been used and optimized for 40 years, but they are physically limited and do not respond to the expectations of future generations of reactors. The last part deals with the thermodynamic cycles that might be involved in the fourth generation of nuclear reactors: cycles with super-critical steam, direct cycles for high temperature gas, indirect gas cycles, and cycles with super-critical CO{sub 2}. The first part of the book gives an account of the situation of nuclear power in France. (A.C.)

  20. The diverse applications of the nuclear power; Les diverses applications du nucleaire



    The three great categories of application in industry and environment of ionizing radiations are the use of ionizing radiation to transport energy in matter it is industrial irradiation, the use of radioactive sources of low activity to analyze and measure, it is the nucleonic instrumentation, the use of radioactive tracers to follow and study the matter transfer. Are explained the treatments to improve the plastic materials and the ionisation of food. (N.C.)