Mesures de procédure spéciales et respect des droits de l'homme Rapport général

NARCIS (Netherlands)

Vervaele, J.A.E.

2009-01-01

Le but du rapport général est de mener une analyse comparative des rapports nationaux en vue de présenter les processus de transformation des systèmes de justice pénale internes, en particulier du procès pénal, étant donné que des mesures procédurales spéciales sont introduites pour appréhender le

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-09-15

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

Mesure magnétique du sextupôle rotatif no.1-2083

CERN Document Server

Chritin, R; Patron, G; CERN. Geneva. SPS and LEP Division

1998-01-01

Lorsque cet aimant nous a été livré afin d'être soumis à un certain nombre de mesures, ses caractéristiques connues se limitaient à celles situées dans l'annexe 1. Il correspond à des sextupôles, INT 2170254 AB L orientables récupérés de l'ancienn

Health physics during work on the G. 2 and G. 3 reactor exchanges; La radioprotection des travaux sur les echangeurs des piles G. 2 et G. 3

Energy Technology Data Exchange (ETDEWEB)

Rodier, J; Chassany, J; Guillermin, P [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1965-07-01

During this work and its preparation, which took place first at G. 2 and then at G. 3 over a period of 11 months, 15000 measurement results were obtained. Their analysis, together with a consideration of the organisation on the site and of the conclusions drawn from the experiment, shows the various factors which determine the importance of the radio-active dangers. (authors) [French] Au cours de ces travaux et de leur preparation, qui ont eu lieu successivement a G. 3 puis a G. 2, pendant 11 mois, 15 000 resultats de mesures ont ete obtenus. Leur etude, mise en parallele avec l'organisation du chantier et les enseignements tires de l'experience, met en evidence les divers facteurs conditionnant les niveaux de risques radioactifs. (auteurs)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Reactor G1: high power experiments

International Nuclear Information System (INIS)

Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

1957-01-01

The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

Induction flowmeters for the measurement of water flow rates; Debitmetre a induction pour mesure des debits d'eau

Energy Technology Data Exchange (ETDEWEB)

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

1954-07-01

This article concerns a induction flow indicator used at the reactor of Chatillon for the measure of the water debits. It has two sensitivities respectively 2,5 m{sup 3}/h and 10 m{sup 3}/h to the maxima of deviation. The precision of the measures is 1 percent of the maximum of the scale. The equipment is constituted an electronic amplifier followed by a synchronous demodulator functioning to the frequency of the sector. (author) [French] L'article concerne un debitmetre a induction utilise a la Pile de Chatillon pour la mesure des debits d'eau. Il y a deux sensibilites respectivement 2,5 m{sup 3}/h et 10 m{sup 3}/h aux maxima de deviation. La precision des mesures est de 1 pour cent du maximum de l'echelle. L'appareillage est constitue d'un amplificateur electronique suivi d'un demodulateur synchrone fonctionnant a la frequence du secteur. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

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

Neutron measurements in the core and blankets of the reactor Rapsodie; Mesures neutroniques dans le coeur et les couvertures de Rapsodie

Energy Technology Data Exchange (ETDEWEB)

Gourdon, J; Edeline, J C [Commissariat a l' Energie Atomique, 13 - Cadarache (France). Centre d' Etudes Nucleaires

1968-07-01

Beside a brief general discussion, the report contains all the core and blanket neutronic measurements. It covers successively the methods, the measurements themselves and the results. The later concern: spectral indexes, axial and radial fission rates, activation foil measurements and neutronic power determination. (authors) [French] Apres une breve description generale de RAPSODIE, le rapport presente l'ensemble des mesures neutroniques faites dans le coeur et les couvertures. Il traite dans l'ordre des methodes, des mesures et enfin des resultats qui concernent: les indices de spectres, les taux de fission axiaux et radiaux, les mesures par detecteurs par activation, la determination de la puissance, neutronique. (auteurs)

Automatic measuring device for atomic oxygen concentrations (1962); Dispositif de mesure automatique de concentrations d'oxygene atomique (1962)

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Deiss, M; Mercier, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

Within the framework of the activities of the Autonomous Reactor Electronics Section we have developed a device, which renders automatic one type of measurement carried out in the Physical Chemistry Department at the Saclay Research Centre. We define here: - the physico-chemical principle of the apparatus which is adapted to the measurement of atomic oxygen concentrations; - the physical principle of the automatic measurement; - the properties, performance, constitution, use and maintenance of the automatic measurement device. It is concluded that the principle of the automatic device, whose tests have confirmed the estimation of the theoretical performance, could usefully be adapted to other types of measurement. (authors) [French] Dans le cadre des activites de la Section Autonome d'Electronique des Reacteurs, il a ete realise et mis au point un dispositif permettant de rendre automatique un type de mesures effectuees au Departement de Physico-Chimie du C.E.N. SACLAY. On definit ici: - le principe physico-chimique de l'appareillage, adapte a la mesure de concentrations de l'oxygene atomique; - le principe physique de la mesure automatique; - les qualites, performances, constitution, utilisation, et maintenance du dispositif de mesure automatique. Il est porte en conclusion, que le principe du dispositif automatique realise, dont les essais ont sensiblement confirme l'evaluation des performances theoriques, pourrait etre utilement adapte a d'autres types de mesures courantes. (auteurs)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1953-12-15

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

Description of a Reactivity Measuring Apparatus; Description d'un Type d'Appareil de Mesure de la Reactivite

Energy Technology Data Exchange (ETDEWEB)

Deiss, M.; Uberschlag, J. [Centre d' Etudes Nucleaires de Saclay (France)

1966-06-15

apparatuses have different dividers in the output unit, in the one case Potentiometric, and in the other electronic. The first apparatus is suitable in principle for most normal measurements. The second, by reason of its shorter response time, is suited both to measurements of higher reactivity values with the reactor level rising, and also with the reactor level falling, even over limited power ranges. (author) [French] La mesure de la periode ou du temps de doublement fournit en general au technicien du controle des reacteurs une information suffisante sur l'evolution du reacteur. Le physicien attache a la determination de parametres physiques ne peut se satisfaire de la mesure ordinaire de ces grandeurs, dont les concepts font abstraction de la nature du phenomene de la fission nucleaire. Habituellement le physicien doit donc convertir la mesure du temps de doublement, effectuee sur un temps suffisamment long, afin d'en eliminer les termes transitoires perturbateurs, et convertir, par intermediaire des courbes de Nordheim, cette mesure en valeur de reactivite. Cette procedure est longue et contraignante. Il a donc semble utile de concevoir un type d'appareil capable d'evaluer directement et instantanement le coefficient de multiplication excedentaire Greek-Small-Letter-Delta K a partir de l'evolution d'une grandeur physique N supposee proportionnelle au flux neutronique regnant dans le coeur du reacteur. Le coefficient Greek-Small-Letter-Delta K est pratiquement assimilable a la reactivite au voisinage de la criticalite. Un appareil de ce type peut en consequence resoudre la relation inverse a celle definie par le systeme des equations differentielles se rapportant au reacteur, en considerant le cas simplifie du reacteur point dans la theorie a un groupe. L'application des techniques du calcul analogique conduit a utiliser un reseau du type Pagels qui sera dispose, soit comme impedance d'entree, soit comme impedance de contre-reaction d'un amplificateur operationnel. Cette

Variation of the material laplacian of G1 with the radius of the uranium bar; Variation du laplacien matiere de G1 avec le rayon du barreau d'uranium

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

In this report are described and interpreted some experiments, carried out in the pile G1 during a period of shut-down, which have made it possible to measure the variation of the material Laplacian of the lattice with the radius of the uranium bar. The variation of the reactivity of the pile is measured when an increasing number of fuel elements are progressively replaced in the central region by fuel elements of greater diameter; it is shown that, starting from measurements based on less than ten per cent of the total number of elements, the variation of reactivity corresponding to the replacement of all the elements can be determined; it is then easy to deduce the variations of the Laplacian. Results: the variations of the Laplacian with the uranium rod diameter are 0 (d. 26 mm), +0.065 {+-} 0.004 m{sup -2} (d. 28 mm) and +0.080 {+-} 0.008 m{sup -2} (d. 32 mm). (author) [French] Dans ce rapport sont decrites et interpretees des experiences realisees sur la pile G1 'froide', experiences qui ont permis de mesurer la variation du Laplacien matiere du reseau avec le rayon du barreau d'uranium. On mesure la variation de reactivite de la pile lorsqu'on remplace progressivement dans la region centrale un nombre croissant de cartouches par des cartouches de plus gros diametre; on montre qu'a partir de mesures portant sur moins de dix pour cent du nombre total de cartouches, on peut determiner la variation de reactivite qui correspondrait au remplacement de toutes les cartouches; il est facile d'en deduire les variations du Laplacien. Resultats: les variations du Laplacien en fonction du diametre du barreau d'uranium sont: 0 (d. 26 mm), +0.065 {+-} 0.004 m{sup -2} (d. 28 mm) and +0.080 {+-} 0.008 m{sup -2} (d. 32 mm). (auteur)

Mesures de procédure spéciales et respect des droits de l'homme Rapport général

OpenAIRE

Vervaele, J.A.E.

2009-01-01

Le but du rapport général est de mener une analyse comparative des rapports nationaux en vue de présenter les processus de transformation des systèmes de justice pénale internes, en particulier du procès pénal, étant donné que des mesures procédurales spéciales sont introduites pour appréhender le terrorisme et la criminalité organisée, et de voir si cela a conduit les pays à se départir de leur propres règles fondamentales, procédures, principes et standards des droits de l’homme applicables...

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Measurement of the anti reactivity of a control rod of G1, by a slow oscillation method; Mesure de l'antireactivite d'une barre de reglage de G1 pour une methode d'oscillation lente

Energy Technology Data Exchange (ETDEWEB)

Breton, D; Leroy, J; Vidal, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

It is possible to determine the effect of the end of a control rod on the reactivity of the pile by measuring the modulation induced in the neutron flux by the slow oscillation of this control rod. The total effect of the control rod can be deduced, given certain hypothesis and corrections, from the experimental curve giving the effect of the end of the rod as a function of its position. This method has the advantage of permitting the measurement of very large anti reactivities, such as p= 10{sup -2} for example, which would not be possible by other kinetic methods. Thus the control rod B{sub 3}, in the low position, brings about a reduction in reactivity equal to 1130 p.c.m. {+-} 30 in the pile charged with 518 fuel elements, on one side only of the slit. We have compared the oscillation method with the classical divergence method, in the fields where the two measurements were possible: a satisfactory agreement was found. We have established that the phase displacement between the oscillation of the rod and the modulation of the flux varied greatly with the position of the rod. This variation cannot be explained on the basis of the dynamic model independent of space; we have attributed it to the influence of spatial harmonics of the flux distribution, and have determined a correction which frees the measurements of this influence. (author) [French] II est possible de determiner l'effet de l'extremite d'une barre de reglage sur la reactivite de la pile, a partir de la mesure de la modulation induite dans le flux neutronique par l'oscillation lente de cette barre de reglage. L'effet total de la barre de reglage peut etre deduit, moyennant certaines hypotheses et certaines corrections, de la courbe experimentale donnant l'effet de l'extremite de la barre en fonction de sa position. Cette methode a l'avantage de rendre possible la mesure d'antireactivites tres grandes, telles que p = 10{sup -2} par exemple, ce qui ne serait pas possible par d'autres methodes

Mesures de procédure spéciales et respect des droits de l'homme
Rapport général

OpenAIRE

John A.E. Vervaele

2009-01-01

Le but du rapport général est de mener une analyse comparative des rapports nationaux en vue de présenter les processus de transformation des systèmes de justice pénale internes, en particulier du procès pénal, étant donné que des mesures procédurales spéciales sont introduites pour appréhender le terrorisme et la criminalité organisée, et de voir si cela a conduit les pays à se départir de leur propres règles fondamentales, procédures, principes et standards des droits de l’homme applicables...

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

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

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

Measurement of a thermal neutron flux using air activation; Mesure de flux de neutrons thermiques par activation d'air

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

It is necessary to know, in irradiation loops, the thermal neutron flux after the irradiation device has been introduced and without being obliged to wait for the discharge of this device. In order to measure the flux and to control it continuously, one possible method is to place in the flux a coiled steel tube through which air passes. By measuring the activity of argon 41, and with a knowledge of the flow rate and the temperature of the air, it is possible to calculate the flux. An air-circulation flux controller is described and the relationship between the flux and the count rate is established The accuracy of an absolute measurement is about 14 per cent; that of a relative measurement is about 3 per cent. The measurement can be carried out equally well whether the reactor is operating at maximum or at low power. The measurement range goes from 10{sup 9} to lO{sup 15} n.cm{sup -2}.sec{sup -1}, and it would be possible after a few modifications to measure fluxes between 10{sup 5} and 10{sup 15} n.cm{sup -2}.sec{sup -1}. Finally, the method is very safe to operate: there is little risk of irradiation because of the low specific activity of the argon-41 formed, and no risk of contamination because the decay product of argon-41 is stable. This method, which is now being used in loops, is thus very practical. (authors) [French] Sur des boucles d'irradiation il est necessaire de connaitre le flux de neutrons thermiques apres mise en place du dispositif d'irradiation et sans etre oblige d'attendre le detournement de ce dispositif. Pour mesurer le flux et le controler en permanence, une methode consiste a placer sous flux un serpentin en acier dans lequel on fait circuler de l'air. La mesure d'activite d'argon 41 permet de calculer le flux, connaissant le debit et la temperature de l'air. Un controleur de flux par circulation d'air est decrit et la relation entre le flux et le taux de comptage est etablie. La precision d'une mesure absolue est de l'ordre de 14 pour

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

Energy Technology Data Exchange (ETDEWEB)

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

1963-10-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

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

Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

International Nuclear Information System (INIS)

1992-04-01

Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

OSIRIS reactor radioprotection, radioprotection measurements performed during the power rise and the first 50 megawatt operation; Radioprotection de la pile OSIRIS, mesures de radioprotection effectuees au cours de la montee en puissance et des premiers fonctionnements a 50 megawatts

Energy Technology Data Exchange (ETDEWEB)

Fanton, B; Lebouleux, P

1967-12-01

The authors supply the results of the measurements that have been made near the Osiris reactor during the power increase and during the first functioning at 50 megawatts. The measurements relate to the absorbed dose rates in the premises, the water activation and the atmospheric contamination. The influence of the heat layer of water movements and the water rate in the core chimney on the absorbed dose rate at the footbridge level overhanging the pile core has been studied. The modifications to the protection devices that have been proposed after the measurements and the effect of these modifications on the results of the measures are given then. The regeneration process of a water purification chain has been examined from the radiation protection point of view. It has been possible to make some twenty radionuclides obvious in the produced effluents and to determine the volume activity of these effluents for each radionuclide. The whole of results show that in a general way, the irradiation levels are low during the usual reactor functioning. [French] Les auteurs fournissent les resultats des mesures de radioprotection oui ont ete effectuees aupres de la pile Osiris pendant la montee en puissance et au cours des premiers fonctionnements a 50 megawatts. Les mesures portent sur les debits de dose absorbee dans les locaux, l'activation de l'eau et la contamination atmospherique. L'influence de la couche chaude des mouvements d'eau et du debit d'eau dans la cheminee du coeur sur le debit de dose absorbee au niveau de la passerelle surplombant le coeur de la pile, a ete etudiee. Les modifications aux dispositifs de protection, qui ont ete proposees a la suite des mesures, et l'effet de ces modifications sur les resultats des mesures sont indiques ensuite. Le processus de regeneration d'une chaine d'epuration de l'eau a ete examine sous l'angle de la radioprotection. Il a ete possible de mettre en evidence une vingtaine de radionucleides dans les effluents produits et de

Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

Les fondements de la mesure du temps comment les fréquences atomiques règlent le monde

CERN Document Server

Audoin, Claude

1998-01-01

La mesure du temps fondée sur des propriétés atomiques est née en 1955, avec le premier étalon de fréquence à jet de césium. Depuis, les horloges atomiques ne cessent de progresser et sont au coeur de nombreuses activités, telles que les comparaisons de temps, l'unification mondiale de l'heure ou la recherche en astronomie, géodésie, géophysique, télécommunications, etc. Cet ouvrage fournira des réponses détaillées au lecteur intéressé par la mesure du temps appliquée aux divers domaines cités.

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

Mesures de teneurs en eau volumique et massique sur du sable

OpenAIRE

FAUCHARD, Cyrille; GUILBERT, Vincent; SAGNARD, Florence; FROUMENTIN, Michel

2009-01-01

Usuellement, la mesure par séchage en étuve et la mesure de la masse volumique humide par méthode nucléaire permettent de relier la teneur en eau massique à la teneur en eau volumique. La méthodologie étudiée ici propose d'associer à la mesure de la masse volumique humide une estimation de la teneur en eau volumique par des méthodes électromagnétiques via la mesure de la permittivité diélectrique apparente du sol. Ce paramètre physique peut être relié à la teneur en eau volumique par une loi ...

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Gamma spectrum measurement in a swimming-pool-type reactor; Mesure du spectre {gamma} d'une pile piscine

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

After recalling the various modes of interaction of gamma rays with matter, the authors describe the design of a spectrometer for gamma energies of between 0.3 and 10 MeV. This spectrometer makes use of the Compton and pair-production effects without eliminating them. The collimator, the crystals and the electronics have been studied in detail and are described in their final form. The problem of calibrating the apparatus is then considered ; numerous graphs are given. The sensitivity of the spectrometer for different energies is determined mainly for the 'Compton effect' group. Finally, in the last part of the report, are given results of an experimental measurement of the gamma spectrum of a swimming-pool type reactor with new elements. (author) [French] Apres un rappel des differents modes d'interaction des rayons gamma avec la matiere, nous decrivons la conception d'un spectrometre pour les energies gamma s'etendant de 0,3 a 10 MeV. Ce spectrometre utilise les effets Compton et creation de paires sans les eliminer. Le collimateur, les cristaux et l'electronique sont entierement etudies et decrits dans leur realisation definitive. Ensuite, le probleme de l'etalonnage de l'appareil est envisage; de nombreuses courbes sont donnees. La sensibilite du spectrometre pour les differentes energies est determinee principalement pour le groupe ''effet Compton''. Enfin, les resultats d'une experience de mesure du spectre gamma d'une pile piscine avec elements neufs sont donnes dans la derniere partie. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-10-15

generateur: le temps de mesure est donc minimum. Les observations enregistrees sur bande perforee sont depouillees par une calculatrice numerique. (author) [Spanish] Los laboratorios de investigacion Philips han construido un generador neutronico de flujo variable en funcion del tiempo. Con una serie de mediciones efectuadas en el reactor BRO2 en estado subcrftico, se ha demostrado su utililidad practica en la esfera de la fisica de los reactores. El funcionamiento del generador es muy flexible debido a su alta estabilidad, a la posibilidad de variar bruscamente la intensidad neutronica, y de pulsar el flujo o modularlo de manera sinusoidal. El generador permite determinar la reactividad ({rho} = {Delta}k/{beta}) y la vida media de los neutrones ( Script-Small-L /{beta}) segun varios metodos independientes. Es posible proceder a una comparacion exacta de esos metodos, dado que pueden aplicarse sin modificar las condiciones de medicion. El autor ha calculado los siguientes valores: a) p, sobre la base de los neutrones retardados, por reduccion instantanea del flujo neutronico; b) p, sobre la base de los neutrones inmediatos, por impulsos neutronicos; c) Script-Small-L /{beta}, combinando 1) y 2), cuando 0, 5 dolares < {rho} < 2 dolares, y d) Script-Small-L /{beta}, sobre la base de la funcion de transferencia del reactor para una fuente modulada. En la memoria se examinan las funciones de transferencia correspondientes a un oscilador de reactividad y a una fuente de modulacion sinusoidal. Se demuestra que es posible medir Script-Small-L /{beta}, cuando 0,1 dolar < {rho} < 10 dolares utilizando una fuente modulada. Por el mismo metodo se obtiene tambien la reactividad partiendo de la razon neutrones inmediatos/neutrones retardados para una frecuencia optima que es practicamente independiente de los datos relativos a los neutrones retardados y del cociente Script-Small-L /{beta}. La precision estadistica de cada metodo puede aumentarse acumulando un gran numero de ciclos en el

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

OSIRIS reactor radioprotection, radioprotection measurements performed during the power rise and the first 50 megawatt operation; Radioprotection de la pile OSIRIS, mesures de radioprotection effectuees au cours de la montee en puissance et des premiers fonctionnements a 50 megawatts

Energy Technology Data Exchange (ETDEWEB)

Fanton, B.; Lebouleux, P

1967-12-01

The authors supply the results of the measurements that have been made near the Osiris reactor during the power increase and during the first functioning at 50 megawatts. The measurements relate to the absorbed dose rates in the premises, the water activation and the atmospheric contamination. The influence of the heat layer of water movements and the water rate in the core chimney on the absorbed dose rate at the footbridge level overhanging the pile core has been studied. The modifications to the protection devices that have been proposed after the measurements and the effect of these modifications on the results of the measures are given then. The regeneration process of a water purification chain has been examined from the radiation protection point of view. It has been possible to make some twenty radionuclides obvious in the produced effluents and to determine the volume activity of these effluents for each radionuclide. The whole of results show that in a general way, the irradiation levels are low during the usual reactor functioning. [French] Les auteurs fournissent les resultats des mesures de radioprotection oui ont ete effectuees aupres de la pile Osiris pendant la montee en puissance et au cours des premiers fonctionnements a 50 megawatts. Les mesures portent sur les debits de dose absorbee dans les locaux, l'activation de l'eau et la contamination atmospherique. L'influence de la couche chaude des mouvements d'eau et du debit d'eau dans la cheminee du coeur sur le debit de dose absorbee au niveau de la passerelle surplombant le coeur de la pile, a ete etudiee. Les modifications aux dispositifs de protection, qui ont ete proposees a la suite des mesures, et l'effet de ces modifications sur les resultats des mesures sont indiques ensuite. Le processus de regeneration d'une chaine d'epuration de l'eau a ete examine sous l'angle de la radioprotection. Il a ete possible de mettre en evidence une vingtaine

Current status and future plan of the G.A. Siwabessy Multipurpose Reactor

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim; Mardi, Alfahari [Centre for Multipurpose Reactor, National Atomic Energy Agency, Serpong (Indonesia)

1998-10-01

Since the first criticality in July 1987, the G.A. Siwabessy Multipurpose Reactor, RSG-GAS, in Serpong-Indonesia has been operated about 26.000 hours up to now. In the last two years the reactor is operated more than 5000 hours per year or equivalent to seven cycles a year. The reactor is utilized for conducting research studies and isotopes production. In the near future the core will be converted to silicide fuel to improve the core performance instead of oxide fuel. The planned maintenance activities are performed according to the schedule. The modifications and remedial maintenance are also performed to ensure that all structures, systems and important to safety are capable of performing as intended. The future activities of the G.A. Siwabessy reactor will be stressed to maintain the reliability and availability of the reactor operation and to optimize the reactor utilization. (author)

Les mesures de métrologie pour le CLIC

CERN Document Server

Cherif, A

2008-01-01

Le projet CLIC est en tout point un défi technique majeur ; c?est le cas également pour la mesure dimensionnelle. Quels sont les équipements et les méthodes qui permettent de caractériser les pièces avec une incertitude de mesure aussi réduite que possible, vu les tolérances micrométriques imposées ? Afin de répondre à cette question, une veille technologique a été maintenue sur une longue période. Les acteurs relevants ont été contactés pour bénéficier d?une ouverture sur les dernières avancées dans le domaine. Différentes techniques ont été étudiées et comparées telles que la digitalisation, la tomographie X, la mesure tridimensionnelle. L'assemblage de haute précision des composants est aussi primordial. Sa mise en ?uvre sous un microscope optique ou à l'aide d'une machine tridimensionnelle est en cours d?étude. L'exposé traitera aussi de la mesure de rugosité, un domaine où nous disposons de moyens adaptés aux exigences spécifiques du projet.

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

method. A scintillator and an electronic system provide a specific signal of the fission products which is then marked on a recorder. In a case where the activity threshold is exceeded, the cell involved is isolated from the prospection system and taker, over by a 'follow-up' detector which follows the evolution of the crack. A year of working on the pile G{sub 1}, which is cooled by air at atmospheric pressure, has made it possible to obtain results on the operation of the canning-burst detection appliance, which has led us to perfect the original device by installing an 'evolution-meter' of the type described above for G{sub 3}. The reactor EL{sub 3}, cooled by heavy water, uses a detection system based on the measurement by GM counters of the activity of the fission gases carried by diluted helium into the heavy water, then extracted by hydro-cyclones. The selectivity of the system gives it a low sensitivity to parasite activities, and an excellent performance. (author) [French] Dans les piles refroidies par gaz carbonique sous pression, du type G{sub 3}, la radioactivite principale du gaz est celle de l'azote 16 creee par reaction {sup 16}O(n, p) {sup 16}N des neutrons rapides sur l'oxygene. Cette activite, de vie courte et de forte energie {beta}, masque l'activite des gaz de fission s'echappant par une fissure de gaine dans le gaz carbonique et oblige a utiliser une methode de separation materielle des produits de fission solides avant la detection proprement dite. Cette detection est faite par une chaine electronique speciale dont l'entree est un scintillateur associe a un photomultiplicateur. Un systeme de mesure d'evolution de fissure avec compensation des variations de puissance permet de suivre la vitesse d'evolution d'une fissure. Cet appareil, baptise evolumetre, est destine a ramener a une methode de zero la mesure de l'activite du gaz de refroidissement des canaux, il permet de s'affranchir: 1) de l'activite propre du gaz restant apres la discrimination

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-07-01

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

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

Canadian supercritical water reactor modeling using G4STORK

International Nuclear Information System (INIS)

Ford, W.; Buijs, A.

2015-01-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Canadian supercritical water reactor modeling using G4STORK

Energy Technology Data Exchange (ETDEWEB)

Ford, W.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Preliminary assessment of an S.G.H.W. type research reactor

International Nuclear Information System (INIS)

Bicevskis, A.; Chapman, A.G.; Hesse, E.W.

1970-08-01

A preliminary design study has been made of a research reactor, based on the enriched S.G.H.W.R. concept, to be used for power reactor fuel irradiation, isotope production, basic research, and training in nuclear technology. A reactor physics assessment established a core size which would allow uninterrupted operation for the required irradiation period consistent with low capital and operating costs. A design was selected with 24 channels, a D 2 O calandria diameter of 2.7 m and an overall core height of 4.0 m. The capital cost was estimated as $750,000 for the fuel and $1,600,000 for the moderator, the refuelling cost being $340,000 per annum. A thermal design study showed that the fission heat of 65 MW could be transmitted to pressurised light water at 200 lb/in 2 abs. and rejected to sea water in two conventional U-tube heat exchangers. The basic design is flexible and can be adapted to meet many special requirements. (author)

Mesure asymétrie avant-arriere des quarks lourds a LEP1 avec le détecteur OPAL

CERN Document Server

Lafoux, H

A partir de l'ensemble des données accumulées par OPAL au cours de la première phase de fonctionnement du LEP, nous avons mesuré l'asymétrie avant-arrière des quarks b et c au voisinage du pic du Zo. Utilisant une méthode traditionnelle, basée sur la détection des leptons produits dans les désintégrations semi-leptoniques des hadrons lourds, nous avons cherché à optimiser chaque étape de la mesure, en mettant en œuvre les algorithmes les plus appropriés. Le recours aux réseaux de neurones artificiels s'est en particulier avéré d'une grande utilité lorsque le problème à résoudre impliquait la prise en compte simultanée de multiples sources d'informations, d'origine et de nature très variées. Nos résultats sont en bon accord avec ceux des autres mesures effectuées à LEP et compatibles avec les prédictions du Modèle Standard pour un quark top de 174 ± 31 GeV/c2 et un boson de Higgs de masse comprise entre 60 et 1000 GeV/c2

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

1994-01-01

The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

Proceedings of the seminar on optimization technology of the use of G.A. Siwabessy Research Reactor

International Nuclear Information System (INIS)

1999-01-01

Seminar on optimization technology of the use of G.A. Siwabessy research reactor was held on March 16, 1999 at the Multipurpose Reactor Center, Serpong, Indonesia. During the seminar, have presented 14 papers about activities or researches on reactor operation technology, use of G.A. Siwabessy research reactor, engineering and nuclear installation development, maintenance and quality assurances. The seminar was held as a tool for developing non-researcher functional workers

Proceedings of the seminar on optimization technology of the use of G.A. Siwabessy Research Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

Seminar on optimization technology of the use of G.A. Siwabessy research reactor was held on March 16, 1999 at the Multipurpose Reactor Center, Serpong, Indonesia. During the seminar, have presented 14 papers about activities or researches on reactor operation technology, use of G.A. Siwabessy research reactor, engineering and nuclear installation development, maintenance and quality assurances. The seminar was held as a tool for developing non-researcher functional workers.

Decommissioning, Dismantling and Disarming: a Unique Information Showroom Inside the G2 Reactor at Marcoule Centre (France) - 12068

Energy Technology Data Exchange (ETDEWEB)

Volant, Emmanuelle [CEA DAM, Bruyeres-le-Chatel (France); Garnier, Cedric [CEA DEN, Marcoule (France)

2012-07-01

The paper aims at presenting the new information showroom called 'Escom G2' (for 'Espace Communication') inaugurated by the French Atomic Energy and Alternative Energies Commission (CEA) in spring 2011. This showroom is settled directly inside the main building of the G2 nuclear reactor: a facility formerly dedicated to weapon-grade plutonium production since the late 1950's at the Marcoule nuclear centre, in south of France. After its shutdown, and reprocessing of the last spent fuels, a first dismantling step was successfully completed from 1986 to 1996. Unique in France and in Europe, Escom G2 is focused on France dismantling expertise and its action for disarmament. This showroom comprises of a 300-square meters permanent exhibition, organized around four themes: France strategy for disarmament, decommissioning and dismantling technical aspects, uranium and plutonium production cycles. Each of these topics is illustrated with posters, photos, models and technical pieces from the dismantled plants. It is now used to present France's action in disarmament to highly ranked audiences such as: state representatives, diplomats, journalists... The paper explains the background story of this original project. As a matter of fact, in 1996 France was the first nuclear state to decide to shut down and dismantle its fissile material production facilities for nuclear weapons. First, the paper presents the history of the G2 reactor in the early ages of Marcoule site, its operating highlights as well as its main dismantling operations, are presented. In Marcoule, where the three industrial-scale reactors G1, G2 and G3 used to be operated for plutonium production (to be then reprocessed in the nearby UP1 plant), the initial dismantling phase has now been completed (in 1980's for G1 and in 1996 for G2 and G3). The second phase, aimed at completely dismantling these three reactors, will restart in 2020, and is directly linked to the opening of

Results and interpretation of spectral indices measurements made with AQUILON; Resultats et interpretation de mesures d'indices de spectre dans aquilon

Energy Technology Data Exchange (ETDEWEB)

Frichet, J P; Mougey, J N; Naudet, R; Taste, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

This report deals with a set of spectral indices measurements made in the heavy water reactor Aquilon on lattices constituted by massive fuel elements of dia. 29,2 mm. The fuel elements were made either of natural uranium or of slightly depleted or slightly enriched uranium, or of an uranium-plutonium alloy. The measurements were carried out for various lattice pitches (square pitch from 110 to 210 mm) and in certain cases for various temperatures (from 20 to 80 deg. C). The results are compared to calculated values obtained by using the latest advances of the thermalization theory developed at Saclay applied to the moderation by heavy water. (authors) [French] Ce rapport est consacre a un ensemble de mesures d'indices de spectre realisees dans la pile a eau lourde Aquilon sur des reseaux d'elements combustibles pleins, de 29,2 mm de diametre. Ces combustibles se composaient ou bien d'uranium naturel, ou bien d'uranium tres legerement appauvri ou enrichi, ou bien d'un alliage uranium plutonium. Les mesures ont ete effectuees pour toute une serie de pas de reseaux (pas carre 110 a 210 mm), certaines d'entre elles a plusieurs temperatures (20 a 80 deg. C). Les resultats des mesures sont compares a des valeurs calculees obtenues en utilisant les plus recents developpements de la theorie de la thermalisation mise au point a Saclay, appliques au cas de la moderation par l'eau lourde. (auteurs)

Dispositif de mesure de pertes dans les conducteurs supraconducteurs utilisés en régime variable

Science.gov (United States)

Le Naour, S.; Lacaze, A.; Laumond, Y.

1998-01-01

A thermometric apparatus to measure AC losses in superconductor wires for 50 Hz applications is described. This method consists in isolating the sample from a helium bath via a thermal resistance. Dissipated power is determined by two thermometers located on both edges of a thermal resistance. The measurement's calibration is done using an ohmic heater. The measurement accuracy is 10% for losses in excess of 2 mW/m. Un dispositif expérimental, pour mesurer les pertes générées dans les conducteurs supraconducteurs utilisés en régime alternatif 50 Hz, est décrit. La méthode, basée sur le principe thermométrique, consiste à isoler l'échantillon du bain d'hélium par une résistance thermique. La puissance dissipée est déterminée à l'aide de deux sondes de température disposées de part et d'autre de la résistance. L'étalonnage de la mesure est assuré par une chaufferette. La précision des mesures est de 10% pour des pertes linéiques supérieures à 2 mW/m.

Observations on the electronic equipment employed for making measurements on the pile G1; Observations sur le materiel electronique utilise pour les mesures sur la pile G1

Energy Technology Data Exchange (ETDEWEB)

Ailloud, J; Belin, P; Meunier, A; Tarabella, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The electronic apparatus employed during the manipulations carried out on the pile G1 is briefly described, with the aim of putting on record the inconveniences encountered in the course of the operation of this equipment. (author) [French] On decrit succinctement l'appareillage electronique utilise durant les manipulations effectuees sur la pile G1, dans le but de noter les inconvenients rencontres au cours de l'exploitation de cet appareillage. (auteur)

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

La productivité et sa mesure en France (1944-1955)

OpenAIRE

Boulat, Régis

2009-01-01

Après la Deuxième Guerre mondiale, grâce à Jean Fourastié et aux modernisateurs du Plan, la productivité est reconnue par les élites et par l'opinion publique française comme « la clef » de la connaissance économique et le moteur de la croissance. Alors que la IVe République institutionnalise une politique d'accroissement de la prductivité dans le cadre d'un programme franco-américain de modernisation, la mesure de la productivité se perfectionne lentement grâce à l'affinement des mesures de ...

MESUR metrics from scholarly usage of resources

CERN Document Server

CERN. Geneva; Van de Sompel, Herbert

2007-01-01

Usage data is increasingly regarded as a valuable resource in the assessment of scholarly communication items. However, the development of quantitative, usage-based indicators of scholarly impact is still in its infancy. The Digital Library Research & Prototyping Team at the Los Alamos National Laboratory's Research library has therefore started a program to expand the set of usage-based tools for the assessment of scholarly communication items. The two-year MESUR project, funded by the Andrew W. Mellon Foundation, aims to define and validate a range of usage-based impact metrics, and issue guidelines with regards to their characteristics and proper application. The MESUR project is constructing a large-scale semantic model of the scholarly community that seamlessly integrates a wide range of bibliographic, citation and usage data. Functioning as a reference data set, this model is analyzed to characterize the intricate networks of typed relationships that exist in the scholarly community. The resulting c...

Fast measure proceeding of weak currents; Un procede de mesure rapide des courants faibles

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

The process of fast measure of the weak currents that we are going to describe briefly apply worthy of the provided currents by the sources to elevated value internal resistance, as it is the case for the ionization chamber, the photocells, mass spectroscopic tubes. The problem to measure weak currents is essentially a problem of amplifier and of input circuit. We intended to achieve a whole amplifier and input circuit with advanced performances, meaning that for a measured celerity we wanted to have an signal/noise ratio the most important as in the classic systems and for a same report signal/noise a more quickly done measure. (M.B.) [French] Le procede de mesure rapide des courants faibles que nous allons brievement decrire s'applique a la mesure des courants fournis par les sources a resistance interne de valeur elevee, comme c'est le cas pour les chambres d'ionisation, les photocellules, les tubes de spectrographe de masse. Le probleme de mesure de courants faibles est essentiellement un probleme d'amplificateur et de circuit d'entree. Nous nous sommes proposes de realiser un ensemble amplificateur et circuit d'entree a performances poussees, c'est a dire que pour une meme rapidite de mesure nous desirions avoir un rapport signal/bruit plus important que dans les systemes classiques et pour un meme rapport signal/bruit une mesure effectuee plus rapidement. (M.B.)

Les mesures de gel et de confiscation en vertu de la Loi sur les embargos

OpenAIRE

Schnyder, Nicolas

2009-01-01

Etude du mécanisme d'adoption des mesures de gel et de confiscation en vertu de la Loi sur les embargos et leurs conséquences tant sur la personne principalement touchée par ces mesures que sur les tiers.

Perceptions locales du changement climatique et mesures d ...

African Journals Online (AJOL)

Perceptions locales du changement climatique et mesures d'adaptation dans la ... village, de l'importance représentée par le karité pour les groupes socioculturels. ... l'adaptation de nouvelles cultures en association, la protection des jeunes ...

The reactor Melusine - radiation measurements carried out at the start of operation and during the first ascents to power; Pile Melusine - mesures de rayonnement effectuees au demarrage et pendant les premieres montees en puissance

Energy Technology Data Exchange (ETDEWEB)

Coutrot, V; Delpuech, J; Fitoussi, L [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This report gives the results of radiation measurements carried out on the first C.E.A. swimming-pool pile, Melusine. The purpose of these measurements, which were carried out during the starting-up period and the first ascents to 1 MW power, is to check the radiation intensity levels near the pile and from this to verify the safety calculation methods used. In addition certain special measurements, such as those performed in and above the water of the swimming-pool, in the channels and outside them etc..., should make it possible eventually to define the conditions under which the pile may be used for special jobs with less protection. In the first part of the report are given the results of radiation measurements carried out during runs at low power plateaux not exceeding 500 W: in particular measurements at variable water levels in the water of the swimming-pool and on the axis of the open channels. The results given in the second part deal with radiation measurements performed in various parts of the premises, studies on the activation of the plugs in the experimental channels and of the materials in the forward compartment, and also of the radioactivity of the swimming-pool water and the air used to cool the channels. (author) [French] Le present rapport a pour but de donner les resultats des mesures de rayonnement effectuees aupres de la premiere pile piscine du C.E.A.: la pile Melusine. Le but de ces mesures, effectuees pendant la periode de demarrage et les premieres montees a la puissance de 1 MW, est de controler les niveaux d'intensite de rayonnement aupres de la pile et par la de juger des methodes de calculs de protection utilisees. D'autre part, certaines mesures plus particulieres, telles que celles effectuees dans l'eau et au-dessus de l'eau de la piscine, dans les canaux et a l'exterieur de ceux-ci, etc..., doivent permettre de definir ulterieurement les conditions d'utilisation de la pile pour des fonctionnements particuliers avec des protections

Mesurer en réseau d'assainissement pour quoi faire ?

OpenAIRE

LAPLACE, Dominique; JOANNIS, Claude; GUIVARCH, Jean Yves

2009-01-01

La mesure de pluie, de niveau d'eau, de vitesse, de débit ou encore de pollution en réseau d'assainissement répond essentiellement à des besoins de contrôle du bon fonctionnement du système, de compréhension et d'amélioration de ces réseaux et d'information des différents acteurs impliqués. Associée à un dispositif de gestion en temps la mesure contribue à permettre de surveiller et maîtriser le fonctionnement du réseau par temps sec et par temps de pluie, et de piloter des actionneurs (pompe...

MESUR: metrics from scholarly usage of resources

CERN Multimedia

CERN. Geneva

2007-01-01

The MESUR project is constructing a large-scale semantic model of the scholarly community that seamlessly integrates a wide range of bibliographic, citation and usage data. Functioning as a reference data set, this model is analyzed to characterize the intricate networks of typed relationships that exist in the scholarly community. The resulting ...

Neutronics analysis of Nigerian Research Reactor-1

International Nuclear Information System (INIS)

Azande, T.S.; Balogun, G.I.

2010-01-01

Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

2013-07-01

Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

MESUR: USAGE-BASED METRICS OF SCHOLARLY IMPACT

Energy Technology Data Exchange (ETDEWEB)

BOLLEN, JOHAN [Los Alamos National Laboratory; RODRIGUEZ, MARKO A. [Los Alamos National Laboratory; VAN DE SOMPEL, HERBERT [Los Alamos National Laboratory

2007-01-30

The evaluation of scholarly communication items is now largely a matter of expert opinion or metrics derived from citation data. Both approaches can fail to take into account the myriad of factors that shape scholarly impact. Usage data has emerged as a promising complement to existing methods o fassessment but the formal groundwork to reliably and validly apply usage-based metrics of schlolarly impact is lacking. The Andrew W. Mellon Foundation funded MESUR project constitutes a systematic effort to define, validate and cross-validate a range of usage-based metrics of schlolarly impact by creating a semantic model of the scholarly communication process. The constructed model will serve as the basis of a creating a large-scale semantic network that seamlessly relates citation, bibliographic and usage data from a variety of sources. A subsequent program that uses the established semantic network as a reference data set will determine the characteristics and semantics of a variety of usage-based metrics of schlolarly impact. This paper outlines the architecture and methodology adopted by the MESUR project and its future direction.

Measurements of fission cross-sections and of neutron production rates; Mesures de sections efficaces de fission et du nombre de neutrons prompts emis par fission

Energy Technology Data Exchange (ETDEWEB)

Billaud, P; Clair, C; Gaudin, M; Genin, R; Joly, R; Leroy, J L; Michaudon, A; Ouvry, J; Signarbieux, C; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

a) Measurements of neutron induced fission cross-sections in the low energy region. The variation of the fission cross sections of several fissile isotopes has been measured and analysed, for neutron energies below 0,025 eV. The monochromator was a crystal spectrometer used in conjunction with a mechanical velocity selector removing higher order Bragg reflections. The fissile material was laid down on the plates of a fission chamber by painting technic. An ionization chamber, having its plates coated with thin {sup 10}B layers, was used as the neutron flux monitor. b) Measurement of the fission cross section of {sup 235}U. We intend to measure the variation of the neutron induced fission cross section of {sup 235}U over the neutron energy range from 1 keV by the time of flight method. The neutron source is the uranium target of a pulsed 28 MeV electron linear accelerator. The detector is a large fission chamber, with parallel plates, containing about 10 g of {sup 235}U (20 deposits of 25 cm diameter). The relative fission data were corrected for the neutron spectrum measured with a set of BF{sub 3} proportional counters. c) Mean number {nu} of neutrons emitted in neutron induced fission. We measured the value of {nu} for several fissile isotopes in the case of fission induced by 14 MeV neutrons. The 14 MeV neutrons were produced by D (t, n) {alpha} reaction by means of a 300 kV Cockcroft Walton generator. (author)Fren. [French] a) Mesures de sectionficaces de fission a basse energie. Nous avons mesure et analyse la variation de la section efficace de fission de divers isotopes fissiles pour des neutrons d'energie inferieure a 0,025 eV. Le monochromateur est constitue par un spectrometre a cristal auquel est associe un selecteur mecanique destine a eliminer les diffractions de Bragg d'ordre superieur au premier. Le materiau fissile est contenu dans une chambre a fission sous forme de depots realises par peinture; une chambre d'ionisation a depots minces de B{sub 10

Checking the sealing of fuel elements by helium sweating - case of the reactors G2 (1960); Controle de l'etancheite des elements combustibles par ressuage d'helium - cas du reacteur G2 (1960)

Energy Technology Data Exchange (ETDEWEB)

Blanc, B; D' Orival, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Choumoff, S [Compagnie Francaise Thomson-Houston, 75 - Paris (France)

1960-07-01

The G2 slug is a welded, hermetically sealed unit; the seal is checked by placing the fuel element in a helium atmosphere under pressure, then measuring the quantity of helium it releases in a vessel under vacuum. The theoretical aspect and the conditions of industrial application are reviewed, and the installations described. (author) [French] La cartouche G2 se presente comme un ensemble soude, hermetique; le controle d'etancheite s'effectue en immergeant l'element combustible dans une atmosphere d'helium sous pression puis en mesurant la quantite d'helium qu'il restitue dans une enceinte sous vide. L'aspect theorique et les conditions d'exploitation industrielle sont evoques et les installations decrites. (auteur)

Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

International Nuclear Information System (INIS)

Perry, E.F.

1995-01-01

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Annex I.G. Demolition of the G1 stack at Marcoule by toppling

International Nuclear Information System (INIS)

2005-01-01

The G1 stack at Marcoule was constructed during the first half of 1956 as a ventilation outlet for the G1 reactor, which is cooled by air. After the G1 reactor was decommissioned, the G1 stack served as a ventilation outlet for two new nuclear facilities on the site. Being no longer in compliance with regulations and having many inadequacies and uncertainties in terms of the prestressed concrete, the stack posed a potential damage risk in extreme wind or in the event of an earthquake. In 1994 it was decided that a new stack would be built to act as an outlet for the existing nuclear facilities, and that the old one would be demolished. The G1 stack was 100 m in height, 10 m in diameter and constructed with 24 vertically stacked concrete rings consisting of nine prefabricated sections, each 3.6 m in height. It was capped by a metal deflector (about 6 m in height and weighing 50 t). The inside consisted of nine semicircular tubes constructed of steel sheet metal weighing 120 t. The base of the stack consisted of the foundation, a plate and a base plate which were constructed at the site. The barrel sections were prefabricated. Construction lasted from January 1956 to June 1956. At the base, the cylindrical portion of the stack widened to form three feet extending to a depth of 7.5 m. The base plate of the stack was formed onsite to the height of 16.7 m and then prestressed using cables. A repair carried out in 1964 included adding a concrete lining of the initial rings of the cylinder up to a level of 22.1 m. Additional prestressing with the base plate and repair of the horizontal and vertical prestressing of the barrel were also carried out, leaving only 22 rings and 43 visible cables. The total mass of the stack was 2170 t, including: - Concrete: cylinder 800 t, base plate 1200 t; - Steel: internal structures 120 t, deflector 50 t. The main radiological risk was the presence of traces of tritium. The radioactive inventory for the entire stack was estimated in 2000

Nuclear material control at IEA-R1 nuclear research reactor

International Nuclear Information System (INIS)

1988-01-01

The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

Extensive utilisation of VR-1 reactor for nuclear education and training

International Nuclear Information System (INIS)

Rataj, J.

2010-01-01

The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

Theoretical and experimental study of a calorimetric technique for measuring energy deposition in materials caused by complex pile irradiation; Etude theorique et experimentale d'une technique calorimetrique de mesure des depots d'energie dans les materiaux dus au rayonnement complexe de pile

Energy Technology Data Exchange (ETDEWEB)

Mas, P; Sciers, P; Droulers, Y [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires, 38 - Grenoble (France)

1962-07-01

Calorimetric methods may be used to measure gamma fluxes greater than 10{sup 6} r/h near the cores of swimming pool reactors. The theory, design, and properties of isothermal calorimeters are discussed, and experimental results obtained with two types are presented. Measurement of energy deposition in materials and the long term integration of energy depositions are other uses of these devices. Results of measurements on heat deposition in steel and water are given. Fluxes were also measured. (authors) [French] Une premiere partie traite de la theorie des calorimetres isothermes mis en oeuvre au C.K.N. Grenoble. La puissance deposee dans le calorimetre par les flux de rayonnement echauffe celui-ci. L'echauffement est mesure a l'aide d'un thermocouple. On montre que l'on a ainsi une mesure absolue de cette puissance. Une deuxieme partie traite de l'etude experimentale de: deux types d'appareils utilises: leur construction, les resultats experimentaux, leurs utilisations. Trois de celles-ci sont particulierement interessantes: - la mesure des hauts flux gamma, - la mesure du depot d'energie dans les materiaux, - l'integration pendant une longue duree des depots d'energie (un modele de calorimetre a fonctionne a ce jour 2 500 heures et a integre 9 x 10 puissance 10 rads gamma et 6 x 10 puissance 18 neutrons rapides). La troisieme partie est consacree a l'etude des qualites de l'appareil: robustesse, fidelite, precision, sensibilite, gamme de mesure. Enfin dans la derniere partie sont decrites deux applications de la methode calorimetrique a la mesure du depot d'energie dans un acier special et dans l'eau. (auteurs)

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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

The success of operation and utilization of the Indonesia multipurpose reactor G.A. Siwabessy

International Nuclear Information System (INIS)

Taryo, Taswanda; Kuntoro, Iman

2000-01-01

The Indonesia Multipurpose Reactor G.A. Siwabessy (RSG-GAS), operated by Multipurpose Reactor Center (MPRC/PRSG-BATAN), went its first criticality in July 1987. The reactor then achieved the power of 30 MW thermal in March 1992. Based on user requirement, the reactor is usually operated at the power of 20 MW thermal. The RSG-GAS is put to use mainly for radioisotope production, R and D on reactor safety and by using beam tubes, the reactor can also be applied for R and D on science and materials. Operation and maintenance of the reactor have been well organized due to well technical and administrative management from the top manager to all people involved in those two activities. Within their support, the RSG-GAS has occupied great advantages not only for man power development in our center but also for scientific cooperation with whoever would like to apply the RSG-GAS for R and D with mutual benefit agreement. (author)

Gas loop - continuous measurement of thermal and fast neutron fluxes; Boucle a gaz - mesure continue de flux de neutrons thermiques et rapides

Energy Technology Data Exchange (ETDEWEB)

Droulers, Y; Pleyber, G; Sciers, P; Maurin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

The measurement method described in this report can be applied both to thermal and fast neutron fluxes. A description is given of two practical applications in each of these two domains. This method is particularly suitable for measurements carried out on 'loop' type equipment. The measurement of the relative flux variations are carried out with an accuracy of 5 per cent. The choice of the shape of the gas circuit leaves a considerable amount of liberty for the adaptation of the measurement circuit to the experimental conditions. (authors) [French] La methode de mesure defrite dans ce rapport s1 applique aussi bien au flux de neutrons thermiques, qu'au flux de neutrons rapides. On donne la description de deux realisations pratiques dans chacun de ces domaines. Cette methode est particulierement adaptee a des mesures effectuees sur des dispositifs du type 'boucle'. La mesure des variations relatives de flux se fait avec une precision de 5 pour cent. Le choix de la configuration du circuit gazeux donne une grande souplesse dans l'adaptation du circuit de mesure aux conditions experimentales. (auteurs)

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

Ethanol production from wet-exploded wheat straw hydrolysate by thermophilic anaerobic bacterium Thermoanaerobacter BG1L1 in a continuous immobilized reactor

DEFF Research Database (Denmark)

Georgieva, Tania I.; Mikkelsen, Marie Just; Ahring, Birgitte Kiær

2008-01-01

was not detoxified, ethanol yield in a range of 0.39-0.42 g/g was obtained. Overall, sugar efficiency to ethanol was 68-76%. The reactor was operated continuously for approximately 143 days, and no contamination was seen without the use of any agent for preventing bacterial infections. The tested microorganism has......Thermophilic ethanol fermentation of wet-exploded wheat straw hydrolysate was investigated in a continuous immobilized reactor system. The experiments were carried out in a lab-scale fluidized bed reactor (FBR) at 70C. Undetoxified wheat straw hydrolysate was used (3-12% dry matter), corresponding...... to sugar mixtures of glucose and xylose ranging from 12 to 41 g/l. The organism, thermophilic anaerobic bacterium Thermoanaerobacter BG1L1, exhibited significant resistance to high levels of acetic acid (up to 10 g/l) and other metabolic inhibitors present in the hydrolysate. Although the hydrolysate...

Mesure et retroaction sur un qubit multi-niveaux en electrodynamique quantique en circuit non lineair

Science.gov (United States)

Boissonneault, Maxime

L'electrodynamique quantique en circuit est une architecture prometteuse pour le calcul quantique ainsi que pour etudier l'optique quantique. Dans cette architecture, on couple un ou plusieurs qubits supraconducteurs jouant le role d'atomes a un ou plusieurs resonateurs jouant le role de cavites optiques. Dans cette these, j'etudie l'interaction entre un seul qubit supraconducteur et un seul resonateur, en permettant cependant au qubit d'avoir plus de deux niveaux et au resonateur d'avoir une non-linearite Kerr. Je m'interesse particulierement a la lecture de l'etat du qubit et a son amelioration, a la retroaction du processus de mesure sur le qubit de meme qu'a l'etude des proprietes quantiques du resonateur a l'aide du qubit. J'utilise pour ce faire un modele analytique reduit que je developpe a partir de la description complete du systeme en utilisant principalement des transfprmations unitaires et une elimination adiabatique. J'utilise aussi une librairie de calcul numerique maison permettant de simuler efficacement l'evolution du systeme complet. Je compare les predictions du modele analytique reduit et les resultats de simulations numeriques a des resultats experimentaux obtenus par l'equipe de quantronique du CEASaclay. Ces resultats sont ceux d'une spectroscopie d'un qubit supraconducteur couple a un resonateur non lineaire excite. Dans un regime de faible puissance de spectroscopie le modele reduit predit correctement la position et la largeur de la raie. La position de la raie subit les decalages de Lamb et de Stark, et sa largeur est dominee par un dephasage induit par le processus de mesure. Je montre que, pour les parametres typiques de l'electrodynamique quantique en circuit, un accord quantitatif requiert un modele en reponse non lineaire du champ intra-resonateur, tel que celui developpe. Dans un regime de forte puissance de spectroscopie, des bandes laterales apparaissent et sont causees par les fluctuations quantiques du champ electromagnetique

Utilization of newly developed immobilized enzyme reactors for preparation and study of immunoglobulin G fragments

Czech Academy of Sciences Publication Activity Database

Korecká, L.; Bílková, Z.; Holčapek, M.; Královský, J.; Beneš, Milan J.; Lenfeld, Jiří; Minc, N.; Cecal, R.; Viovy, J.-L.; Przybylski, M.

2004-01-01

Roč. 808, č. 1 (2004), s. 15-24 ISSN 1570-0232. [International Symposium on Polymer Design for BioSeparation and Nanobiotechnology /8./. Compiegne, 27.11.2003-29.11.2003] Grant - others:GA ČR(CZ) GA203/02/0023 Program:GA Institutional research plan: CEZ:AV0Z4050913 Keywords : immobilized enzyme reactors * immunoglobulin G Subject RIV: CE - Biochemistry Impact factor: 2.176, year: 2004

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Apparatus for examination of irradiated fuel elements of industrial reactors at Marcoule; Appareillage d'examen des elements combustibles des piles industrielles de Marcoule

Energy Technology Data Exchange (ETDEWEB)

Pesenti, P; Wallet, Ph [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

The authors describe a viewing and measurement cell for the slugs of Marcoule industrial reactors. This cell allows visual inspection, and photography of slugs. Length measurements are also made possible by horizontal motion of the slug both in translation and rotation. (author) [French] Les auteurs decrivent une cellule d'observation et de mesure des elements combustibles des piles industrielles de Marcoule. La cellule permet l'examen a vue, la photographie, la radioscopie et la radiographie des elements combustibles. Elle permet en outre la mesure de longueurs sur ces elements, ces derniers pouvant etre deplaces horizontalement en translation, et en rotation. (auteur)

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

Two further years of operation of the reactor G1 (july 1958 - july 1960)

International Nuclear Information System (INIS)

Mathot, P.; Bauzit, J.; Cante, R.; Hebrard, L.

1960-01-01

The aim of the present report is to present certain observations and to give the results obtained during the period from july the 1 st 1958 to july the 1 st 1960. The main operations carried out during this period were, chronologically: - From july the 5 th to october the 18 th 1958: preparation and execution of the first annealing of the graphite. - From dec. the 15 th 1958 to july the 15 th 1959: a discharging campaign which resulted in the complete renewal of the fuel elements. During the monthly stoppages of this campaign, it was possible to make certain observations concerning the packing of the graphite, while at the same time measurements of the temperature of the element cans were made at an increased number of points. - From september the 25 th 1959 to december the 9 th 1959: preparation and execution of the second annealing. At the end of the annealing, the thorium lattice was modified and extra thermocouples were installed for measuring the temperature of the body of the graphite. An apparatus was built for measuring the radial flux. - From december the 9 th 1959 to july 1960: a continuous operation campaign, with a minimum of stoppages. The experimental results are re-assembled, independently of their chronological order, under three main headings which describe the reactors history: - continuous operation, - discharges, - annealing of the reactor. (author) [fr

La mesure comme représentation de l’objet

Directory of Open Access Journals (Sweden)

Danielle Laberge

2011-04-01

Full Text Available Adoptant une perspective d’interpénétration des méthodes et nous centrant sur le rôle des actes méthodiques dans la production des connaissances, nous procédons à l’examen de la place et du rôle de l’interprétation dans le processus de la mesure. Partant de la définition d’Abraham Kaplan selon laquelle « la mesure est l’assignation de nombres à des objets, des événements ou des situations à partir d’un système de règles définissant des propriétés pouvant être quantifiées » (Kaplan, 1964, nous considérons la mesure comme une activité méthodique de recherche constituée d’un ensemble d’actes interprétatifs distincts, mis en œuvre à des moments divers du processus de recherche. Nous montrons que la mesure est susceptible à la fois de réduire la complexité et de la restaurer. La mesure ne peut être limitée à sa dimension quantitative. Elle se construit dans l’interrelation permanente avec les autres actes de connaissance.Measurement process as object’s representation. Analysis and interpretationIn this article we examine the status and the role of interpretation in the measurement process, from the point of view of mixed methods. Starting with Abraham Kaplan’s definition of “measurement as the assignment of numbers to objects (or events or situations in accord with some rule defining properties that can be quantified”, we state that measurement is produced through a set of different interpretations at various moments during the research process. It cannot be seen only as a reduction of complexity and a quantification of reality since it is also a way of restoring complexity and quality. Measurement must be understood in relation with all the other knowledge operations.La medida como objeto de representaciones. Análisis e interpretaciónEn este artículo examínanos el estatuto y el papel de los métodos de medida desde el punto de vista de lo que constituye un acto metódico. A partir

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-09-15

direct reading of the permeability and stainless- steel components. The correlation between permeability and {Delta} ferrite content is explained. Measurements of the {Delta} ferrite percentage across welds in stainless-steel tubes and measurements of the {Delta} ferrite precipitations as a function of the plastic strain are discussed (hammer-forging of reactor fuel-elements). (author) [French] Les auteurs decrivent un instrument permettant de mesurer et d'enregistrer automatiquement le module de Young, le module de cisaillement et la capacite d'amortissement en fonction de la temperature et du temps. On mesure le module de Young en excitant des specimens de diverses dimensions a leur frequence propre. On mesure la capacite d'amortissement d'apres la libre decroissance de la vibration ou la largeur a mi-hauteur de la courbe de resonance. Le memoire donne des exemples de mesures de la guerison apres irradiation et apres deformation inelastique, ainsi que des exemples du degre de graphitisation. Les auteurs demontrent que l'on peut detecter des defauts et des variations de densite dans les banes de graphite. Ils expliquent, en outre, une methode d'etude de la fixation de pastilles d' UO{sub 2} sur des tubes en acier austenitique a parois minces. Us decrivent un four special pour l'etude du comportement elastique ou inelastique de specimens 'chauds ' a des temperatures variant entre 20 et 1000 Degree-Sign C. Les auteurs discutent le controle de la qualite de metaux non ferreux par mesure de ia conductivite electrique au moyen de courants de Foucault et decrivent un instrument permettant de mesurer sans aucun contact la conductivite electrique de metaux non ferreux. Ils expliquent la correlation entre la conductivite electrique et l'allongement sous l'effet des contraintes dans le cas de metaux et d'alliages non ferreux. Ils s'attachent particulierement a la mesure d'echantillons de petites dimensions. Ils decrivent un dispositif pour la mesure directe a distance dans la

An overview of the RECH-1 reactor conversion

International Nuclear Information System (INIS)

Klein, J.; Medel, J.; Daie, J.; Torres, H.

2000-01-01

The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

Energy Technology Data Exchange (ETDEWEB)

Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

1999-08-01

The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

International Nuclear Information System (INIS)

Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

1999-01-01

The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Training courses at VR-1 reactor

International Nuclear Information System (INIS)

Sklenka, L.; Kropik, M.

2006-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-03-15

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

Le temps mesurable, réversible, insaisissable ?

CERN Document Server

Fink, Martin; Leduc , Michèle

2016-01-01

Depuis l'Antiquité, la nature du temps a fasciné nombre de grands penseurs. Cet ouvrage expose ce que la physique est capable de dire aujourd'hui sur le sujet. La mesure du temps, ou plus exactement celle d'une durée, se fait grâce à des horloges atomiques dont l'exactitude peut atteindre une seconde sur plusieurs milliards d'années. Nous décrivons la façon dont s'effectue le transfert du temps qui permet la synchronisation d'horloges en différents points de la Terre ou de l'espace au milliardième de seconde près, ou même mieux. Les relativités, restreinte et générale, ont bouleversé notre conception du temps et ont un impact considérable sur certains problèmes de la vie quotidienne comme l'utilisation du GPS. On abandonne l'idée d'un temps absolu, le temps devient multiple et insaisissable, et peut-être même une illusion. Enfin la flèche du temps, ou l'irréversibilité, implique que les phénomènes physiques se déroulent toujours dans un sens déterminé, en relation avec la croissan...

Development of the multipurpose reactor G.A. Siwabessy (MPR-GAS)

International Nuclear Information System (INIS)

Handoyo, Demon; Deswandri; Mulyanto, Dwijo; Kusmono, Slamet

1998-01-01

To analyze the reliability of its components, the development of the Multi Purposes Reactor G.A. Siwabessy (MPR-GAS) data base has been done. The analysis was done base on the components operating experiences and the data collection from the operation data have been collected from February 1997 to January 1998. Only 22 components could be analyzed since these components operation cycles were shows very complete. The operation data provide values, such as; the operation time of components, number of failure, number of demand and number of failure on demand. The reliability parameters of the components were calculated using these values. In the calculation, the components were classified by the operating components, which gave a operation failure rate and the standby components, which gave a failure probability on demand. On calculating these parameters, not only the 1997/1998 data were used but also the past data (since September 1990), because the reliability are a cumulative value. Number of operation failure rate of components in 1997∼1998 are FAK=1,08 10 - 4 hr - 1, JE=2,53 10 - 4 hr - 1, KBE01=2,28 10 - 5 hr - 1, KBE02=3,72 10 - 5 hr - 1, PA-AP=3,01 10 - 4 hr - 1, PA A H=4,66 10 - 4 hr - 1

Current status of the Thai Research Reactor (TRR-1/M1)

International Nuclear Information System (INIS)

Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

2006-01-01

The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

Measurement of the local void fraction at high pressures in a heating channel; Mesure du taux de vide a haute pression dans un canal chauffant

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

Void fraction measurements were made in two phase flow boiling systems at high pressures in a uniformly heated, rectangular channel with a high aspect ratio. The local void fraction values were calculated from measurements of the absorption of a thin collimated X-ray beam (2 mm x 0.05 mm). The mean void fraction in a horizontal section results from integration of the local values across the section. At a fixed measuring station the quality and- void fraction were varied by changing the heat flux, flow rate and pressure systematically. Two channels were used differing in length and thickness (150.8 cm x 5.3 cm x 0.2 cm and the significant features of this study are: -1) The void fraction measurements are among the first obtained at such high pressure (80 to 140 kg/cm{sup 2}); -2) In the experimental region under consideration the measurements are systematic and numerous enough to allow accurate interpolations: mass velocity from 50 to 220 g/cm{sup 2}.s, heat flux from 40 to 170 W/cm{sup 2} and calculated steam quality from -0.2 to 0.2; -3) Many tests were performed under local boiling conditions with the mean temperature of the fluid below the saturation temperature; and -4) These results were compared to the predictions of certain models presented in the literature and simple empirical formulae were developed to fit the experimental results. (author) [French] Des mesures de taux de vide ont ete effectuees sur un ecoulement eau-vapeur a haute pression dans un canal vertical, de section rectangulaire tres allongee et chauffe a flux uniforme. Les valeurs du taux de vide local sont obtenues a partir des mesures de l'absorption d'un faisceau de rayons X finement collimate (2 mm x 0,05 mm). La valeur du taux de vide moyen dans une section droite s'en deduit par integration. Cette section droite ou sont realisees les mesures est fixe et, a pression, debit et flux donnes, les variations du titre et du taux de vide sont obtenues par variations de l'enthalpie d'entree. Deux

Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Gestion génétique des populations naturelles de poissons

Directory of Open Access Journals (Sweden)

1989-07-01

Il propose un ertain nombre de mesures simples permettant de conserver pendant un période relativement longue (20 à 30 ans des populations produisants des sujets de repeuplement "génétiquement proches" des individus sauvages de la population d'origine.

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-12-15

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10{sup -3}) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

International Nuclear Information System (INIS)

2013-01-01

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10 -3 ) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the Proceedings of

Measurement of the dead time of a G.M. counter and of the secondary emission of the cathode by the method of the delayed coincidences; Mesure du temps mort d'un compteur G.M. et de l'emission secondaire de la cathode par la methode des coincidences retardees

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

The dead time of a G.M counter is measured with the method of delayed coincidences. The pulses of the counter that supplies the circuit of coincidences, arrive there, on the one hand, directly, and in the other part, after a known and variable delay. This method permits besides, to study the parasitic impulses coming from the impact of the positive ions on the cathode of the meter. From the results relative to several counters working in various conditions are given. (author) [French] Le temps mort d'un compteur G.M. est mesure a l'aide d'un methode de coincidences retardees. Les impulsions du compteur qui alimentent le circuit de coincidences, y parviennent, d'une part, directement, et, dautre part, apres un retard connu et variable. Cette methode permet de plus, d'etudier les impulsions parasites provenant de l'impact des ions positifs sur la cathode du compteur. Des resultats relatifs a plusieurs compteurs fonctionnant dans des conditions diverses sont donnes. (auteur)

Mesure de la fraction d'evenements a quatre quarks dans les desintegrations multihadroniques du boson Z au LEP

Science.gov (United States)

Lefebvre, Eric

Cette thèse propose de tester la Chromodynamique Quantique (QCD) en effectuant une mesure précise d'une des trois constantes fondamentales du groupe de symétrie SU(3) utilisé pour décrire la physique des interactions fortes. Cette constante fondamentale, appelée TF, est reliée à certains états finaux spécifiques des désintégrations du Z0. Ces états apparaissent sous forme de perturbations du deuxième ordre en as et sont illustrés par des diagrammes de Feynman. À cet ordre, la chromodynamique prévoit deux types de diagrammes de Feynman distincts; le premier contient, un quark, un antiquark et deux gluons, et le second, deux quarks et deux antiquarks. La constante TF est directement proportionnelle à la fraction d'événements à deux quarks et deux antiquarks qui est l'objet de notre mesure. Notre mesure est fondée sur l'étude des événements à quatre partons dans l'état final. Ces quatre partons, en s'hadronisant, produisent quatre jets de particules qui peuvent être détectés expérimentalement et identifiés à l'aide d'algorithmes de reconstruction des jets. Des observables angulaires nous permettent de faire une discrimination parmi les états finaux de la désintégration du Z0, et ainsi déterminer la valeur de la fraction d'événements à deux quarks et deux antiquarks fq. Cette fraction peut s'exprimer par le rapport de la fraction observée expérimentalement fexq sur la fraction théorique fthq , R4q=fex qfthq. Afin d'améliorer la mesure de cette fraction et de diminuer le bruit causé par une contamination des événements d'ordres supérieurs, nous avons développe une méthode d'extrapolation. Cette méthode s'appuie sur la caractérisation de l'espace de phase des événements, exprimée par les paramètres yij. Les mesures de R4q que nous obtenons sont: R4q=2,27+0,29 -0,56 à l'aide d'une méthode conventionnelle avec la condition de sélection y34 > 0,020 > y45, et R4q=1,22+0,56 -0,71 à l'aide de notre méthode d

A neutronic feasibility study for LEU conversion of the SAFARI-1 reactor

International Nuclear Information System (INIS)

Pond, R.B.; Hanan, N.A.; Matos, J.E.; Ball, G.

2000-01-01

A neutronic feasibility study to convert the SAFARI-1 reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with NECSA. Comparisons were made of the reactor performance with the current 90% enriched HEU fuel type (UAl) and two 19.75% enriched LEU fuel types (U 3 Si 2 and U7Mo). The thermal fluxes with the LEU fuels were 3 - 9% lower than with the current HEU fuel. For the same fuel assembly design, a uranium density of approximately 4.5 g/cm 3 was required with U 3 Si 2 -Al fuel and a uranium density of about 4.6 g/cm 3 was required with U7Mo-Al fuel to match the 24.6-day cycle of the UAl-alloy fuel with 0.92 gU/cm 3 . The selection of a suitable LEU fuel and the decision to convert SAFARI-1 will be an economic matter that depends upon the fuel type, fuel assembly design, experiment performance and fuel cycle costs. (author)

Absolute measurement of {beta} emitters with a 4 {pi} counter; Mesure absolue des emetteurs {beta} au compteur 4 {pi}

Energy Technology Data Exchange (ETDEWEB)

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

1959-06-15

The object of this work is to investigate the conditions under which the activity of {beta}-emitting radionuclides may be measured with a maximum of precision, and as a result to study the relevant corrections. The various problems relating to activity measurements with a 4 {pi} counter have been examined successively: - comparison of 4 {pi}, GM and proportional counters; - study of the preparation of sources; - corrections on the counting of sources; - self-absorption; - correction for absorption. The precision obtained on these measurements varies from 1.2 to 3 per cent, with the result that the 4 {pi} counter can be considered a very satisfactory calibration instrument. (author) [French] Le but de ce travail est de rechercher les conditions permettant d'obtenir avec le maximum de precision, la mesure de l'activite des radionuclides se desintegrant par emission et par consequent d'etudier les corrections qui s'y rapportent. Nous avons examine successivement les differents problemes se rapportant aux mesures d'activite au compteur 4 {pi}: - Comparaison des compteurs 4 {pi}, GM et proportionnel; - etude de la preparation des sources; - corrections sur la numeration des sources; - auto-absorption; - correction d'absorption. La precision obtenue dans ces mesures, variant de 1,2 a 3 pour cent, on peut donc considerer le compteur 4 {pi} comme un instrument d'etalonnage tres satisfaisant. (auteur)

Absolute measurement of {beta} emitters with a 4 {pi} counter; Mesure absolue des emetteurs {beta} au compteur 4 {pi}

Energy Technology Data Exchange (ETDEWEB)

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

1959-06-15

The object of this work is to investigate the conditions under which the activity of {beta}-emitting radionuclides may be measured with a maximum of precision, and as a result to study the relevant corrections. The various problems relating to activity measurements with a 4 {pi} counter have been examined successively: - comparison of 4 {pi}, GM and proportional counters; - study of the preparation of sources; - corrections on the counting of sources; - self-absorption; - correction for absorption. The precision obtained on these measurements varies from 1.2 to 3 per cent, with the result that the 4 {pi} counter can be considered a very satisfactory calibration instrument. (author) [French] Le but de ce travail est de rechercher les conditions permettant d'obtenir avec le maximum de precision, la mesure de l'activite des radionuclides se desintegrant par emission et par consequent d'etudier les corrections qui s'y rapportent. Nous avons examine successivement les differents problemes se rapportant aux mesures d'activite au compteur 4 {pi}: - Comparaison des compteurs 4 {pi}, GM et proportionnel; - etude de la preparation des sources; - corrections sur la numeration des sources; - auto-absorption; - correction d'absorption. La precision obtenue dans ces mesures, variant de 1,2 a 3 pour cent, on peut donc considerer le compteur 4 {pi} comme un instrument d'etalonnage tres satisfaisant. (auteur)

Développement de la pupillométrie pour la mesure objective des émotions dans le contexte de la consommation alimentaire

OpenAIRE

Lemercier, Anaïs

2014-01-01

Les perceptions sensorielles et hédoniques résultent de processus complexes d’intégration, qui ne sont pas seulement rationnels, mais aussi fondés sur des sentiments, des émotions et des souvenirs. Afin d'appréhender au mieux le comportement du consommateur, il est devenu indispensable de mesurer les émotions afin de comprendre leur rôle fondamental dans la prise de décision. En science du consommateur, les émotions sont principalement mesurées par questionnaire. Malheureusement, cette mesure...

Sr and Pb isotopic composition of five USGS glasses (BHVO-2G, BIR-1G, BCR-2G, TB-1G, NKT-1G)

NARCIS (Netherlands)

Elburg, M.A.; Vroon, P.Z.; van der Wagt, R.A.C.A.; Tchalikian, A.

2005-01-01

Sr isotopic compositions and Rb/Sr ratios of three USGS glasses (BHVO-2G, BIR-1G, BCR-2G) are identical to those of the original USGS reference materials. NKT-1G and TB-1G give values of 0.70351 and 0.70558, respectively. Pb isotopic ratios were measured by the standard-sample bracketing technique

The Chernobyl reactor accident. Pt. 1 and 2

International Nuclear Information System (INIS)

1986-06-01

The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

Annual report on JEN-1 reactor

International Nuclear Information System (INIS)

Montes, J.

1972-01-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Measuring device for strong gamma-ray sources; Dispositif de mesure des fortes sources emettrices {gamma}

Energy Technology Data Exchange (ETDEWEB)

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

1956-07-01

We are given the description of a hollow argon-filled cylindrical ionisation chamber which is to be used to measure gamma-emitting sources. The instrument is currently used at the Measures Department in routine gauging of some radioelements. Sources are introduced into the central part of the chamber through a remote handling device. Measures are directly registered, it is not worth while removing the source from the container; a deviation of the source has little effect on the ionization current. The chamber was gauged to test such elements as: {sup 198}Au, {sup 60}Co, {sup 192}Ir, {sup 24}Na, {sup 137}Cs. Its measuring power approximately ranges from 100 micro-curies to 5 curies. (author) [French] On decrit une chambre d'ionisation cylindrique creuse, a remplissage d'argon, destinee a la mesure des sources emettrices {gamma}. Cet appareil est utilise couramment par la Section Mesures pour l'etalonnage de routine d'un certain nombre de radioelements. Les sources sont mises en place au centre de la chambre par un dispositif de manipulation a distance. La mesure est faite directement, sans qu'il soit necessaire d'extraire la source de son container; un decentrement de la source n'a en effet pas d'influence sensible sur le courant d'ionisation. Cette chambre d'ionisation a ete etalonnee pour divers radioelements: {sup 198}Au, {sup 60}Co, {sup 192}Ir, {sup 24}Na, {sup 137}Cs. La flamme d'activite mesurable s'etend de 100 microcuries a 5 curies, environ. (auteur)

Qualification process of dispersion fuels in the IEAR1 research reactor

International Nuclear Information System (INIS)

Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

2010-01-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

Nuclear reactor (1960); Reacteurs nucleaires (1960)

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

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

Automatic magnetic susceptibility measurements between 4 K and 1200 K; Mesure automatique des susceptibilites magnetiques de 4 K a 1200 K

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

We give a detailed description of a Faraday magnetic susceptibility balance which operates from 4 K to 1200 K. Some preliminary results on platinum and tantalum illustrate the precision and the sensitivity of the measurements. The apparatus has been designed for measurements on the plutonium compounds which present severe health hazards. (author) [French] Nous decrivons en detail un appareil permettant la mesure des susceptibilites magnetiques de 4 K a 1200 K par la methode de FARADAY. Quelques resultats preliminaires sur le platine et le titane montrent la precision et la sensibilite des mesures, L'appareil a ete adapte aux mesures sur les composes particulierement dangereux du plutonium. (auteur)

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP

Directory of Open Access Journals (Sweden)

Lim Mi Joung

2016-01-01

Full Text Available The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP-OPR (Optimized Power Reactor 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

Apparatus of irradiation of steel test pieces in the Marcoule pile G 1

International Nuclear Information System (INIS)

Marinot, R.; Wallet, Ph.

1960-01-01

Test pieces of steel were irradiated in the reactor G1 at Marcoule, in convectors replacing fuel elements, and in vertical channels in furnace-heated containers. The apparatus designed for this irradiation is described: containers, converter-rods, suspension fixtures and clamps, temperature measurement devices, lead castles and unloading set-ups. (author) [fr

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Mirror hybrid reactor optimization studies

International Nuclear Information System (INIS)

Bender, D.J.

1976-01-01

A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

Dynamic method for the measurement of Young'S modulus. Application to nuclear graphites; Methode de mesure dynamique du module d'Young. Application aux graphites nucleaires

Energy Technology Data Exchange (ETDEWEB)

Pattou, F; Trutt, J C

1963-07-01

A dynamic method has been developed for measuring Young's modulus and the rigidity modulus using the 'Forster Elastomat'. The principle consists in the determination of the resonance frequencies of graphite samples submitted to transverse, longitudinal, and torsional vibrations. The first two modes of vibration make it possible to calculate the elasticity modulus or the Young's modulus E, the third mode makes possible the calculation of the rigidity modulus G. The relationships from which the moduli E and G are measured are given. A systematic study has been made of graphite samples produced by extrusion or compression and submitted afterwards to one or several impregnations with pitch. For graphites made from the same coke by the same method, a linear relationship has been found for Young's modulus as a function of the apparent density. For the same apparent density, graphites made from different starting materials have generally different Young's moduli that bear a relationship to the crystalline characteristics of the material. The measurements of the rigidity modulus C made on different graphites also show the influence of crystallite orientation. (authors) [French] Une methode de mesure dynamique du module d'Young et du module de rigidite du graphite utilisant 'l'Elastomat Forster' a ete mise au point. Le principe consiste a determiner les frequences de resonance d'echantillons de graphite soumis a des vibrations transversales, longitudinales et de torsion. Les deux premiers modes de vibration permettent de calculer le module d'elasticite ou module d'Young E, le troisieme mode de vibration permet de calculer le module de rigidite G. Apres avoir decrit la methode de mesure, on rappelle les relations qui permettent de calculer les modules E et G. L'etude systematique d'echantillons de graphite, fabriques par filage ou pressage et ayant subi eventuellement une ou plusieurs impregnations au brai a ete effectuee. Pour les graphites issus du meme coke et fabriques

Alignement général du CLIC: stratégie et progrès

CERN Document Server

Mainaud-Durand, H

2008-01-01

La faisabilité concernant le pré-alignement actif du CLIC sera démontrée si l?on peut prouver qu?il existe une référence et ses capteurs associés permettant l?alignement des composants à mieux que 3 microns (1?). Pour répondre à ce challenge, une méthode de mesure d?écarts à un fil tendu est proposée, basée sur 40 ans de pratique de cette technique au CERN. Quelques problèmes demeurent concernant cette méthode : la connaissance de la forme du fil tendu utilisé comme référence droite, la détermination du géoïde à la précision souhaitée et le développement de capteurs bas coût permettant des mesures sub-micrométriques. Des études ont été entreprises afin de lever les derniers points en suspens, pendant que cette solution est intégrée dans une proposition concernant l?alignement général du CLIC. Cela implique un grand nombre d?interactions au niveau du projet, dans des domaines aussi différents que le génie civil, l?intégration, la physique du faisceau, la métrologie des �...

Efficiency of the Shut-Down and Safety Equipment and the Kinetic Characteristics of the G2 and G3 Reactors; Efficacite des dispositifs de secours et de securite et caracteristiques cinetiques des piles G2 et G3; Ehffektivnost' sistem avarijnoj zashchity reaktorov G.2 i G.3 i kineticheskie kharakteristiki ehtikh sistem; Caracteristicas cineticas y eficacia de los dispositivos de auxilio y de seguridad de los reactores G2 y G3

Energy Technology Data Exchange (ETDEWEB)

Henri, C.; Plisson, J.; Teste duBailler, A. [Centre d' Etudes Nucleaires de Saclay (France)

1963-10-15

The experience gained in several years of operating the G2 and G3 reactors confirms that natural uranium-graphite-gas reactors are extremely safe. The built-in shut-down and safety mechanisms which minimize operational incidents such as lack of power from the mains, blower failure, lack of water etc., together with accidents such as cladding bursts, local overheating, loss of coolant etc. are described and their operation explained by means of diagrams. The main points examined are as follows: (a) power distribution and controlability during accident conditions; (b) distribution of emergency water; and (c) the safety chain. The performance of the installations and the successive improvements incorporated in them are mentioned. The built-in safety characteristics of the reactors are shown by means of an experimental study of their behaviour in transient operation. These studies make it possible to check the validity of the calculation model. The machine calculation programmes can subsequently be used to study the consequences of possible accidents. Special attention is given to the depressurization accident, taking into account the performance of the safety device installed. (author) [French] L'experience acquise'au cours de plusieurs annees d'exploitation des piles G2 et G3 permet de confirmer le haut degre de securite du fonctionnement des piles de la filiere uranium naturel-graphitegaz. Les installations fixes de secours et de securite permettant de pallier, d'une part aux incidents d'exploitation tels que manque d'alimentation du reseau de distribution, arret de soufflage, manque d'alimentation en eau, etc., d'autre part, a des accidents tels que rupture de gaine, echauffements locaux, perte de fluide caloporteur, etc., sont decrites et leur fonctionnement explicite au moyen de schemas de principe. On examine principalement (a) la distribution ''puissance'' et ''controle'' des installations secourues, (b) la distribution d'eau secourue, et (c) la chaine de

Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-02-28

The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

Compilation of reports of the Advisory Committee on Reactor Safeguards, 1957-1984. Volume 4. Generic Subjects A-G

International Nuclear Information System (INIS)

1985-04-01

This six-volume compilation contains over 1000 reports prepared by the Advisory Committee on Reactor Safeguards from September 1957 through December 1984. The reports are divided into two groups: Part 1: ACRS Reports on Project Reviews, and Part 2: ACRS Reports on Generic Subjects. Part 1 contains ACRS reports alphabetized by subject name and within project name by chronological order. Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order. This volume contains generic reports arranged alphabetically from A to G

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Karel, Matejka; Lubomir, Sklenka

2005-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

COMPARISON OF UASB AND FLUIDIZED-BED REACTORS FOR SULFATE REDUCTION

Directory of Open Access Journals (Sweden)

S. M. Bertolino

2015-03-01

Full Text Available Abstract Reactor hydrodynamics is important for sulfidogenesis because sulfate reduction bacteria (SRB do not granulate easily. In this work, the sulfate reduction performance of two continuous anaerobic bioreactors was investigated: (i an upflow anaerobic sludge blanket (UASB reactor and (ii a fluidized bed reactor (FBR. Organic loading, sulfate reduction, and COD removal were the main parameters monitored during lactate and glycerol degradation. The UASB reactor with biomass recirculation showed a specific sulfate reduction rate of 0.089±0.014 g.gSSV-1.d-1 (89% reduction, whereas values twice as high were achieved in the FBR treating either lactate (0.200±0.017 g.gSSV-1.d-1 or glycerol (0.178±0.010 g.gSSV-1.d-1. Sulfate reduction with pure glycerol produced a smaller residual COD (1700 mg.L-1 than that produced with lactate (2500 mg.L-1 at the same COD.sulfate-1 mass ratio. It was estimated that 50% of glycerol degradation was due to sulfate reduction and 50% to fermentation, which was supported by the presence of butyrate in the FBR effluent. The UASB reactor was unable to produce effluents with sulfate concentrations below 250 mg.L-1 due to poor mixing conditions, whereas the FBR consistently ensured residual sulfate concentrations below such a value.

Some particular aspects of control in nuclear power reactors; Quelques aspects particuliers du controle dans les piles atomiques de puissance

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

There are still many problems in the field of measurement and control of neutron flux. The present studies in connexion with high flux reactors contribute to the solution of these problems which concern specialists in reactor control. The present state of this investigation and the results of different studies carried out in France by the C A and the EDF are pointed out: A - In the nuclear instrumentation field, work is at present devoted to the technologies used to develop detectors and cables, which have to work at high temperature and in a high {gamma} background; fast electronic techniques are applied to fission counters to measure low neutron fluxes in a high {gamma} background (10 Rh). B - In the control and safety field, there is a real need for studies on the behaviour of reactors in the subcritical state. This increases the margin of security during restarts when poison effects must be overcome The perturbations due to control rod movements necessitate a new organisation of power level safety and control assemblies, in connexion with thermal or activation measurements. Two methods of fast start-up are described. They are related to the fission rate measurement as a function of time. This is done either continuously by a constant and high reactivity change, or step by step. The application of automatic techniques to detector motion seems to give the answer to control and safety in normal start-up. C - The scope of these studies covers the methods used for the control of E.D.F. 3, which are described. (authors) [French] La mesure et le controle du flux neutronique dans les piles de puissance posent encore de nombreux problemes. Les etudes actuellement entreprises dans le domaine des piles a haut flux, doivent apporter une contribution importante a la solution de ces problemes qui interessent les specialistes du controle des piles de puissance. On analyse l'etat actuel de ces etudes et on donne les resultats des differents travaux effectues en France, dans

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Caractérisation de la végétation ligneuse du bassin versant de la ...

African Journals Online (AJOL)

Le bassin versant de la Maggia est située dans la région de Tahoua. La présente étude vise à caractériser la végétation ligneuse du bassin versant sur le plan de sa composition floristique et de sa structure. Les données ont été collectées au moyen des relevés floristiques, de mesures de diamètres à 1,30 m pour les.

Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

International Nuclear Information System (INIS)

Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

2013-01-01

Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

Mesure de la masse volumique des particules des eaux résiduaires urbaines : protocole et évaluation des incertitudes

OpenAIRE

RUBAN, Gwenaël; LABORATOIRE CENTRAL DES PONTS ET CHAUSSEES - LCPC

2004-01-01

La connaissance de la masse volumique des particules des eaux résiduaires urbaines est intéressante notamment pour l'étude de leur traitement par décantation ainsi que des relations entre concentrations en matières en suspension et paramètres optiques (turbidimétrie).Une étude des différentes méthodes de mesure de la masse volumique appliquées aux eaux résiduaires a été effectuée par G. Chebbo (1992). Il en a conclu que le pycnomètre à gaz type Beckman est le seul appareil donnant satisfactio...

Developpement d'une methode calorimetrique de mesure des pertes ac pour des rubans supraconducteurs a haute temperature critique

Science.gov (United States)

Dolez, Patricia

Le travail de recherche effectue dans le cadre de ce projet de doctorat a permis la mise au point d'une methode de mesure des pertes ac destinee a l'etude des supraconducteurs a haute temperature critique. Pour le choix des principes de cette methode, nous nous sommes inspires de travaux anterieurs realises sur les supraconducteurs conventionnels, afin de proposer une alternative a la technique electrique, presentant lors du debut de cette these des problemes lies a la variation du resultat des mesures selon la position des contacts de tension sur la surface de l'echantillon, et de pouvoir mesurer les pertes ac dans des conditions simulant la realite des futures applications industrielles des rubans supraconducteurs: en particulier, cette methode utilise la technique calorimetrique, associee a une calibration simultanee et in situ. La validite de la methode a ete verifiee de maniere theorique et experimentale: d'une part, des mesures ont ete realisees sur des echantillons de Bi-2223 recouverts d'argent ou d'alliage d'argent-or et comparees avec les predictions theoriques donnees par Norris, nous indiquant la nature majoritairement hysteretique des pertes ac dans nos echantillons; d'autre part, une mesure electrique a ete realisee in situ dont les resultats correspondent parfaitement a ceux donnes par notre methode calorimetrique. Par ailleurs, nous avons compare la dependance en courant et en frequence des pertes ac d'un echantillon avant et apres qu'il ait ete endommage. Ces mesures semblent indiquer une relation entre la valeur du coefficient de la loi de puissance modelisant la dependance des pertes avec le courant, et les inhomogeneites longitudinales du courant critique induites par l'endommagement. De plus, la variation en frequence montre qu'au niveau des grosses fractures transverses creees par l'endommagement dans le coeur supraconducteur, le courant se partage localement de maniere a peu pres equivalente entre les quelques grains de matiere

Banque d’instruments de mesure en recherche : Une innovation au service des membres chercheurs en sciences infirmières

Directory of Open Access Journals (Sweden)

Sylvie Le May

2017-04-01

Full Text Available Résumé : Introduction : Face aux difficultés que rencontrent ses enseignants et étudiants à retrouver des instruments de mesure valides dans les bases de données, le Réseau de Recherche en Interventions en Sciences Infirmières du Québec (RRISIQ a récemment choisi de développer une banque d’instruments de mesure accessible et bien documentée utilisant le logiciel bibliographique Zotero. Cet article a pour but de décrire la Banque d’instruments du RRISIQ, d’en exposer les défis et ses perspectives de développement. Description : La Banque comprend plus de 1400 liens ou références à des instruments de mesure reliés aux interventions cliniques, à l’organisation des services infirmiers et à la formation infirmière. L’utilisateur a accès à des références bibliographiques d’articles scientifiques sur les instruments, en anglais et en français. En naviguant dans la Banque, il clique sur l'article de son choix, obtenant ainsi une description bibliographique complète, dont une adresse web lui permettant d’accéder en ligne au plein texte. Résultats : La Banque d’instruments Zotero nécessite un faible coût d’entretien technique pour effectuer des sauvegardes, résoudre les difficultés et gérer les demandes d'accès. Elle est appréciée par ses utilisateurs. Discussion : La Banque prendra de l’ampleur dans les années à venir et des démarches sont actuellement réalisées par l’équipe pour la publiciser davantage auprès de ses membres et de leurs étudiants. L’équipe envisage de la rendre disponible à d’autres équipes de recherche du Québec.

Mesure de la vitesse d'infiltration des eaux dans le sol : Cas des sols ...

African Journals Online (AJOL)

SARAH

30 avr. 2016 ... Pour ceux, il est nécessaire d'opter pour une agriculture moderne ... à une telle agriculture. Mots clés : mesure, vitesse, infiltration, sols, pollutions, eau, Niari, Congo .... une porosité tubulaire d'origine biologique. Toutes ces.

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

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

The rehabilitation/upgrading of Philippine Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Renato, T Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

1998-10-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

The rehabilitation/upgrading of Philippine Research Reactor

International Nuclear Information System (INIS)

Renato T, Banaga

1998-01-01

The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E 1 -U-Z 1 -H 1.6 TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-06-15

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

Critical sizes and flux distributions in the shut down pile; Tailles critiques et cartes de flux a froid

Energy Technology Data Exchange (ETDEWEB)

Banchereau, A; Berthier, P; Genthon, J P; Gourdon, C; Lattes, R; Martelly, J; Mazancourt, R de; Portes, L; Sagot, M; Schmitt, A P; Tanguy, P; Teste du Bailler, A; Veyssiere, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

An important part of the experiments carried out on the reactor G1 during a period of shut-down has consisted in determinations of critical sizes, and measurements of flux distribution by irradiations of detectors. This report deals with the following points: 1- Critical sizes of the flat pile, the long pile and the uranium-thorium pile. 2- Flux charts of the same piles, and study of an exponential experiment. 3- Determination of the slit effect. 4- Calculation of the anisotropy of the lattice. 5- Description of the experimental apparatus of the irradiation measurements. (author) [French] Une part importante des experiences a froid effectuees sur le reacteur G1 a consiste en des determinations de tailles critiques et des mesures de distributions de flux par irradiations de detecteurs. Le present rapport traite les points suivants: 1- Tailles critiques de la pile plate, de la pile longue, de la pile a uranium-thorium. 2 - Cartes de flux des memes piles et etude d'une experience exponentielle. 3 - Determination de l'effet de fente. 4 - Calcul de l'anisotropie du reseau. 5 - Description de l'appareillage experimental des mesures d'irradiations. (auteur)

Critical sizes and flux distributions in the shut down pile; Tailles critiques et cartes de flux a froid

Energy Technology Data Exchange (ETDEWEB)

Banchereau, A.; Berthier, P.; Genthon, J.P.; Gourdon, C.; Lattes, R.; Martelly, J.; Mazancourt, R. de; Portes, L.; Sagot, M.; Schmitt, A.P.; Tanguy, P.; Teste du Bailler, A.; Veyssiere, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

An important part of the experiments carried out on the reactor G1 during a period of shut-down has consisted in determinations of critical sizes, and measurements of flux distribution by irradiations of detectors. This report deals with the following points: 1- Critical sizes of the flat pile, the long pile and the uranium-thorium pile. 2- Flux charts of the same piles, and study of an exponential experiment. 3- Determination of the slit effect. 4- Calculation of the anisotropy of the lattice. 5- Description of the experimental apparatus of the irradiation measurements. (author) [French] Une part importante des experiences a froid effectuees sur le reacteur G1 a consiste en des determinations de tailles critiques et des mesures de distributions de flux par irradiations de detecteurs. Le present rapport traite les points suivants: 1- Tailles critiques de la pile plate, de la pile longue, de la pile a uranium-thorium. 2 - Cartes de flux des memes piles et etude d'une experience exponentielle. 3 - Determination de l'effet de fente. 4 - Calcul de l'anisotropie du reseau. 5 - Description de l'appareillage experimental des mesures d'irradiations. (auteur)

Interactions of RuO4(g) with different surfaces in nuclear reactor containments

International Nuclear Information System (INIS)

Holm, J.; Glaenneskog, H.; Ekberg, C.

2008-07-01

During a severe nuclear reactor accident with air ingress, ruthenium in the form of RuO4 can be released from the nuclear fuel. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. This work has investigated the distribution of RuO4 between an aqueous and gaseous phase in the temperature interval of 20-50 deg. C by on-line measurements with an experimental set-up made of glass. The experiments showed that RuO4 is almost immediately distributed in the aqueous phase after its introduction in the set-up in the entire temperature interval. However, the deposition of ruthenium on the glass surfaces in the system was significant. The speciation of the ruthenium on the glass surfaces was studied by SEM-EDX and ESCA and was determined to be the expected RuO2. Experiments of interactions between gaseous ruthenium tetroxide and the metals aluminium, copper and zinc have been investigated. The metals were treated by RuO4 (g) at room temperature and analyzed with ESCA, SEM and XRD. The analyses show that the black ruthenium deposits on the metal surfaces were RuO2, i.e. the RuO4 (g) has been transformed on the metal surfaces to RuO2(s). The analyses showed also that there was a significant deposition of ruthenium tetroxide especially on the copper and zinc samples. Aluminium has a lower ability to deposit gaseous ruthenium tetroxide than the other metals. The conclusion that can be made from the results is that surfaces in nuclear reactor containments will likely reduce the source term in the case of a severe accident in a nuclear power plant. (au)

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Slug-Burst Detection in the G3 Reactor; La detection de rupture de gaine au reacteur G3; Obnaruzhenie razryva obolochki v reaktore G3; Deteccion de fallas del revestimiento en el reactor G3

Energy Technology Data Exchange (ETDEWEB)

Plisson, J. [Centre d' Etudes Nucleaires de Marcoule (France)

1963-10-15

The author explains the principles underlying slug-burst detection and describes the construction of the apparatus concerned. The main features are a) fully automatic operation, b) centralization of data in the control room and c) measurement by electrostatic collection on a turntable. (author) [French] Dans ce memoire, l'auteur expose les principes sur lesquels est fondee la detection de rupture de gaines et il decrit la realisation des installations. Les caracteristiques principales sont a) l'automatisme integral, b) la centralisation des informations dans la salle de commande et c) mesure par collection electrostatique sur plaque tournante. (author) [Spanish] El autor expone los principios en que se basa la deteccion de las fallas en los revestimientos de los elementos combustibles y describe las caracteristicas principales de la instalacion, que son: a) automatizacion integral, b) centralizacion de las informaciones en la sala de mandos, y c) medicion por recoleccion electrostatica sobre una placa giratoria. (author) [Russian] Izlagayutsya printsipy, na kotorykh osnovano obnaruzhenie razryva obolochki, opisyvaetsya konstruirovanie ustanovok. Osnovnye kharakteristiki takovy: a) integral'nyj avtomatizm, b) tsentralizatsiya informatsii v komandnom zale i c) izmerenie putem ehlektrostaticheskogo sobiraniya na povorachivayushchejsya plastinke. (author)

Effect of increase in salinity on ANAMMOX-UASB reactor stability.

Science.gov (United States)

Xing, Hui; Wang, Han; Fang, Fang; Li, Kai; Liu, Lianwei; Chen, Youpeng; Guo, Jinsong

2017-05-01

The effect of salinity on the anaerobic ammonium oxidation (ANAMMOX) process in a UASB reactor was investigated by analysing ammonium, nitrite, nitrate and TN concentrations, and TN removal efficiency. Extracellular polymeric substances (EPSs) and specific ANAMMOX activity (SAA) were evaluated. Results showed the effluent deteriorated after salinity was increased from 8 to 13 g/L and from 13 to 18 g/L, and TN removal efficiency decreased from 80% to 30% and 80% to 50%, respectively. However, ANAMMOX performance recovered and TN removal efficiency increased to 80% after 40 days when the influent concentrations of [Formula: see text] and [Formula: see text] were 200 mg/L and salinity levels were at 13 and 18 g/L, respectively. The amount of EPSs decreased from 58.9 to 37.1 mg/g volatile suspended solids (VSS) when the reactor was shocked by salinity of 13 g/L, and then increased to 57.2 mg/g VSS when the reactor recovered and ran stably at 13 g/L. The amount of EPSs decreased from 57.2 to 49.1 mg/g VSS when the reactor was shocked by salinity of 18 g/L, and then increased to 60.7 mg/g VSS when the reactor recovered and ran stably at 18 g/L. The amount of EPS and the amounts of polysaccharide, protein and humus showed no evident difference when the reactor recovered from different levels of salinity shocks. Batch tests showed salinity shock load from 8 to 38 g/L inhibited the SAA. However, when the reactor recovered from salinity shocks, SAA was higher compared to that when the reactor was subjected to the same level of salinity shock.

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Shutdown channels and fitted interlocks in atomic reactors

International Nuclear Information System (INIS)

Furet, J.; Landauer, C.

1968-01-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

G.M. counter and pre-determined dead time; Compteur G.M. et temps mort impose

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

This paper is divided into two main parts. - The first section recalls the principle on which a G.M. counter works, and examines the factors which lead to inaccuracies in counting. The concept of dead time, although simple risen associated with the counter alone, becomes complicated as soon as an electronic dead time is introduced to meet the demands of a measurement or an experiment. The resulting dead time, due to the coexistence of these dead times created by a single motivating factor, shows up as a function of certain laws of probability. From the analysis of the various cases of possible combinations, the conditions which must be fulfilled by a system with pre-determined dead time may be determined. This leads to a method for measuring the dead time of a G.M. counter, and the possibility of studying the latter under the utilisation conditions foreseen. - In the second part the principle, construction and characteristics of two systems with pre-determined dead time are discussed. To conclude, a comparison of several experimental results justifies an extension of the possibilities of a G.M. counter used in conjunction with such a system. (author) [French] Deux parties essentielles scindent cet expose. - La premiere partie rappelle le principe de fonctionnement d'un compteur G.M. et examine les facteurs d'imprecisions affectant les comptages. La notion de temps mort, simple quand elle est associee au compteur seul, se complique des qu'intervient un temps mort electronique introduit pour les besoins d'une mesure ou d'une experience. Le temps mort resultant, du a la coexistence de ces temps morts engendres par une meme cause, se manifeste en fonction de certaines lois de probabilites. L'analyse des differents cas de combinaisons possibles permet de preciser les imperatifs auxquels doit repondre un systeme a temps mort impose. Il en decoule une methode de mesure du temps mort d'un compteur G.M. et la possibilite d'etudier celui-ci dans les conditions d

Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

International Nuclear Information System (INIS)

Wheeler, J.K.; Pallotta, A.S.

1992-01-01

The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

Measurement of the strange quark content of nucleon: G{sup 0} experiment; Mesure du contenu etrange du nucleon: experience G{sup 0}

Energy Technology Data Exchange (ETDEWEB)

Batigne, G

2003-12-01

The G{sup 0} project is a parity violation experiment dedicated to the measurement of the proton weak and axial form factors by means of electron-proton scattering. Combining these weak form factors with the electromagnetic ones makes possible the extraction of the contribution of strange quarks to the charge and magnetization distribution in the nucleon. This thesis presents the strategy used for the G{sup 0} experiment, the different subsystems and the first results from its engineering run. The counting rate asymmetries, at the order of 10-5, are measured over a large range in transferred momentum (Q{sup 2} = 0.1 to 1 (GeV/c){sup 2}) with expected precision at the level of 10{sup -7}. A deadtime correction program has been developed which allows to correct 90% of the counting losses and to reduce associated false asymmetries at the level of 10-8. A method has been defined to extract the measured values of Q{sup 2} with a precision of 1%. The first preliminary results of G{sup 0} on parity violation asymmetries are also shown. (author)

The reactor Cabri; La pile cabri

Energy Technology Data Exchange (ETDEWEB)

Ailloud, J; Millot, J P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m{sup 3}/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

Mesure de la teneur en eau en continu durant le séchage du foin en balles

OpenAIRE

Cormier, Étienne

2008-01-01

Une mesure en continu et précise de la teneur en eau (TEE) permettrait d'optimiser le séchage du foin dans un séchoir commercial à grande échelle. Pour mesurer la précision des lectures dans ces conditions, un capteur électronique relié à 16 sondes a été utilisé pour estimer la TEE dans un séchoir expérimental. Deux sondes et un thermocouple étaient insérés dans huit couches de foin superposées, de 135 mm d'épaisseur chacune. Les TEE estimées par les sondes ont été comparées à des TEE exactes...

Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

International Nuclear Information System (INIS)

Rokhmadi

2007-01-01

Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

Application of a proportional counter to some particular cases of {alpha} measurements; Applications d'un compteur proportionnel a quelques cas particuliers de mesures {alpha}

Energy Technology Data Exchange (ETDEWEB)

Ferret, J; Gasc, M T; Le Du, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

I - A measurement method based on {alpha} {gamma} coincidences is used for activity measurements on {sup 241}Am in the presence of a strong {alpha} activity due to Pu. The sensitivity of the apparatus makes it possible to detect: less than 10{sup -7} {mu}C {sup 241}Am in the presence of a Pu activities several hundred times greater. II - The same equipment has been used for {alpha} measurements on {alpha}-emitting sources in the presence of very strong {beta} and {gamma} activities. In particular results are given for {alpha}-activity measurements on powder samples. It is even possible in these conditions, to detect or measure {sup 241}Am. III - The equipment makes possible also: - the absolute calibration of a pure {sup 241}Am source - {sup 237}Np measurements - simultaneous measurements of {alpha}, {beta} and {gamma} activities in solid samples in various forms. IV - The assembly includes a small-size proportional counter operating in conjunction with a {gamma} probe, together with auxiliary electronic equipment (stabilized high voltages, amplifiers, a coincidence unit, a sealer). (authors) [French] I - Une methode de mesures par coincidences {alpha} {gamma} est utilisee pour des mesures d'activite de {sup 241}Am en presence d'activite {alpha} (due au Pu) importante. -La sensibilite de l'appareillage permet de deceler: moins de 10{sup -7} {mu}C {sup 241}Am dans les activites {alpha} Pu plusieurs centaines de fois plus importantes. II - Le meme appareillage a ete utilise pour des mesures {alpha} sur des sources emettrices {alpha} en presence d'activites {beta} et {gamma} tres importantes. En particulier des mesures d'activite {alpha} sur des echantillons de poudre sont exposees. Il est meme possible, dans ces conditions, de deceler ou mesurer {sup 241}Am. III - L'ensemble permet egalement: - l'etalonnage absolu d'une source de {sup 241}Am pur - des mesures de {sup 237}Np - des mesures simultanees d'activites {alpha}, {beta} et {gamma} dans des echantillons solides

Control panel for radiation around reactors (1963); Tableaux de controle des radiations aupres des piles (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

This report outlines the general philosophy of radiation control in French reactors and their annexes. The supervision is carried out continuously from a central control panel on which appear all the measurements made and the alarm signals. The equipment is described; one item makes it possible to measure simultaneously the radioactive dusts and gases. The specifications of the alarm system, which is considered to be the most important are given. Finally a new measuring technique is proposed which makes it possible to reduce considerably the cost of radiation control while at the same time providing the results in a form in which they can be easily treated, in particular in the case of the calculation of total doses. (authors) [French] Ce rapport definit la philosophie generale du controle des radiations dans les piles francaises et dans leurs annexes. La surveillance se fait d'une maniere continue a partir d'un tableau de controle centralise ou sont reportees toutes les mesures et les signalisations d'alarme. On decrit les appareils utilises, dont un permet la mesure simultanee des poussieres et gaz radioactifs, et on definit les specifications de la fonction alarme qui est consideree comme la plus importante. Enfin on propose une nouvelle technique de mesure qui permettrait de reduire considerablement le cout du controle des radiations tout en fournissant des resultats plus facilement exploitables, en particulier pour le calcul des doses integrees. (auteurs)

Burnup measurements at the RECH-1 research reactor

International Nuclear Information System (INIS)

Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

2002-01-01

The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

High temperature elastic constant measurements: application to plutonium; Mesure des constantes elastiques a haute temperature application au plutonium

Energy Technology Data Exchange (ETDEWEB)

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

1969-03-01

We present an apparatus with which we have measured the Young's modulus and the Poisson's ratio of several compounds from the resonance frequency of cylinders in the temperature range 0 deg. C-700 deg. C. We especially studied the elastic constants of plutonium and measured for the first time to our knowledge the Young's modulus of Pu{sub {delta}} and Pu{sub {epsilon}}. E{sub {delta}} 360 deg. C = 1.6 10{sup 11} dy/cm{sup 2}; E{sub {epsilon}} 490 deg. C = 1.1 10{sup 11} dy/cm{sup 2}, {sigma}{sub {epsilon}} = 0.25 {+-} 0.03 Using our results, we have calculated the compressibility, the Debye temperature, the Grueneisen constant and the electronic specific heat of Pu{sub {epsilon}}. (author) [French] Nous decrivons un appareil qui permet de mesurer les constantes elastiques (module de Young et module de Poisson) jusqu'a 700 deg. C a partir des frequences de resonance de barreaux cylindriques. Nous avons plus specialement etudie le plutonium et determine pour la premiere fois a notre connaissance le module de Young des phases {delta} et {epsilon}: E{sub {delta}} 360 deg. C = 1.6 10{sup 11} dy/cm{sup 2}; E{sub {epsilon}} 490 deg. C = 1.1 10{sup 11} dy/cm{sup 2}, {sigma}{sub {epsilon}} = 0.25 {+-} 0.03 Nos mesures nous ont permis de calculer la compressibilite, la temperature de Debye, la constante de Gruneisen et la chaleur specifique electronique de Pu{sub {epsilon}}. (auteur)

Training and research on the nuclear reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

1998-01-01

The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Analysis of Chemical Species Along the Process Stages of Demineralized Water Production at Reactor G.A. Siwabessy

International Nuclear Information System (INIS)

Nurul Huda; Setyono; Sumijanto; Diah E L; Ihsan, M.

2003-01-01

The tank water of multipurpose reactor G.A. Siwabessy is supplied from a water demineralization plant which works based on ion exchange processes. Controlling the quality of the water produced by this plant is one of many factor which effects the quality of the reactor tank water. This experiment resulted a characteristic pattern data of water and its chemical species content along process stages of demineralized water production at the reactor. The experiment results showed that the pH (degree of acidity), electric conductivity and dissolved cation (Ca 2+ , Mg 2+ ) lied at the permissible range. The value fluctuation of these variables showed a right pattern. It can be concluded that the water produced by this plant met the requirements to be used as primary cooling water of the reactor. However, the value of pH is still too low although it lied in the tolerance limit. Beside that, it isn't all of water impurities concentration can be predicted by the value of pH and conductivity. Therefore, the determination of water quality for the need of reactor tank water quiet to be done continually to keep the water condition in order to meet the quality required, and to evaluate and developed its production technology. (author)

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

High speed pulsed magnetic fields measurements, using the Faraday effect; Mesures de champs magnetiques pulses rapides a l'aide de l'effet Faraday

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-01

For these measures, the information used is the light polarization plane rotation induced by the magnetic field in a glass probe. This rotation is detected using a polarizer-analyzer couple. The detector is a photomultiplier used with high-current and pulsed light. In a distributed magnet (gap: 6 x 3 x 3 cm) magnetic fields to measure are 300 gauss, lasting 0.1 {mu}s, with rise times {<=} 35 ns, repetition rate: 1/s. An oscilloscope is used to view the magnetic field from the P.M. plate signal. The value of the field is computed from a previous static calibration. Magnetic fields from 50 to 2000 gauss (with the probe now used) can be measured to about 20 gauss {+-} 5 per cent, with a frequency range of 30 MHz. (author) [French] Pour faire de telles mesures, on utilise comme information la rotation du plan de polarisation de la lumiere provoquee par le champ magnetique dans une sonde en verre. On detecte cette rotation au moyen d'un polariseur et d'un analyseur, qui sont regles a 45 deg. pour conserver un phenomene lineaire. Le detecteur est un photomultiplicateur travaillant en fort courant en lumiere pulsee. Dans un aimant distribue d'entrefer 6 x 3 x 3 cm, on obtient des champs magnetiques a mesurer de 300 gauss, durant 0.1 {mu}s, avec des temps de montee {<=} 35 ns; au taux de 1 fois par seconde. L'observation du champ se fait sur oscilloscope a partir du signal de plaque du P.M. La valeur absolue du champ est obtenue au moyen d'un etalonnage statique prealable. On peut ainsi mesurer a 20 gauss et {+-} 5 pour cent pres environ des champs magnetiques de 50 a 2000 gauss (avec la sonde actuelle) et avec une bande passante de 30 MHz. (auteur)

Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

IEA-R1 reactor - Spent fuel management

International Nuclear Information System (INIS)

Mattos, J.R.L. De

1996-01-01

Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

A experimental system for the checking of the absorption of E.C.A.G. graphite; Empilement pour le controle du graphite E.C.A.G

Energy Technology Data Exchange (ETDEWEB)

Raievski, V; Vidal, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1958-07-01

A system is described for measuring the mean absorption cross section in thermal neutrons of graphite. This system consists of a graphite stack containing a Ra-Be source and a BF3 counter. A cavity in the stack receives the graphite to be studied or the graphite standard. By comparing the counting rates their absorption ratio can be deduced. The measurement is performed on graphite rods which have been machined before being placed in the pile. It provides the possibility of detecting over a batch of 1 ton of graphite, in a single measurement, a difference in absorption of 0.1 milli barn. (author) [French] On decrit un dispositif permettant de mesurer la section efficace moyenne d'absorption en neutrons thermiques du graphite. Ce dispositif est constitue par un empilement de graphite contenant une source de Ra-Be et un compteur a BF3. Une cavite menagee dans l'empilement peut recevoir le graphite a etudier ou le graphite etalon. Par comparaison des taux de comptage, on en deduit leur rapport d'absorption. La mesure est effectuee sur des barres de graphite usinees avant leur mise en place dans la pile. Elle permet de deceler sur un lot de graphite de 1 tonne, en une seule mesure, une difference d'absorption de 0,1 millibarn. (auteur)

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-02-15

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

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

The reactor Cabri; La pile cabri

Energy Technology Data Exchange (ETDEWEB)

Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m{sup 3}/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

2014-08-15

The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

International Nuclear Information System (INIS)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A.

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U 3 O 8 and U 3 Si 2 dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U 3 O 8 -Al and five containing U 3 Si 2 -Al), with densities of 3.2 gU/cm 3 and 4.8 gU/cm 3 respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

International Nuclear Information System (INIS)

Ardani

2011-01-01

Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

Energy Technology Data Exchange (ETDEWEB)

Palmer, E.R. [Westinghouse Savannah River Company, AIKEN, SC (United States)

1997-07-01

Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA.

Removal Site Evaluation Report to the C-Reactor Seepage Basins (904-066, -067 and -068G)

International Nuclear Information System (INIS)

Palmer, E.R.

1997-07-01

Removal Site Evaluation Reports are prepared in accordance with Section 300.410 of the National Contingency Plan (NCP) and Section X of the Federal Facility Agreement (FFA). The C-Reactor Seepage Basins (904-066G,-067G,-068G) are listed in Appendix C, Resource Conservation and Recovery Act (RCRA)/Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Units List, of the FFA. The purpose of this investigation is to report information concerning conditions at this unit sufficient to assess the threat (if any) posed to human health and the environment and to determine the need for additional CERCLA action. The scope of the investigation included a review of past survey and investigation data, the files, and a visit to the unit.Through this investigation unacceptable conditions of radioactive contaminant uptake in on-site vegetation were identified. This may have resulted in probable contaminant migration and become introduced into the local ecological food chain. As a result, the SRS will initiate a time critical removal action in accordance with Section 300.415 of the NCP and FFA Section XIV to remove, treat (if required), and dispose of contaminated vegetation from the C-Reactor Seepage Basins. Erosion in the affected areas will be managed by an approved erosion control plan. further remediation of this unit will be conducted in accordance with the FFA

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

Mesures en matière de taxation des produits du tabac en Afrique de ...

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

Mesures en matière de taxation des produits du tabac en Afrique de l'Ouest. En décembre 2007, le programme Recherche pour la lutte mondiale contre le tabac (RMCT) a entrepris une initiative afin de comprendre les facteurs cruciaux qui peuvent déterminer le succès de la lutte antitabac en Afrique subsaharienne.

Heavy water reactors physics; Physique des reacteurs a eau lourde

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

Lightning protection system analysis at Multipurpose Reactor G A. Siwabessy building

International Nuclear Information System (INIS)

Teguh-Sulistyo

2003-01-01

Analysis to the part of lightning protection system at Multi Purpose Reactor GA Siwabessy (RSG-GAS) have been done. Observation examined the damage of some part of the earthing system caused by human error of chemically system. The analysis performed some assumptions and simulations to the points of lightning stroke. From this analysis obtained that the reactor building do not have vertical finial which can protect effectively to the whole reactor building and auxiliary building. Installing some new finials at some places are needed to protect building therefore the reactor building and auxiliary building well safe from lighting stroke

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

Operation and maintenance of 1MW PUSPATI TRIGA reactor

International Nuclear Information System (INIS)

Adnan Bokhari; Mohammad Suhaimi Kassim

2006-01-01

The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Perspectives on reactor safety. Revision 1

International Nuclear Information System (INIS)

Haskin, F.E.; Hodge, S.A.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Perspectives on reactor safety. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Utilization of titanium sponge in H. T. G. R

Energy Technology Data Exchange (ETDEWEB)

Tone, H [Japan Atomic Energy Research Inst., Oarai, Ibaraki. Oarai Research Establishment

1977-10-01

The high temperature, gas-cooled reactor (H.T.G.R.) uses helium as a coolant and graphite as both the moderator and the fuel tube material. At first sight, there should not be any problem concerning the compatibility of these materials in the H.T.G.R. core region where temperature exceeds 700/sup 0/C, however, it is possible that the graphite core and other structural materials are oxidized by traces of impurities in the coolant. In large-power H.T.G.R., water inleakage from both heat exchangers and coolant circulation pumps will probably be the major source of impurity which will react with the graphite-producing H/sub 2/, CO and CO/sub 2/. In the near future, the nuclear heat of H.T.G.R. will be used as a major heat source for steel production and the chemical industry. For these purposes, it will be necessary to construct a reactor using a helium coolant of greater than 1000/sup 0/C. Therefore, not only the development of refractory metals as structural materials but also an effective helium coolant purification system are the keys for H.T.G.R. construction. Recently, in the helium coolant purification system of H.T.G. Reactors, which have been developed in the several nations advanced in atomic reactors, titanium sponge is used very frequently to remove hydrogen gas as an impurity in helium coolant. Titanium sponge can absorb very large quantities of hydrogen and its absorption-capacity can be very easily controlled by controlling the temperature of the titanium sponge-since titanium hydride is formed by endothermic reaction. The titanium sponge trap is used also in OGL-1 (Oarai Gas Loop-1), helium coolant purification system for large scale irradiation apparatus which is used for nuclear fuels of H.T.G.R. This apparatus has been installed in the Japan Material Testing Reactor. In this report, the coolant purification system of H.T.G.R., OGL-1 and the experimental results of the titanium sponge trap are explained briefly.

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Advances in Reactor Physics, Mathematics and Computation. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-06-15

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

Kinetics of Ar+*(2G9/2) metastable ions and transport of argon ions in ICP reactor

NARCIS (Netherlands)

Sadeghi, N.; Derouard, J.; Grift, van de M.; Kroesen, G.M.W.; Hoog, de F.J.; Tachibana, K.; Watanabe, Y.

1997-01-01

The decay time of the argon Ar~~(2G912) metastable ions was measured in the afterglow of a low pressure pulsed helicon reactor. From the argon pressure and electron density dependence of this decay time, rate coefficients for quenching of these ions by argon atoms and by plasma electrons have been

Reactor utilization, Part 1

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1981-01-01

The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

Mesure de la vitesse d'infiltration des eaux dans le sol : Cas des sols ...

African Journals Online (AJOL)

C'est donc une infiltration superficielle qui ne peut pas modifier la composition des eaux de l'aquifère. Ces sols sont alors favorables à une telle agriculture. Mots clés: mesure, vitesse, infiltration, sols, pollutions, eau, Niari, Congo. English Title: Measuring the speed of the water infiltration into the soil: case of the soil of the ...

Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

International Nuclear Information System (INIS)

Salahuddin, A.; Israr, M.; Hussain, M.

2002-01-01

Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Prévalence du syndrome métabolique et de ses facteurs de risque chez les enfants et les adolescents canadiens : Enquête canadienne sur les mesures de la santé, cycle 1 (2007-2009 et cycle 2 (2009-2011

Directory of Open Access Journals (Sweden)

M. MacPherson

2016-01-01

Full Text Available Introduction : Nous avons étudié la prévalence du syndrome métabolique (SMet et de ses facteurs de risque ainsi que l'influence du statut socioéconomique chez les enfants et les adolescents canadiens. Méthodologie : Nous avons inclus dans notre étude les 1228 répondants de l'Enquête canadienne sur les mesures de la santé, cycle 1 (2007-2009 et cycle 2 (2009-2011, âgés de 10 à 18 ans et ayant fourni un échantillon de sang à jeun. Nous avons utilisé les définitions consensuelles du SMet proposées par la Fédération internationale du diabète (FID pour les enfants et adolescents (10 à 15 ans et pour les adultes (16 ans et plus. Nous avons mesuré la prévalence du SMet et de ses facteurs de risque ainsi que les différences en fonction du statut socioéconomique au moyen de tests du x2. Résultats : La prévalence du SMet était de 2,1 %. Le tiers (37,7 % des répondants présentaient au moins un facteur de risque, les plus répandus étant l'obésité abdominale (21,6 %, un faible taux de C-HDL (19,1 % et un taux de triglycérides élevé (7,9 %. Cette combinaison d'obésité abdominale, de faible taux de C-HDL et de taux élevé de triglycérides correspondait à 61,5 % des cas de SMet. Les participants des ménages de la tranche supérieure de revenu et bénéficiant d'un niveau de scolarité élevé présentaient la plus faible prévalence d'un ou de plusieurs facteurs de risque du SMet, d'obésité abdominale et de faible taux de C-HDL. Conclusion :La prévalence du SMet (2,1 % s'est révélée inférieure à celle mesurée auparavant pour le Canada (3,5 % et les États-Unis (4,2 % à 9,2 %, sans doute en raison de l'application stricte des critères de la FID pour l'étude du SMet. Le tiers des enfants et des adolescents canadiens présentaient au moins un facteur de risque de SMet. Comme le risque de SMet augmente avec l'âge, ces estimations de la prévalence, couplées à une prévalence nationale de l'obésité d

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

The MAP, M/G1,G2/1 queue with preemptive priority

Directory of Open Access Journals (Sweden)

Bong Dae Choi

1997-01-01

Full Text Available We consider the MAP, M/G1,G2/1 queue with preemptive resume priority, where low priority customers arrive to the system according to a Markovian arrival process (MAP and high priority customers according to a Poisson process. The service time density function of low (respectively: high priority customers is g1(x (respectively: g2(x. We use the supplementary variable method with Extended Laplace Transforms to obtain the joint transform of the number of customers in each priority queue, as well as the remaining service time for the customer in service in the steady state. We also derive the probability generating function for the number of customers of low (respectively, high priority in the system just after the service completion epochs for customers of low (respectively, high priority.

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-01-01

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

Final Remediation Report for the K-Area Bingham Pump Outage Pit (643-1G); FINAL

International Nuclear Information System (INIS)

Morganstern, M.

2002-01-01

The K-Area Bingham Pump Outage Pit (K BPOP) Building Number 643-1G, is situated immediately south and outside the K-Reactor fence line and is approximately 400 feet in length and 60 feet in width. For the K BPOP operable unit, the Land Use Control (LUC) objectives are to prevent contact, removal, or excavation of buried waste in the area and to preclude residential use of the area

Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

1993-01-01

The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

Pseudo-harmonics method: an application to thermal reactors

International Nuclear Information System (INIS)

Silva, F.C. da; Rotenberg, S.; Thome Filho, Z.D.

1985-10-01

Several applications of the Pseudo-Harmonics method are presented, aiming to calculate the neutron flux and the perturbed eigenvalue of a nuclear reactor, like PWR, with three enrichment regions as Angra-1 reactor. In the reference reactor, perturbations of several types as global as local were simulated. The results were compared with those from the direct calculation. (E.G.) [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-02-15

Critical and exponential experiments in general produce overlapping information on reactor lattices. Over the past ten years the Savannah River Laboratory has been operating a heavy-water critical, the PDP, and an exponential, the SE, in parallel. This paper summarizes SRL experience to give results and recommendations as to the applicability of the two kinds of facilities in different experiments. Six types of experiments are considered below: (1) Buckling measurements in uniform isotropic lattices Here Savannah River has made extensive comparisons between single-region criticals, exponentials, substitution criticals, and PCTR type measurements. The only difficulties in the exponentials seem to lie in the radial-buckling determinations. If these can be made successfully, the exponentials can offer good competition to the criticals. Material requirements are greatest for the single-region criticals, roughly comparable for the substitution criticals and exponentials, and least for the PCTR measurements. (2) Anisotropic and void effects SRL experiments with the criticals and with critical-exponential comparisons are reviewed briefly here and at greater length in a companion paper. (3) Evaluation of control systems Adequately analysed exponential experiments appear to give good results for total-worth measurements. However, for adequate study of overall flux shaping, flux tilts, etc. a full-sized critical such as the PDP is required. (4) Temperature coefficients Exponential experiments provide an excellent method for determining the temperature coefficient of buckling for uniform lattice heating. A special facility, the PSE, at Savannah River permits such measurements up to temperatures of 215 Degree-Sign C. For non-uniform lattice heating criticals are generally preferred. (5) Mixed lattices Actual reactors rarely use the simple uniform lattices to which the exponentials basically apply. Critical experiments with mixed loadings are used at SRL both in measuring

Design and construction of a faraday cup for measuring small electron currents; Etudes et realisation d'une ''coupe de faraday'' pour les mesures de faibles courants electroniques

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

This paper describes the design of a device for measuring and integrating very small currents generated by the impact of a charged particle beam upon a Faraday cup. Part one considers the detector as such. The main component is a graphite bloc capable of stopping practically all the incident charges. Part two describes the associated electrode apparatus required to measure better than 10{sup -13} ampere with a precision- of 1 per cent: Integration of such weak currents over periods of several hours, in the presence of a strong background current, is also discussed. (author) [French] Ce rapport decrit l'etude et la realisation d'un ensemble permettant de mesurer et d'integrer sur des periodes de plusieurs heures des courants electroniques tres faibles provenant d'un faisceau de particules chargees, Dans la premiere partie du rapport nous etudierons le capteur proprement dit qui se compose essentiellement d'un bloc de graphite dont la forme et les dimensions sont telles, que la majeure partie des charges (positons et negatons de 60 MeV) est captee (1 pour mille reussissent a s'echapper). Dans la deuxieme partie nous decrivons l'appareillage associe au capteur capable de mesurer moins de 10{sup -13} ampere avec une precision de l'ordre du pour cent et d'integrer ce courant sur des periodes de temps variables, compte tenu de l'ambiance 'bruyante' (Accelerateur Lineaire) dans laquelle la mesure s'effectue. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-07-01

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

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Report on the activity of the RA reactor operation for the period from July 1 1961 - Sept. 30 1961; Tromesecni izvestaj za period od 1.VII do 30.IX 1961. g

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V [Laboratorija za eksploataciju reaktora RA, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-09-15

During the reporting period the reactor was permanently ready for operation and responding to the demands of the experimenters. The reactor was operating for 408.5 hours at power levels from 50 - 5000 kW, or 1985 MWh in total, burnup of the first batch of fuel was 22.55%. Reactor core was made of 56 fuel channels. Activities related to construction of new and improvement of the existing equipment was continued in order to enable safe operation and successful utilization of the RA reactor. Exchange of the electronic tubes was continued in order to increase the stability of the reactor control and reactor protection systems. About 65% of tubes planned to be exchanged this year was done. Cooperation with the CEN Saclay, France related to construction of experimental loops VISA-1 and VISA-2 was continued as well as cooperation with Poland concerned with exchange of experts. The problem of lack of properly trained staff was nor solved. [Serbo-Croat] U izvestajnom periodu reaktor je uvek bio spreman za rad i odgovarao je zahtevima eksperimentatora. Reaktor je radio 408,5 casova na snagama od 50-5000 kW, odnosno ukupno 1985 MWh, pri cemu je ukupan utrosak prve sarze goriva bio 22,55%. Jezgro reaktora bilo je sacinjeno od 56 tehnoloskih kanala. Radi bezbednijeg pogona i sto uspesnije eksploatacije reaktora RA nastavljeno je sa radovima na realizaciji novih i poboljsanju postojecih uredjaja. U cilju povecanja stabilnosti elektronskih instrumenta u sistemu upravljanja i zastiti reaktora nastavljeno je sa zamenom elektronskih cevi, do sada je zamenjeno oko 65% radova predvidjenih za ovu godinu. Nastavljena je saradnja sa nuklearnim centrom u Saclay, Francuska na izradi petlje VISA-1 i VISA-2 i saradnja sa Poljskom u razmeni strucnjaka. Problem nedostatka kadrova nije resen.

New human machine interface for VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

2002-01-01

The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

Sensitivity analysis of power excursion in RSG-GAS reactor due to reactivity insertion

International Nuclear Information System (INIS)

Pinem, Surian; Sembiring, Tagor Malem

2002-01-01

Reactor kinetics has a very important role in reactor operation safety and nuclear reactor control. One of the important aspects in reactor kinetics is power behavior as function of time due to chain reaction in the core. The calculation was performed using kinetic equation and feedback reactivity and evaluated using static power coefficient. Analysis was performed for oxide 250 g, silicide 250 g and silicide 300 g fuel elements with insertion of positive reactivity, negative reactivity and reactivity close to delay neutron fraction. The calculation of power excursion sensitivity showed that the insertion of 0,5 % Δk/k, in the fuel element of silicide 300 g is bigger 5 % than the one of oxide 250 g or silicide 250 g. If inserted by - 1,2 % Δk/k, there is no change among three fuel elements. Therefore, in kinetic point of view, it is showed there is no significant influence in the RSG-GAS reactor operation safety is the current core of oxide 250 g is converted to silicide 250 g or to silicide 300 g

Stability of an anaerobic single reactor filled with dolomitic limestone with increased organic load of sugarcane

Directory of Open Access Journals (Sweden)

Maria Magdalena Ribas Döll

2017-12-01

Full Text Available The anaerobic single-stage reactor was evaluated to treat vinasse and to evaluate its stability. This bench reactor was filled with dolomitic limestone with a horizontal plug flow to simulate a drainage channel. The experiment lasted 129 days while the reactor was submitted to different applied organic concentrations (chronologically applied: 3.0; 5.0; 12.0; 9.0 and 7.5 g L-1 as COD, chemical oxygen demand. COD removals were 50% and 9% with 3.0 and 7.5 g L-1, respectively. With 12.0 g L-1, reactor efficiency increased to 33%, with an abrupt drop to 3% on the 84th day. Therefore, in order to avoid reactor collapse, a remedial measure was necessary. The system remained in batch without feeding for 19 days (from the 85th to the 104th day with 9.0 g L-1. Afterwards, it was observed that the performance of the system tended to stabilize, reaching 47% with 7.5 g L-1 in the 118th day. At the end of the experiment, the potassium content of the wastewater decreased from 800 mg L-1 to 594 mg L-1 (on an average 25% and calcium and magnesium increased within the reactor liquor. The dissolution of the limestone inside the liquor reactor probably caused this result. After the treatment with limestone, the average pH value of the effluent increased from 4.9 to over 6.0 in all organic concentrations. It could be concluded that the reactor filled with dolomitic limestone in these operational conditions assured a low efficiency in COD removal, potassium reduction, increasing values of pH, alkalinity, calcium and magnesium. The instability was observed when there was increase in organic load to 12 g L-1 with subsequent recovery.

Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

Lidiam - direct-link computer for the photo measurement apparatus of a bubble chamber; Lidiam - liaison directe d'une calculatrice aux appareils de mesure de photos de chambre a bulles

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

The treatment of bubble chamber pictures can be considerably improved by connecting conventional measuring machines to a small computer providing continuous control of the later as well as the immediate detection of errors and their immediate corrections. The computer will also perform first processing of the measured data. In addition the system described will improve the effective yield of each apparatus and facilitates the control of the data. A description is given here of the apparatus and of some future extensions. (author) [French] L'exploitation des cliches de chambre a bulles peut etre sensiblement amelioree par la jonction directe des appareils de mesure a une calculatrice qui permettra la detection immediate des erreurs et leurs corrections, le controle continu des appareils et un premier traitement des evenements mesures. De plus, le rendement effectif des appareils de mesure sera notablement augmente et la gestion ainsi que l'exploitation des mesures grandement facilitees. Nous presentons ici la description d'un dispositif de ce genre ainsi que les projets d'extensions envisagees.

Department of Reactor Technology annual progress report 1 January -31 December 1977

International Nuclear Information System (INIS)

1978-04-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

Measurement of Placental Blood Flow with {sup 133}Xe in Normal and Pathological Human Pregnancy; Mesure du Debit Placentaire dans les Grossesses Normales et Pathologiques

Energy Technology Data Exchange (ETDEWEB)

Pontonnier, G.; Delmas, H.; Farre, J.; Favretto, R. [Clinique Obstetricale, Hopital de la Grave, Toulouse (France)

1971-02-15

practical standpoint, it provides one more element for the surveillance of pregnancies involving high foetal risk and makes it possible to study the action of drugs administered to the mother on placental blood flow. (author) [French] La plupart des auteurs s'accordent a reconnaitre l'importance des alterations de la circulation placentaire dans la genese de la souffrance foetale chronique. Or il n'existait pas jusqu'ici de technique qui permette une evaluation quantitative de l'hemodynamique placentaire. Les auteurs presentent une methode de mesure du debit placentaire qui est une application des radioisotopes a la mesure des debits locaux. Ils utilisent le xenon-133 en solution dans du serum physiologique; ce gaz radioactif presente l'avantage d'etre inerte et instantanement diffusible. Apres localisation radiographique ou ultrasonique du placenta, ils injectent par voie trans-abdominale 50 {mu}Ci de xenon dans le placenta. L'enregistrement de la courbe de clearance du xenon-133 est effectue au moyen d'un detecteur a scintillation. La courbe obtenue est en meme temps enregistree sur un inscripteur lineaire et transmise a un calculateur automatique. Les auteurs ont ainsi realise 111 mesures de debit placentaire: 45 dans des grossesses normales, 59 dans des grossesses pathologiques et 7 apres perfusion medicamenteuse. Les mesures effectuees ont permis pour la premiere fois de chiffrer la valeur du debit placentaire chez la femme. Pour les grossesses normales, entre la 32{sup e} et la 41{sup e} semaine, il est de 145 ml/100 g/min. Les mesures effectuees au cours de grossesses pathologiques (hypertension arterielle, dysgravidie, infection urinaire, diabete, grossesse prolongee) ont montre que celles-ci s'accompagnent d'une diminution du debit placentaire statistiquement significative. L'importance de cette diminution est en rapport avec la clinique et l'etat de l'enfant a la naissance. Cette methode de mesure, facilement reproductible chez la meme femme, presente donc un

A semi-automatic device for measuring osmotic pressures (1962); Un dispositif semi-automatique pour la mesure des pressions osmotiques (1962)

Energy Technology Data Exchange (ETDEWEB)

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

1962-07-01

A cryoscopic apparatus for measuring osmotic pressure in small samples (0.1 ml) is described. The sample is frozen by air cooled dry ice or liquid nitrogen; the temperature is measured by a thermistor resistance and a recording millivoltmeter. (author) [French] Un appareil cryometrique pour la mesure des pressions osmotiques sur des petits echantillons (0,1 ml) est decrit. L'echantillon est congele par une circulation d'air refroidi par de la carboglace ou de l'azote liquide; sa temperature est mesuree par une thermitance associee a un millivoltmetre enregistreur. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

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

1965-10-15

distance intercalaire fournit a un enregistreur une tension stable de sortie de 1 V. Il est facile de mesurer les variations des distances intercalaires avec une precision de 5 {mu}m. Les mesures ont ete faites pour plusieurs elements combustibles. Les resultats et la reproductibilite sont tres satisfaisants. (author) [Spanish] Los elevados flujos termicos que suelen alcanzarse en los recientes reactores de ensayo de materiales, exigen recorridos adecuados para lograr una transmisiun uniforme de calor y una disipacion segura del mismo, evitando asf la formacion de vapor en la masa. Ademas, a fin de controlar el hinchamiento y el comportamiento del combustible en el reactor, tambien debe medirse la capa de agua en experimentos realizados despues de la irradiacion, con elementos combustibles agotados. A tal efecto se ha disenado una sonda ultrasonica destinada a medir, en una longitud de 1 m el espesor de 3 mm de agua correspondiente al elemento combustible BR-2. En el caso de los experimentos posteriores a la irradiacion, es necesario trabajar con el elemento combustible sumergido en un tanque de agua, a profundidad no menor de 6 m. La sonda puede resistir una prolongada inmersion en agua, y no le afectan las dosis normales de radiacion gamma. Aunque proyectado conforme al metodo usual de reflexion de impulsos, el sistema permite separar pulsos emitidos y reflejados, usando un cristal ferro-electrico de 10 MHz, con elevada disipacion inherente de energia. Puede usarse un osciioscopio para la lectura, en cuyo caso el tiempo se representa en el eje horizontal, regulandose la velocidad de barrido de manera que sea directamente proporcional a la velocidad de propagacion de la onda, es decir, al espesor de la capa de agua. Este tipo de representacion da resultados satisfactorios cuando setrata de un numero limitado de mediciones, pero sin duda resulta mas conveniente el registro grafico. En este caso, se da a los impulsos emitidos y reflejados la forma deseada y se les inyecta

Backfitting of research reactors

International Nuclear Information System (INIS)

Delrue, R.; Noesen, T.

1985-01-01

The backfitting of research reactors covers a variety of activities. 1. Instrumentation and control: Control systems have developed rapidly and many reactor operators wish to replace obsolete equipment by new systems. 2. Pool liners: Some pools are lined internally with ceramic tiles. These may become pervious with time necessitating replacement, e.g. by a new stainless steel liner. 3. Heat removal system: Deficiencies can occur in one or more of the cooling system components. Upgrading may require modifications of the system such as addition of primary loops, introduction of deactivation tanks, pump replacement. Recent experience in such work has shown that renewal, backfitting and upgrading of an existing reactor is economically attractive since the related costs and delivery times are substantially lower than those required to install a new research reactor

Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

International Nuclear Information System (INIS)

1989-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

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)

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

Anaerobic digestion of cheese whey using up-flow anaerobic sludge blanket reactor

Energy Technology Data Exchange (ETDEWEB)

Yan, J.Q.; Lo, K.V.; Liao, P.H.

1989-01-01

Anaerobic treatment of cheese whey using a 17.5-litre up-flow anaerobic sludge blanket reactor was investigated in the laboratory. The reactor was studied over a range of influent concentration from 4.5 to 38.1 g chemical oxygen demand per litre at a constant hydraulic retention time of 5 days. The reactor start-up and the sludge acclimatization were discussed. The reactor performance in terms of methane production, volatile fatty acids conversion, sludge net growth and chemical oxygen demand reduction were also presented in this paper. Over 97% chemical oxygen demand reduction was achieved in this experiment. At the influent concentration of 38.1 g chemical oxygen demand per litre, an instability of the reactor was observed. The results indicated that the up-flow anaerobic sludge blanket reactor process could treat cheese whey effectively.

Mesurer le droit de participer à la vie culturelle: le développement des indicateurs

NARCIS (Netherlands)

Donders, Y.; Bouchard, J.; Gandolfi, S.; Meyer-Bisch, P.

2013-01-01

Dans les domaines des droits de l'homme, de la culture et du développement, des cadres et mécanismes sont développés afin de mesurer et de qualifier les relations entre ces domaines et leur impact sur la vie des individus en des communautés. Les indicateurs constituent un élément important de ces

Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

International Nuclear Information System (INIS)

Ratanatongchai, Wichian

2009-07-01

Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

Mesophilic and thermophilic biotreatment of BTEX-polluted air in reactors.

Science.gov (United States)

Mohammad, Balsam T; Veiga, María C; Kennes, Christian

2007-08-15

This study compares the removal of a mixture of benzene, toluene, ethylbenzene, and all three xylene isomers (BTEX) in mesophilic and thermophilic (50 degrees C) bioreactors. In the mesophilic reactor fungi became dominant after long-term operation, while bacteria dominated in the thermophilic unit. Microbial acclimation was achieved by exposing the biofilters to initial BTEX loads of 2-15 g m(-3) h(-1), at an empty bed residence time of 96 s. After adaptation, the elimination capacities ranged from 3 to 188 g m(-3) h(-1), depending on the inlet load, for the mesophilic biofilter with removal efficiencies reaching 96%. On the other hand, in the thermophilic reactor the average removal efficiency was 83% with a maximum elimination capacity of 218 g m(-3) h(-1). There was a clear positive relationship between temperature gradients as well as CO(2) production and elimination capacities across the biofilters. The gas phase was sampled at different depths along the reactors observing that the percentage pollutant removal in each section was strongly dependant on the load applied. The fate of individual alkylbenzene compounds was checked, showing the unusually high biodegradation rate of benzene at high loads under thermophilic conditions (100%) compared to its very low removal in the mesophilic reactor at such load (<10%). Such difference was less pronounced for the other pollutants. After 210 days of operation, the dry biomass content for the mesophilic and thermophilic reactors were 0.300 and 0.114 g g(-1) (support), respectively, reaching higher removals under thermophilic conditions with a lower biomass accumulation, that is, lower pressure drop. (c) 2007 Wiley Periodicals, Inc.

New measuring and protection system at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

International Nuclear Information System (INIS)

1991-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

International Nuclear Information System (INIS)

1992-08-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Application and expression of HSV gG1 protein from a recombinant strain.

Science.gov (United States)

Yan, Hua; Yan, Huishen; Huang, Tao; Li, Guocai; Gong, Weijuan; Jiao, Hongmei; Chen, Hongju; Ji, Mingchun

2010-11-01

According to the homologous sequence of glycoprotein G1 (gG1) genes from different strains of herpes simplex virus type 1 (HSV-1), a pair of primers was designed to amplify the gG1 gene fragment by PCR. Both the PCR product and the pGEX-4T-1 vector were digested with EcoR I and Sal I. The gG1 gene fragment was subcloned into the digested pGEX-4T-1 vector to construct a recombinant plasmid (pGEX-4T-1-gG1). The resultant plasmid was identified by dual-enzyme digestion and sequence analysis, and then transformed into Escherichia coli BL21 for expression under the induction of isopropyl β-D-1-thiogalactoside (IPTG). The expressed GST-gG1 fragment was detected by SDS-PAGE and purified by affinity chromatography. The properties of GST-gG1 fragment were evaluated by immunoblot analysis. Enzyme-linked immunosorbent assays (ELISAs) based on the GST-gG1 fragment were used for determining IgG or IgM to HSV-1. The GST-gG1 fragment-specific ELISA was also compared with ELISA with whole-HSV-1 antigen and commercial ELISA kits. The gG1-specific IgG and IFN-γ producing CD8+ T cells were induced in mice immunized with the GST-gG1 fragment. These results indicated that the GST-gG1 fragment could be used for replacing whole-virus antigen to detect IgM and IgG to HSV-1 in human sera, which provided a strategy for developing vaccines to protect HSV-1 infection using gG1 fragment. Copyright © 2010 Elsevier B.V. All rights reserved.

Aspects et mesure de la qualité de vie : évolution et renouvellement des tableaux de bord métropolitains

Directory of Open Access Journals (Sweden)

Pierre J. Hamel

2008-02-01

Full Text Available La mesure de la qualité de vie à l’intérieur des espaces urbains préoccupe les administrations publiques depuis nombre d’années. Cet article passe en revue les modèles de mesure de la qualité de vie développés par les métropoles canadiennes. Il s’interroge sur l’évolution de ces modèles de mesure et sur leur capacité à rendre compte des différentes problématiques désormais associées à la notion de qualité de vie comme le développement social, l’environnement, la société du risque, les ambiances urbaines ou la compétitivité urbaine.For a number of years now, government bodies at all levels have been concerned with measuring quality of life within urban areas. This paper reviews the models used by Canada’s metropolises to measure quality of life. It examines how the models have evolved and their capacity to consider various issues which have become associated with the notion of quality of life, such as social development, environment, risk society, urban surroundings, or urban competitiveness.

Thermohydraulic analysis for power increase of IEAR-1 reactor

International Nuclear Information System (INIS)

Umbehaun, Pedro E.; Bastos, Jose L.F.

1996-01-01

In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

Determination of the concentration of {alpha} emitting radioactive aerosols; Mesure de la concentration des aerosols radioactifs emetteurs {alpha}

Energy Technology Data Exchange (ETDEWEB)

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

1953-06-15

In the first part of this work the techniques used for the quantitative measurement of the concentrations of aerosols carrying short lived (radon or thoron daughters) or long lived (uranium) {alpha} radioactive emitters are described. In the second part the author investigates the problem of the determination of radon concentration in air by means of activity determinations on airborne dusts. Special reference is made to the measurement of the radon active deposit on two types of dusts (iron oxide (yellow) and uranium oxide) in small chambers (6 liters). In the third part are given data resulting from determinations of radon and thoron concentrations in atmospheric air in the south of Paris area using this method. (author) [French] Dans la premiere partie de ce travail on expose les techniques utilisees pour la mesure quantitative des concentrations d'aerosols contenant des emetteurs radioactifs {alpha}, tant pour ceux a vie courte (derives du radon ou du thoron) que pour ceux a vie longue (uranium). Dans la seconde partie on traite le probleme de la determination de la concentration de l'air en radon par la mesure de l'activite des poussieres ayant sejourne dans cet air. En particulier, on indique pour de petits volumes (6 litres) la proportion de depot actif du radon qui est fixee sur deux types de poussieres (limonite et oxyde d'uranium) en fonction de la concentration de celles-ci. Dans la troisieme partie on donne quelques exemples de mesure par cette methode de la concentration en radon et en thoron de l'atmosphere de la region parisienne. (auteur)

Concentration urinaire de bisphénol A et obésité chez les adultes : résultats de l’Enquête canadienne sur les mesures de la santé

Directory of Open Access Journals (Sweden)

Minh T. Do

2017-01-01

Full Text Available Introduction : Des études chez l’animal ont révélé que l’exposition au bisphénol A (BPA affecte le métabolisme des lipides et favorise la prise de poids. Des études épidémiologiques récentes appuient aussi l’existence d’un lien entre le BPA et l’obésité chez l’humain, mais la plupart d’entre elles se limitent à une seule mesure de l’adiposité ou ne tiennent pas compte des facteurs de confusion alimentaires possibles. Cette étude vise à examiner les associations entre les concentrations urinaires de BPA et les mesures de l’adiposité dans un échantillon national représentatif des adultes canadiens. Méthodologie : Nous avons réalisé des analyses à l’aide de données de biosurveillance et de données anthropométriques mesurées directement auprès de 4 733 adultes de 18 à 79 ans dans le cadre de l’Enquête canadienne sur les mesures de la santé (2007-2011. Nous avons utilisé des modèles de régression logistique multinomiale et binaire pour estimer les associations entre, d’une part, les concentrations urinaires de BPA et, d’autre part, les diverses catégories d’indice de masse corporelle (IMC (embonpoint par opposition à poids insuffisant ou normal; obésité par opposition à poids insuffisant ou normal et un tour de taille élevé (102 cm ou plus pour les hommes; 88 cm ou plus pour les femmes en tenant compte des facteurs de confusion possibles. Des analyses de régression linéaire ont aussi été effectuées pour évaluer les associations entre les concentrations urinaires de BPA et des mesures de l’IMC et du tour de taille. Résultats : On a pu associer positivement la concentration urinaire de BPA et l’obésité (définie par l’IMC. Le rapport de cotes s’est situé à 1,54 (intervalle de confiance [IC] à 95 % : 1,002 à 2,37 dans le quartile de BPA le plus élevé par rapport au plus bas (test de tendance, p = 0,041. La concentration urinaire de BPA n’a pas été associ

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Analysis of the accuracy of certain methods used for measuring very low reactivities; Analyse de la precision de certaines methodes de mesure de tres basses reactivites

Energy Technology Data Exchange (ETDEWEB)

Valat, J; Stern, T E

1964-07-01

The rapid measurement of anti-reactivities, in particular very low ones (i.e. a few tens of {beta}) appears to be an interesting method for the automatic start-up a reactor and its optimisation. With this in view, the present report explores the various methods studied essentially from the point of view of the time required for making the measurement with a given statistical accuracy, especially as far as very low activities are concerned. The statistical analysis is applied in turn to: the methods for the natural background noise (auto-correlation and spectral density); the sinusoidal excitation methods for the reactivity or the source, with synchronous detection ; the periodic source excitation method using pulsed neutrons. Finally, the statistical analysis leads to the suggestion of a new method of source excitation using neutronic random square waves combined with an intercorrelation between the random excitation and the resulting output. (authors) [French] La mesure rapide des antireactivites, en particulier celle des tres basses (soit quelques dizaines de {beta}), apparait comme une voie interessante pour le demarrage automatique d'un reacteur et son optimalisation. Dans cette optique, le present rapport explore diverses methodes etudiees essentiellement sous l'angle de la duree de mesure necessaire a une precision relative statistique donnee, plus particulierement en ce qui concerne les tres basses reactivites. L'analyse statistique porte successivement sur: les methodes du bruit de fond naturel (autocorrelation et densite spectrale); les methodes d'excitation sinusoidale de reactivite ou de source, avec detection synchrone; la methode d'excitation periodique de source par neutrons pulses. Enfin l'analyse statistique amene a proposer une methode nouvelle d'excitation de source par creneaux neutroniques aleatoires alliee a une intercorrelation entre l'excitation aleatoire et la sortie resultante. (auteurs)

Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

Energy Technology Data Exchange (ETDEWEB)

Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

1966-07-01

The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

Bonheur et progrès : mesurer le bien-être au Bhoutan et au Canada ...

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

31 janv. 2011 ... Étant l'un des pays les moins développés, le Bhoutan s'inquiète de la mondialisation et il est déterminé à protéger sa spécificité culturelle. Il veut préserver ses valeurs sociales en les ancrant dans des termes que le monde entier pourra comprendre et respecter, c'est-à-dire en les quantifiant. En mesurant ...

Mesures de contraintes in-situ. Méthode de relaxation des carottes Measuring in-Situ Stresses. Relaxation Method with Core Samples

Directory of Open Access Journals (Sweden)

Perreau P.

2006-11-01

Full Text Available Dans cet article, on se propose de présenter les premiers résultats de l'étude de la méthode d'évaluation des contraintes par mesure de déformations différées d'une carotte après son extraction. Le travail correspondant a été réalisé dans le cadre du projet ARTEP Fracturation hydraulique . Les principes de cette méthode et les quelques éléments d'interprétation récemment publiés dans la littérature sont exposés dans un premier temps. Les résultats de deux campagnes de mesures sur deux puits de la SNEA-P (Soudron, novembre 1985 et Lanot, juillet 1986 sont ensuite présentés. Ces essais ont mis en évidence que les déformations différées d'une carotte dues au relachement des contraintes sont effectivement mesurables. Cependant, une interprétation quantitative de ces mesures nécessite une amélioration des conditions expérimentales (stabilisation thermique, stabilisation de l'état de saturation. This article describes the first results of research on a method of evaluating stresses by measuring the differred deformations of a core sample after it has been extracted. The corresponding research was done within the framework of an ARTEP project on Hydraulic Fracturing . The principles of this method and several interpretation aspects published recently in the literature are described in the first part. Then the results of two measurement campaigns using two SNEA-P wells (Soudron in November 1985 and Lanot in July 1986 are described. These tests revealed that the differed deformations of a core sample due to the relaxing of stresses can effectively be measured. However, a quantitative interpretation of these measurements requires an improvement to be made in the experimental conditions (thermal stabilization, stabilization of the state of saturation.

Department of Reactor Technology: annual progress report 1 January - 31 December 1976

International Nuclear Information System (INIS)

1977-06-01

The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

Measurement of the scattering cross-section of non-fissile nuclei by the time-of-flight method. Determination of the spin of some excited States; Mesure de la section efficace de diffusion de noyaux non fissiles par la methode du temps de vol. Determination du spin de quelques etats excites

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

The Saclay 45 MeV Linear Accelerator is used as a pulsed neutron source to carry out experiments in the resonance energy region. Sometimes, when the right conditions are available, the transmission on measurements can give all the resonance parameters such as E{sub r}, {gamma}{sub n}, 2 g {gamma}{sub n} and J = I {+-} 1/2, when I is the spin of the target nucleus; but usually the complementary data from scattering measurements are required to obtain the value of spin J. This experiment is being run here and we present some data on Rh{sup 103} and Au{sup 197} for energy range between 40 to 700 eV. (author) [French] L'utilisation de l'Accelerateur Lineaire de Saclay de 45 MeV comme source de neutrons pulsee a permis la realisation d'un certain nombre d'experiences dans la region des resonances. Lorsque les conditions sont satisfaisantes, les mesures de transmission peuvent donner tous les parametres des resonances : E{sub r}, {gamma}{sub n}, 2 g {gamma}{sub n} et J = I {+-} 1/2 ou I est le spin du noyau cible; mais souvent les resultats complementaires des mesures de diffusion sont necessaires a l'obtention de la valeur du spin J. Cette experience a ete realisee ici et nous presentons les resultats obtenus sur le {sup 103}Rh et {sup 197}Au dans la gamme d'energie comprise entre 40 et 700 eV. (auteur)

Limit to the measurement of feeble activities using ionization chambers; Limite des possibilites de mesure de faibles activites au moyen de chambres d'ionisation

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The present account based on experience acquired in the Biology Service, in measuring feeble activities of tritium and carbon-14 has been prepared for the benefit hose who have to carry out measurements of feeble activities using ionization chambers. Precision are given on the behaviour and actual performance of vibrating condenser electrometers, based on approximately two years operating experience. It is shown that the possibilities of utilisation are not limited as is generally believed by insufficient sensitivity and stability of the electronic equipment, but by the existence of various parasitic phenomena coming from the ionization chamber itself, which make very difficult the measurement of ionization currents which are less than 10{sup -14} A. (author)Fren. [French] Le present compte rendu, base sur l'experience acquise au Service de Biologie dans la mesure de faibles activites de tritium et de carbone-14, est redige a l'intention des personnes ayant a effectuer des mesures de tres faibles activites au moyen de chambres d'ionisation. Il donne des precisions sur le comportement et les performances reelles des electrometres a condensateur vibrant, basees sur environ deux ans d'utilisation, et demontre que les possibilites de mesure ne sont pas limitees - comme on le croit generalement - par l'insuffisance de sensibilite et de stabilite de l'appareillage electronique, mais par l'existence de divers phenomenes parasites dont la chambre d'ionisation est le siege et qui rendent tres difficiles la mesure de courants d'ionisation inferieurs a 10{sup -14} A. (auteur)

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Neutron density optimal control of A-1 reactor analoque model

International Nuclear Information System (INIS)

Grof, V.

1975-01-01

Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

Application of the similitude principle to gamma-gamma density measurements; Application du principe de similitude a la mesure gamma-gamma de densite

Energy Technology Data Exchange (ETDEWEB)

Czubek, J A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires. Departement d' Electronique Generale, Service d' Electronique Industrielle; Institut de Recherches Nucleaires, Dep. VI, Cracow (Poland)

1966-07-01

The work presented here deals with the problem of the application of the similitude principle to rock density measurements by the gamma-gamma method. A formula is presented which makes it possible to transform results of gamma-gamma measurements carried out on models in order to make them suitable for comparison with results obtained under actual field conditions. Both the space coordinates and the densities are transformed. This transformation makes it possible to obtain a calibration curve as a function of the density for a gamma-gamma probe using only a single model of given density. The influence has also been studied of the chemical composition on the results obtained from gamma-gamma measurements. A method has been developed for estimating the equivalent Z parameter of the medium; the possibility of completely eliminating the influence of the chemical composition of the medium on the measurement results has been studied. (author) [French] L'etude presentee ci-dessous traite le probleme de l'application du principe de similitude aux mesures de densite des roches par la methode gamma-gamma. Nous indiquons une formule qui permet de transformer les resultats des mesures gamma-gamma effectuees sur les modeles pour les comparer aux resultats obtenus dans les conditions reelles sur le terrain. On transforme les coordonnees spatiales ainsi que les densites. Cette transformation donne la possibilite d'obtenir une courbe d'etalonnage (en fonction de la densite) pour une sonde gamma-gamma en utilisant un seul modele de densite donnee. On a etudie aussi l'influence de la composition chimique sur les resultats obtenus des mesures gamma-gamma. On a etabli une methode d'estimation du parametre Z equivalent du milieu, ainsi que la possibilite d'eliminer completement l'influence de la composition chimique du milieu sur les resultats des mesures de densite. (auteur)

Resistivity measurements using a direct current induction method (1963); Mesure de resistivite par la methode d'induction en courant continu (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

The conventional methods for measuring electrical resistivities necessitate the fixing of electrical contacts on the sample either mechanically or by soldering. Furthermore it is also necessary to carry,out the measurements on low cross-section samples which are not always easy to obtain. Our direct-current induction method on the other hand requires no contacts and can easily be applied to samples of large cross-section. The sample is placed in a uniform magnetic field; at the moment when the current is cut, eddy currents appear in the sample which tend to oppose the disappearance of the field. The way in which the magnetic flux decreases in the sample makes it possible to determine the resistivity of the material. This method has been applied to samples having diameters of between 1 and 30 mm in the case of metals which are good conductors. It gives a value for the local resistivity and makes it possible to detect any variation along a sample. The measurements can be carried out at all temperature from a few degrees absolute to 500 deg. C. We have used the induction method to follow the purification of beryllium by zone-melting; it is in effect possible to estimate the purity of a material by resistivity measurements. We have measured the resistivity along each bar treated by the zone-melting technique and have thus, localised the purest section. High temperature measurements have been carried out on uranium carbide and on iron-aluminium alloys. This method constitutes an interesting means of investigation the resistivity of solid materials. Its accuracy and rapidity make it particularly adapted both to fundamental research and to production control. (authors) [French] Les methodes classiques de mesure de resistivite electrique imposent la realisation sur l'echantillon de contacts electriques obtenus soit mecaniquement, soit par soudure. En outre, elles demandent, le plus souvent, d'effectuer les mesures sur des echantillons de faible section qu'il n'est pas

Department of Reactor Technology annual progress report 1 January - 31 December 1978

International Nuclear Information System (INIS)

1979-04-01

The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1993-March 31, 1994

International Nuclear Information System (INIS)

1994-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

International Nuclear Information System (INIS)

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Amorphous silicon cells for the measurement of photosynthetically active radiation; Utilisation des cellules au silicium amorphe pour la mesure du rayonnement photosynthétiquement actif (400-700 nm)

Energy Technology Data Exchange (ETDEWEB)

Chartier, M. [Institut National de la Recherche Agronomique, Thiverval-Grignon (France); Bonchretien, P.; Allirand, J. M.; Gosse, G.

1989-07-01

Numerical simulation and experimental measurements from amorphous silicon cells in comparison with these now used in ecophysiology illustrate the interest of amorphous silicon cells for the measurement of PAR incident, reflected, and transmitted below the canopy [French] La simulation et la confrontation expérimentale ces réponses des cellules au silicium amorphe par rapport à celles des capteurs existants montrent l’intérêt des cellules au silicium amorphe pour la mesure du rayonnement PAR (exprimé en densité de flux de photons) incident, réfléchi et transmis sous un couvert végétal.

Modification of the IAN-R1 reactor

International Nuclear Information System (INIS)

Jaime, J.; Ahumada, S.; Spin, R.A.

1990-01-01

The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

The E142 SLAC experiment: measurement of the neutron g{sup n}{sub 1}(x) spin structure function; Experience E142 au SLAC: mesure de la fonction de structure en spin g{sup n}{sub 1}(x) du neutron

Energy Technology Data Exchange (ETDEWEB)

Roblin, Y

1995-04-21

This thesis describes the E142 experiment which has been carried out at the Stanford Linear Accelerator (SLAC), USA, from October to December 1992. This experiment of polarized inelastic scattering of a 22.6 GeV electron beam on a polarized helium 3 target has allowed the first measurement of the neutron g{sup n}{sub 1}(x) spin structure function. The knowledge of this structure function gives informations on the nucleon spin structure. On the other hand, the g{sup n}{sub 1}(x) structure function integral value on the 01 domain has been obtained for a Q{sup 2} mean value of 2 GeV{sup 2} after some extrapolations. This value is at about two standard deviations away from the theoretical predictions of the Ellis-Jaffe rule. Thanks to the existing experimental results for the proton (E143 experiment), the Bjorken sum rule has been precisely tested and is perfectly compatible with the theoretical value. The results have allowed to estimate the nucleon spin fraction carried by the quarks. (J.S.). 86 refs., 58 figs., 13 tabs.

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Treatment of pig excreta using an SCFBR anaerobic reactor

Directory of Open Access Journals (Sweden)

Kevin G. Molina T.

1999-01-01

Full Text Available A new anaerobic reactor called the Sludge Central Fixed Bed Reactor (SCFBR was built and evaluated for the treatment of liquid residue from the pig farms. The SCFBR has three main parts. The lower area is for sludge, the middle part consists of a concentrically packed zone and an upper area for the separation of solids, liquids and gases. The 28.51 SCFBR reactor was evaluated over a period of 210 days, using three organic loads of 0.548,0.421 and 1.239 g COD/1 day. Initially, the reactor was fed non-continuously using 10 and 10.7 days Hydraulic Retention Times (HRT. The HRT was later decreased to 3.87 days using continuous feeding. For the three 0.548, 0.421 and 1.239 g COD/1 day organic loads, Chemical Oxygen Demand (COD removal was 68%, 81% and 73% respectively and Volatile Solids (VS removal was 53.5%, 55.8% and 50.1% respectively. The SCFBR performed well, as shown by the removal efficiency and stability obtained. A microphotograph of sludge from the lower zone is presented, showing high Methanosaeta (Methanothrix presence.

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-02-15

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

Seismic analysis of a large pool-type LMR [liquid metal reactor

International Nuclear Information System (INIS)

Wang, C.Y.; Gvildys, J.

1989-01-01

This paper describes the seismic study of a 450-MWe liquid metal reactor (LMR) under 0.3-g SSE ground excitation. Two calculations were performed using the new design configuration. They deal with the seismic response of the reactor vessel, the guard vessel and support skirt, respectively. In both calculations, the stress and displacement fields at important locations of those components are investigated. Assessments are also made on the elastic and inelastic structural capabilities for other beyond-design basis seismic loads. Results of the reactor vessel analysis reveal that the maximum equivalent stress is only about half of the material yield stress. For the guard vessel and support skirt, the stress level is very small. Regarding the analysis if inelastic structural capability, solutions of the Newmark-Hall ductility modification method show that the reactor vessel can withstand seismics with ground ZPAs ranging from 1.015 to 1.31 g, which corresponds to 3.37 to 4.37 times the basic 0.3-g SSE. Thus, the reactor vessel and guard vessel are strong enough to resist seismic loads. 4 refs., 10 figs., 5 tabs

Feasibility studies on large sample neutron activation analysis using a low power research reactor

International Nuclear Information System (INIS)

Gyampo, O.

2008-06-01

Instrumental neutron activation analysis (INAA) using Ghana Research Reactor-1 (GHARR-1) can be directly applied to samples with masses in grams. Samples weights were in the range of 0.5g to 5g. Therefore, the representativity of the sample is improved as well as sensitivity. Irradiation of samples was done using a low power research reactor. The correction for the neutron self-shielding within the sample is determined from measurement of the neutron flux depression just outside the sample. Correction for gamma ray self-attenuation in the sample was performed via linear attenuation coefficients derived from transmission measurements. Quantitative and qualitative analysis of data were done using gamma ray spectrometry (HPGe detector). The results of this study on the possibilities of large sample NAA using a miniature neutron source reactor (MNSR) show clearly that the Ghana Research Reactor-1 (GHARR-1) at the National Nuclear Research Institute (NNRI) can be used for sample analyses up to 5 grams (5g) using the pneumatic transfer systems.

Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

International Nuclear Information System (INIS)

1996-04-01

The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

New instruments and methods for measuring the concentration of radioactive products in the atmosphere; Appareils recents et methodes nouvelles pour la mesure de la concentration des produits radioactifs dans l'atmosphere

Energy Technology Data Exchange (ETDEWEB)

Jehanno, C; Blanc, A; Lallemant, C; Roux, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

) [French] Differents enregistreurs pour aerosols radioactifs ont ete realises pour mesurer la pollution de l'atmosphere des laboratoires ou l'atmosphere exterieure. EAR 600. - Permet de mesurer en continu instantanement et 3 a 10 heures apres le prelevement des concentrations d'aerosols emetteurs {alpha} ou {beta} allant de quelques 10{sup -11} a quelques 10{sup -8} curies par metre cube d'air. EAR 800. - Permet de mesurer en continu des concentrations d'aerosols emetteurs {alpha} allant de 10{sup -11} a 10{sup -5} curie par metre cube d'air, et des concentrations d'aerosols emetteurs {beta} allant de 10{sup -11} a 10{sup -1} curie par metre cube d'air. EAR PLUTONIUM. - Permet de detecter en quelques minutes 1000 DMP (2.10{sup -9} curie par metre cube) et en 8 heures 1 DMP (2.10{sup -12} curie par metre cube). Deux methodes sont utilisees pour separer l'activite due au plutonium de celle due aux derives du radon et du thoron: a) par discrimination d'amplitude, b) par les coincidences (a,b) RaC' et ThC-ThC'. SP 4. - Monte sur jeep, cet ensemble permet la mesure de l'irradiation produite au sol par le panache des piles. La sensibilite est de 5{mu}R/h. A.D.I.R. - Cet appareil autonome et portatif est destine a la mesure instantanee des teneurs en radon de l'atmosphere des mines. Il permet de mesurer des teneurs de l'air variant entre 0,4 et 400.10{sup -10} curie par litre d'air (0,4 et 400 DPM). La mesure des retombees radioactives est faite apres collection de cette activite par un pluviometre special comportant une surface adhesive et une cartouche de resines echangeuses d'ions. La radioactivite des retombees varie entre quelques 10{sup -9} et quelques 10{sup -7} curie par metre carre et par mois. La mesure des concentrations de l'atmosphere en produits de fission est faite apres collection sur papier filtre. Les concentrations mesurees dans l'air au niveau du sol varient entre 10{sup -13} et 10{sup -12} curie par metre cube. (auteur)

4G/5G polymorphism modulates PAI-1 circulating levels in obese women.

Science.gov (United States)

Fernandes, Karla S; Sandrim, Valéria C

2012-05-01

The increase in plasminogen activator inhibitor type 1 (PAI-1) has been described as a risk factor to thrombosis-related diseases. In addition, it has been demonstrated that the variant 4G of polymorphism 4G/5G located in promoter region of PAI-1 gene is associated with higher PAI-1 levels. We investigate the role of this polymorphism on circulating PAI-1 concentration in a population of 57 obese women (23%, 4G/4G; 49%, 4G/5G and 28%, 5G/5G genotypes). Our results demonstrate a genotype-specific modulation on PAI-1 levels in obese women, thus 5G/5G genotype presented significantly lower levels of plasma PAI-1 when compared to 4G/4G group (46 ± 19 ng/mL vs. 63 ± 13 ng/mL, respectively). Our findings indicate that obese carriers of 4G/4G genotype may have increased risk to develop thrombotic diseases.

Background and Measurability of Test Samples in an Emergency Situation; Bruit de Fond et Possibilite de Mesurer des Echantillons en Cas d'Accident Radiologique; 0424 041e 041d 0418 0418 0417 041c 0414 ; Radiactividad de Fondo y Posibilidad de Medir Muestras en Casos de Urgencia

Energy Technology Data Exchange (ETDEWEB)

Widell, C. -O. [AB Atomenergi, Nykoeping (Sweden)

1965-06-15

Background radiation levels in the vicinity of a reactor, after a ''maximum credible accident'', have been calculated. Based on this the measurability of test samples from air and of human bodies is studied. It was found possible to analyse radiometrically air-filter samples from 0.1 m{sup 3} air contaminated to 100 MPC (40 h) {alpha}-activity, 100 MPC (40 h) iodine-131 activity and 10 MPC (40 h) {alpha}-activity in a gamma background of 2 r/h. These measurements were carried out in the normal equipments in the counting-room 400 m away from the actual reactor. It is, however, necessary to use well-shielded counters (e.g. 10-cm lead shield). It was also found satisfactory to make rapid {beta}-analyses with an end-window G-M tube ordinarily used in routine control in a 10-cm lead shield. The analyses of iodine-131 had to be carried out in a gamma spectrometer but measuring times of 100 s were found to be satisfactory. Also the spectrometers need to be shielded with at least 10 cm of lead. It might be difficult to investigate body burdens of the order of one permissible quarterly intake in an ordinary human body counter if the gamma background is increased to very high levels. In such instances it would be useful either to be able to move the human body counter off the site or to have a transportable counter available. The influence of a raised background is illustrated with figures for an iron-room with 16-cm walls and a 4 in X 8 in Nal crystal. All the figures above are calculated with large errors due to large fluctuations in the gamma fields. It is necessary to count the background before as well as after the counting of the sample if no monitor is available. (author) [French] L'auteur a calcule le bruit de fond au voisinage d*un reacteur apres un 'accident maximum previsible'. Sur la base de ces calculs, il etudie la possibilite de mesurer la radioactivite d'echantillons d'air et de tissus humains. Il a ete possible de proceder a l'analyse radiometrique d

Core conversion study from silicide to molybdenum fuel in the Indonesian 30 MW multipurpose reactor G.A. Siwabessy (RSG-GAS)

International Nuclear Information System (INIS)

Sembiring, T.M.; Kuntoro, I.

2005-01-01

This paper describes the core conversion from silicide to molybdenum core through a series of silicide (2.96 gU cm -3 ) - molybdenum (3.55 gUcm -3 ) mixed transition cores for the Indonesian 30 MW-Multipurpose G.A. Siwabessy (RSGGAS) reactor. The core calculations are carried out using the two-dimensional multigroup neutron diffusion method code of Batan-EQUIL-2D. The calculated results showed that the proposed silicide-molybdenum mixed transition cores, using the same refueling/reshuffling scheme, meet the safety criteria and it can be used in safely converting from an all-silicide core to an all-molybdenum core. (author)

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Measurement of the C / H ratio using neutrons; Mesure du rapport C / H au moyen des neutrons

Energy Technology Data Exchange (ETDEWEB)

Martinelli, P [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Ricci, H [Universite de Lima (Peru)

1960-07-01

A probe made up of a Ra ({alpha}, n) Be neutron source and a proportional counter filled with boron trifluoride has been used to measure the C/H ratio in hydrocarbons. The intensity of the thermal neutron flux in the neighbourhood of the detector increases with the concentration of the hydrocarbon hydrogen surrounding it. By measuring the density it is possible to deduce the C/H ratio. It is thus possible to evaluate the C/H ratio with a precision equal to that given by the {beta}-ray transmission method. The errors arising from the chemical nature of the hydrocarbon can be reduced to a minimum. This method has the advantage of allowing the measurement of the C/H ratio of hydrocarbons contained in recipients or thick steel tubing by means an independent portable apparatus. (author) [French] Une sonde constituee d'une source de neutrons Ra ({alpha}, n) Be et d'un compteur proportionnel a remplissage de trifluorure de bore a ete utilisee pour mesurer le rapport C/H dans les hydrocarbures. Le flux des neutrons thermiques au voisinage du detecteur est d'autant plus intense que la concentration en hydrogene de l'hydrocarbure qui entoure la sonde est plus elevee. Une mesure de densite permet d'en deduire le rapport C/H. On peut ainsi evaluer le rapport C/H avec une precision aussi bonne que celle que l'on obtient par transmission de rayons {beta}. Les erreurs provenant de la nature chimique de l'hydrocarbure peuvent etre minimisees. Cette methode presente l'avantage de permettre la mesure du rapport C/H d'hydrocarbures contenus dans des recipients ou des canalisations epaisses en acier a l'aide d'un appareil exterieur transportable. (auteur)

RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Jumel, Stephanie; Van-Duysen, Jean Claude

2005-01-01

Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

Science.gov (United States)

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Cytokinetically quiescent (G0/G1) human multiple myeloma cells are susceptible to simultaneous inhibition of Chk1 and MEK1/2.

Science.gov (United States)

Pei, Xin-Yan; Dai, Yun; Youssefian, Leena E; Chen, Shuang; Bodie, Wesley W; Takabatake, Yukie; Felthousen, Jessica; Almenara, Jorge A; Kramer, Lora B; Dent, Paul; Grant, Steven

2011-11-10

Effects of Chk1 and MEK1/2 inhibition were investigated in cytokinetically quiescent multiple myeloma (MM) and primary CD138(+) cells. Coexposure to the Chk1 and MEK1/2 inhibitors AZD7762 and selumetinib (AZD6244) robustly induced apoptosis in various MM cells and CD138(+) primary samples, but spared normal CD138(-) and CD34(+) cells. Furthermore, Chk1/MEK1/2 inhibitor treatment of asynchronized cells induced G(0)/G(1) arrest and increased apoptosis in all cell-cycle phases, including G(0)/G(1). To determine whether this regimen is active against quiescent G(0)/G(1) MM cells, cells were cultured in low-serum medium to enrich the G(0)/G(1) population. G(0)/G(1)-enriched cells exhibited diminished sensitivity to conventional agents (eg, Taxol and VP-16) but significantly increased susceptibility to Chk1 ± MEK1/2 inhibitors or Chk1 shRNA knock-down. These events were associated with increased γH2A.X expression/foci formation and Bim up-regulation, whereas Bim shRNA knock-down markedly attenuated lethality. Immunofluorescent analysis of G(0)/G(1)-enriched or primary MM cells demonstrated colocalization of activated caspase-3 and the quiescent (G(0)) marker statin, a nuclear envelope protein. Finally, Chk1/MEK1/2 inhibition increased cell death in the Hoechst-positive (Hst(+)), low pyronin Y (PY)-staining (2N Hst(+)/PY(-)) G(0) population and in sorted small side-population (SSP) MM cells. These findings provide evidence that cytokinetically quiescent MM cells are highly susceptible to simultaneous Chk1 and MEK1/2 inhibition.

Status of the spent fuel in the reactor buildings of Fukushima Daiichi 1–4

Energy Technology Data Exchange (ETDEWEB)

Jäckel, Bernd S., E-mail: bernd.jaeckel@psi.ch

2015-03-15

The ratios of the radionuclides Cs-134g and Cs-137 deduced from measurements of liquid samples from the spent fuel pools in Fukushima Daiichi 1–4 are used to interpret the status of the spent fuel assemblies in the pools of the damaged reactor buildings. The different natures of the production of Cs-134g (neutron capture product of Cs-133) and Cs-137 (cumulative fission product from mass chain 137) and the different half-lives (2.06 years and 30.17 years respectively) require a complicated calculation of the mass and activity of the two nuclides. These masses are depending on the local burn up of the fuel, the burn up history and the radioactive decay. Calculation of the neutron capture product Cs-134g is particularly complicated, because the production of Cs-133 (stable cumulative fission product from mass chain 133) has to be taken into account. The neutron capture cross section for Cs-133 for thermal neutrons is well known, but the energy spectrum of the neutrons in a reactor includes higher energies according to the degree of moderation. Therefore the cross section was fitted from a gamma scan of spent fuel rods in a hot cell. The method of the calculation of the nuclide activities and the interpretation of the gamma measurements of the spent fuel pool samples from Fukushima Daiichi 1–4 are described in detail. It could be shown that at most only very minor mechanical damage of some spent fuel elements occurred during the accident and the later phase of the clearing work.

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

International Nuclear Information System (INIS)

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

Energy Technology Data Exchange (ETDEWEB)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

International Nuclear Information System (INIS)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Plasminogen Activator Inhibitor-1 (PAI-1) gene 4G/5G alleles frequency distribution in the Lebanese population.

Science.gov (United States)

Shammaa, Dina M R; Sabbagh, Amira S; Taher, Ali T; Zaatari, Ghazi S; Mahfouz, Rami A R

2008-09-01

Plasminogen activator inhibitor-1 (PAI-1) is an inhibitor of fibrinolysis. Increased plasma PAI-1 levels play an essential role in the pathogenesis of cardiovascular risk and other diseases associated with thrombosis. The 4G/5G polymorphism of the PAI-1 promoter region has been extensively studied in different populations. We studied 160 healthy unrelated Lebanese individuals using a reverse hybridization PCR assay to detect the 5G/5G, 4G/5G and, 4G/4G genotypes of the PAI-1 gene and the frequencies of the 4G and 5G alleles. We found that 4G/5G genotype was the most prevalent (45.6%) followed by 5G/5G (36.9%) and 4G/4G (17.5%). The frequencies of the 4G and 5G alleles were calculated to be 0.403 and 0.597, respectively. Compared to other ethnic communities, the Lebanese population was found to harbour a relatively high prevalence of the rare 4G allele. This, in turn, may predispose this population to develop cardiovascular diseases and other thrombotic clinical conditions. This study aids to enhance our understanding of the genetic features of the Lebanese population.

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Results of concentration measurements of artificial radioactive aerosols in the lower atmosphere; Resultats des mesures de concentration, dans la basse atmosphere, des aerosols radioactifs artificiels

Energy Technology Data Exchange (ETDEWEB)

Ardouin, B; Jehanno, C; Labeyrie, J; Lambert, G; Tanaevsky, O; Vassy, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

This report gives the results of the measurements of artificial gross-{beta}-radioactivity in aerosols in the lower atmosphere; these measurements have been made by the Electronic Physics Service of the Electronic Department, and by the Atmospheric Physics Laboratory of the Paris Science Faculty. The measurements were begun in September 1956 and were continued in an increasing number of stations both in France and in the rest of the world. The present report deals with the period up to the end of august 1961, that is up to the end of the nuclear moratorium. After recalling the constitution and the properties of radioactive aerosols present in the atmosphere, the authors describe the measurement methods, estimate their accuracy and discuss various aspects of the results. (authors) [French] Ce rapport contient les resultats des mesures de radioactivite {beta} globale d'origine artificielle des aerosols dans la basse atmosphere, effectuees conjointement par le Service d'Electronique Physique du Departement d'Electronique et le Laboratoire de Physique de l'Atmosphere de la Faculte des Sciences de Paris. Les mesures ont commence en septembre 1956 et ont ete poursuivies dans un nombre croissant de stations, tant en France que dans le reste du monde. Le present rapport s'arrete a la fin aout 1961, c'est-a-dire au moment de la reprise des essais nucleaires. Apres avoir rappele la constitution et les proprietes des aerosols radioactifs presents dans l'atmosphere, les auteurs indiquent les methodes de mesure utilisees, evaluent leur precision et discutent les differents aspects des resultats de leurs mesures. (auteurs)

Butanol production by Clostridium acetobutylicum in a continuous packed bed reactor.

Science.gov (United States)

Napoli, Fabio; Olivieri, Giuseppe; Russo, Maria Elena; Marzocchella, Antonio; Salatino, Piero

2010-06-01

In this study, we report on a butanol production process by immobilized Clostridium acetobutylicum in a continuous packed bed reactor (PBR) using Tygon rings as a carrier. The medium was a solution of lactose (15-30 g/L) and yeast extract (3 g/L) to emulate the cheese whey, an abundant lactose-rich wastewater. The reactor was operated under controlled conditions with respect to the pH and to the dilution rate. The pH and the dilution rate ranged between 4 and 5, the dilution rate between 0.54 and 2.4 h(-1) (2.5 times the maximum specific growth rate assessed for suspended cells). The optimal performance of the reactor was recorded at a dilution rate of 0.97 h(-1): the butanol productivity was 4.4 g/Lh and the selectivity of solvent in butanol was 88%(w).

Present and future activities of TRIGA RC-1 Reactor

International Nuclear Information System (INIS)

Festinesi, A.

1986-01-01

A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

Neutronic studies in the enrichment reduction of research reactor IEAR-1

International Nuclear Information System (INIS)

Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

1987-01-01

In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

The plasminogen activator inhibitor-1 (PAI-1) gene -844 A/G and -675 4G/5G promoter polymorphism significantly influences plasma PAI-1 levels in women with polycystic ovary syndrome.

Science.gov (United States)

Lin, Sun; Huiya, Zhang; Bo, Liu; Wei, Wei; Yongmei, Guan

2009-12-01

Mutations in the plasminogen activator inhibitor-1 (PAI-1) gene, along with increased PAI-1 levels, have been implicated in the pathogenesis of polycystic ovarian syndrome (PCOS). We investigated a possible influence of the promoter polymorphism (-844 A/G and -675 4G/5G) in the PAI-1 gene on plasma PAI-1 levels in 126 PCOS patients and 97 healthy controls. Levels of total testosterone, luteinizing hormone (LH), follicle stimulating hormone (FSH), fasting plasma glucose (FPG), fasting insulin, and PAI-1 were measured, and body mass index (BMI), waist-to-hip ratio (WHR), LH/FSH ratio, and homeostasis model assessment for insulin resistance (HOMA-IR) were calculated. PAI-1 -675 4G/5G and -844 A/G gene polymorphisms were also performed. Total testosterone, fasting insulin, and PAI-1 levels; BMI, LH/FSH, and HOMA-IR were significantly higher in PCOS patients than controls (P 5G or 5G/5G genotype. The plasma PAI-1 levels of the combination of the PAI-1 -844 A/A and -675 4G/4G or 4G/5G genotypes, or the coadunation of 4G/4G and -844 non-G/G (A/A + A/G) genotypes were significantly high in PCOS women compared with controls. A trend to a positive interaction between PAI-1 -675 4G/5G and -844 A/G gene polymorphism may elevate plasma PAI-1 levels and hypofibrinolysis, which is probably an important hereditary risk factor in PCOS.

Plasminogen activator inhibitor 1 4G/5G and -844G/A variants in idiopathic recurrent pregnancy loss.

Science.gov (United States)

Magdoud, Kalthoum; Herbepin, Viviana G; Touraine, Renaud; Almawi, Wassim Y; Mahjoub, Touhami

2013-09-01

Plasminogen activator inhibitor type 1 (PAI-1) regulates fibrinolysis, and the common promoter region variants -675G/A (4G/5G) and -844G/A are associated with increased thrombotic risk. Despite evidence linking altered fibrinolysis with adverse pregnancy events, including idiopathic recurrent pregnancy loss (RPL), the contribution of PAI-1 variants to RPL risk remains controversial. We investigated the association between the PAI-1 -844G/A and 4G/5G (-675G/A) variants with altered risk of RPL. This was a case-control study involving 304 women with confirmed RPL and 371 age- and ethnically matched control women. PAI-1 genotyping was performed by PCR single-specific primer -675 (G/A) and real-time PCR (-844G/A) analysis. Minor allele frequency (MAF) of 4G/5G (P 5G single-nucleotide polymorphism (SNP) was significantly associated with RPL under additive, dominant, and recessive genetic models; no association of -844G/A with RPL was seen irrespective of the genetic model tested. Taking common -844G/5G haplotype as reference (OR = 1.00), multivariate analysis confirmed the association of 4G-containing -844A/4G (P 5G, but not -844G/A, PAI-1 variant is associated with an increased risk of RPL. © 2013 John Wiley & Sons Ltd.

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Création d'outils pour la mesure de l'intégration des technologies de ...

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

En collaborant avec les ministères de l'Éducation de la Colombie, du Mexique, du Pérou et de l'Uruguay, les chercheurs élaboreront des méthodologies communes et des indicateurs pertinents localement en vue de surveiller et de mesurer l'intégration des TIC en classe. Les résultats obtenus fourniront aux décideurs les ...

Reliabitity study of the accumulator system for Angra-1 reactor

International Nuclear Information System (INIS)

Santos Maciel, C.C.R.

1980-01-01

The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

Design of a new research reactor : 1st year conceptual design

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T.

2004-01-01

A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

Feasibility study of a magnetic fusion production reactor

Science.gov (United States)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

Radioactive apparatus for measuring the thickness of hot sheet-metal; Jauge d'epaisseur radioactive pour la mesure a chaud de plaques metalliques; Radioaktivnyj izmeritel' tolshchiny goryachego listovogo prokata; Calibrador radiactivo para la medicion en caliente de chapas

Energy Technology Data Exchange (ETDEWEB)

Vasichev, B N; Latyshev, V K; Pliskin, Yu S; Felinger, A K

1962-01-15

In order to achieve a high precision of measurement under rolling conditions, a dynamic method of compensation was chosen, by which the comparison parameter moves through all the values in the range of measurement, whereas the zero component designates only the moment when the measured parameter and the comparison parameter are equal. This method eliminates the mechanical return communications and variable displacements of the static-compensation method, and gives an instrument which is both sufficiently accurate and operates with sufficient speed in the complicated process of rolling. The basic design of the instrument, for checking the thickness of the sheet during the rolling process is described and the factors affecting the accuracy of measurement of a sheet (temperature, undulation, superficial moisture, composition of the sheet, and its position on the roller train) are discussed. The construction of an experimental instrument for industrial use over a thickness range of 14-44 mm is also described. The industrial tests carried out with the instrument showed that, with a Cs{sup 137} source of activity equivalent to 9.2 g radium, the accuracy of measurement of the thickness of hot sheet within the specified range is {+-}0.1 mm. The instrument's operating speed is one measurement per second. It works reliably under the conditions of the rolling mill. (author) [French] Pour obtenir une grande precision dans les mesures en cours de laminage, les auteurs ont choisi la methode dynamique de compensation, dans laquelle le parametre de comparaison passe par toutes les valeurs de la gamme de mesures, la composante zero designant uniquement le moment ou le parametre mesure et le parametre de comparaison sont egaux. Cette methode permet d'eliminer les systemes mecaniques de retour et les deplacements variables qui caracterisent la methode statique de compensation, et d'avoir un appareil a la fois suffisamment precis et d'un fonctionnement suffisamment rapide. Les auteurs

Predictive maintenance technology development at G.A. Siwabessy multipurpose reactor

Energy Technology Data Exchange (ETDEWEB)

Jupiter Sitorus Pane; Imron, M.; Sapto Hartoko; Sentot Alibasya Harahap [Multipurpose Research Reactor G.A. Siwabessy, National Nuclear Energy Agency (Indonesia)

1999-10-01

Safe operation of reactor is certainly influenced by condition of system and component equipped to the reactor's system. In order to maintain the condition of that systems and components, RSG-GAS has arranged maintenance program with time-basis. All 6 (six) groups of reactor systems are maintained within interval of weekly, monthly, three monthly, six-monthly, yearly, five-yearly appropriately. The experience showed that event though the maintenance was performed persistently, the condition of system and component are still not able to determine exactly. The possibility of accidental failure is open since the failure factor are varied and complicated. In order to limit an uncertainty of the component condition a based maintenance shall be introduced. An infrared investigation and manual vibration analysis had been used to diagnose the condition of some RSG-GAS' components. In addition, other alternative technology for predictive maintenance was developed. It is started by computerizing the database maintenance and doing historical review for its aging management, and developing data acquisition and processing equipment using Lab View computer program for collecting and processing signal data from dynamics system. This paper describes briefly the status of those development results. (author)

Predictive maintenance technology development at G.A. Siwabessy multipurpose reactor

International Nuclear Information System (INIS)

Jupiter Sitorus Pane; Imron, M.; Sapto Hartoko; Sentot Alibasya Harahap

1999-01-01

Safe operation of reactor is certainly influenced by condition of system and component equipped to the reactor's system. In order to maintain the condition of that systems and components, RSG-GAS has arranged maintenance program with time-basis. All 6 (six) groups of reactor systems are maintained within interval of weekly, monthly, three monthly, six-monthly, yearly, five-yearly appropriately. The experience showed that event though the maintenance was performed persistently, the condition of system and component are still not able to determine exactly. The possibility of accidental failure is open since the failure factor are varied and complicated. In order to limit an uncertainty of the component condition a based maintenance shall be introduced. An infrared investigation and manual vibration analysis had been used to diagnose the condition of some RSG-GAS' components. In addition, other alternative technology for predictive maintenance was developed. It is started by computerizing the database maintenance and doing historical review for its aging management, and developing data acquisition and processing equipment using Lab View computer program for collecting and processing signal data from dynamics system. This paper describes briefly the status of those development results. (author)

Devices for launching 0.1-g projectiles to 150 km/s or more to initiate fusion. Part 1. Magnetic-gradient and electrostatic accelerators

International Nuclear Information System (INIS)

Brittingham, J.N.

1979-01-01

The feasibility of using magnetic-gradient and electrostatic accelerators to launch a 0.1-g projectile to hypervelocities (150 km/s or more) is studied. Such hypervelocity projectiles could be used to ignite deuterium-tritium fuel pellets in a fusion reactor. For the magnetic-gradient accelerator, several types of projectile were studied: shielded and unshielded copper, ferromagnetic, and superconducting. The calculations revealed the superconducting projectile to be the best of those materials. It would require a 3.2-km-long magnetic-gradient accelerator and achieve a 92% efficiency. This accelerator-projectile combination would be the one most likely to launch a 0.1-g projectile to 150 km/s or more. Its components would cost $58.9 million. The electrostatic accelerator was found to be impractical because of its excessive length of 23 km

Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-04-01

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

Graphite stack corrosion of BUGEY-1 reactor (synthesis)

International Nuclear Information System (INIS)

Petit, A.; Brie, M.

1996-01-01

The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

Sandia reactor kinetics codes: SAK and PK1D

International Nuclear Information System (INIS)

Pickard, P.S.; Odom, J.P.

1978-01-01

The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

Biological sulphide removal from anaerobically treated domestic sewage: reactor performance and microbial community dynamics.

Science.gov (United States)

Garcia, Graziella Patrício Pereira; Diniz, Renata Côrtes Oliveira; Bicalho, Sarah Kinaip; Franco, Vitor Araujo de Souza; Gontijo, Eider Max de Oliveira; Toscano, Rodrigo Argolo; Canhestro, Kenia Oliveira; Santos, Merly Rita Dos; Carmo, Ana Luiza Rodrigues Dias; Lobato, Livia Cristina S; Brandt, Emanuel Manfred F; Chernicharo, Carlos A L; Calabria de Araujo, Juliana

2015-01-01

We developed a biological sulphide oxidation system and evaluated two reactors (shaped similar to the settler compartment of an up-flow anaerobic sludge blanket [UASB] reactor) with different support materials for biomass retention: polypropylene rings and polyurethane foam. The start-up reaction was achieved using microorganisms naturally occurring on the open surface of UASB reactors treating domestic wastewater. Sulphide removal efficiencies of 65% and 90% were achieved with hydraulic retention times (HRTs) of 24 and 12 h, respectively, in both reactors. However, a higher amount of elemental sulphur was formed and accumulated in the biomass from reactor 1 (20 mg S(0) g(-1) VTS) than in that from reactor 2 (2.9 mg S(0) g(-1) VTS) with an HRT of 24 h. Denaturing gradient gel electrophoresis (DGGE) results revealed that the the pink and green biomass that developed in both reactors comprised a diverse bacterial community and had sequences related to phototrophic green and purple-sulphur bacteria such as Chlorobium sp., Chloronema giganteum, and Chromatiaceae. DGGE band patterns also demonstrated that bacterial community was dynamic over time within the same reactor and that different support materials selected for distinct bacterial communities. Taken together, these results indicated that sulphide concentrations of 1-6 mg L(-1) could be efficiently removed from the effluent of a pilot-scale UASB reactor in two sulphide biological oxidation reactors at HRTs of 12 and 24 h, showing the potential for sulphur recovery from anaerobically treated domestic wastewater.

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

OpenAIRE

Gotteland, David; Boulé, Jean-Marie

2004-01-01

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

Ageing problems and renovation programme of ET-RR-1 reactor

International Nuclear Information System (INIS)

Khattab, M.S.; Sultan, M.A.

1995-01-01

Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

Mesures de spectrométrie et de dosimétrie neutron aux postes de travail pour l'étalonnage de dosimètres individuels PGP-DIN

OpenAIRE

Itié, C.; Muller, H.; Asselineau, B.; Médioni, R.; Crovisier, P.; Valier-Bradier, P.; Groetz, J. E.; Piot, J.

2002-01-01

International audience; (ManuAcrit r e p le 3 juillet 2002, accepté le 29 septembre 2002) Dans le cadre de la mise en application des recommandations décrites dans la publication 60 de la CIPR, des mesures de spectrométrie neutron ont été réalisées à plusieurs postes de travail au CEA de Valduc. Le but de ces mesures était la détermination de nouveaux coefficients d'étalonnage h affecter aux dosimètres individuels neutron PGP-DIN afin de restituer correctement les doses reçues par les opérate...

Generating the flux map of Nigeria Research Reactor-1 for efficient ...

African Journals Online (AJOL)

One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

The prevalence of PAI-1 4G/5G gene variant in Serbian population

Directory of Open Access Journals (Sweden)

Đorđević Valentina

2013-01-01

Full Text Available Introduction: Plasminogen activator inhibitor 1 (PAI-1 has a major role in inhibition of firinolysis and normal haemostasis. The presence of the PAI-1 4G/4G genotype leads to increased expression of PAI-1. High blood level of PAI-1 is associated with many diseases such as thrombosis, cerebral insult, myocardial infarction, pregnancy loss, preeclampsia, insulin resistance, type 2 diabetes, breast cancer and asthma. In this study, the prevalence of PAI-1 4G/5G gene variant was determined in healthy subjects from Serbian population. Methods: The study was carried out in a group of 210 healthy subjects (105 women and 105 men. The presence of PAI-1 4G/5G gene variant was detected by PCR-RFLP analysis. Results: The prevalence of PAI-1 4G/4G genotype was 34.76% and it was increased compared to PAI-1 5G/5G genotype (19.05%. The most frequent was PAI-1 4G/5G genotype (46.19%. Allelic frequency for 4G allele was higher (0.58 compared to 5G allele (0.42. Conclusions: The prevalence of PAI-1 4G/5G gene variant in Serbian population is similar to the neighboring populations. Results of this study represent the first data for Serbian population. This study could be useful for further research where the role of PAI-1 4G/5G gene variant will be assessed in the pathogenesis of many diseases.

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1987-01-01

RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

The H,G_1,G_2 photometric system with scarce observational data

Science.gov (United States)

Penttilä, A.; Granvik, M.; Muinonen, K.; Wilkman, O.

2014-07-01

The H,G_1,G_2 photometric system was officially adopted at the IAU General Assembly in Beijing, 2012. The system replaced the H,G system from 1985. The 'photometric system' is a parametrized model V(α; params) for the magnitude-phase relation of small Solar System bodies, and the main purpose is to predict the magnitude at backscattering, H := V(0°), i.e., the (absolute) magnitude of the object. The original H,G system was designed using the best available data in 1985, but since then new observations have been made showing certain features, especially near backscattering, to which the H,G function has troubles adjusting to. The H,G_1,G_2 system was developed especially to address these issues [1]. With a sufficient number of high-accuracy observations and with a wide phase-angle coverage, the H,G_1,G_2 system performs well. However, with scarce low-accuracy data the system has troubles producing a reliable fit, as would any other three-parameter nonlinear function. Therefore, simultaneously with the H,G_1,G_2 system, a two-parameter version of the model, the H,G_{12} system, was introduced [1]. The two-parameter version ties the parameters G_1,G_2 into a single parameter G_{12} by a linear relation, and still uses the H,G_1,G_2 system in the background. This version dramatically improves the possibility to receive a reliable phase-curve fit to scarce data. The amount of observed small bodies is increasing all the time, and so is the need to produce estimates for the absolute magnitude/diameter/albedo and other size/composition related parameters. The lack of small-phase-angle observations is especially topical for near-Earth objects (NEOs). With these, even the two- parameter version faces problems. The previous procedure with the H,G system in such circumstances has been that the G-parameter has been fixed to some constant value, thus only fitting a single-parameter function. In conclusion, there is a definitive need for a reliable procedure to produce

Experimental measurement of fission fragments paths in uranium gold, molybdenum, zirconium and silicon; Mesure experimentale des parcours des fragments de fission dans l'uranium, l'or, le molybdene, le zirconium et le silicium

Energy Technology Data Exchange (ETDEWEB)

Faraggi, H; Garin-Bonnet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The measurement of total number of fissiongments emerging from an homogeneous, thick alloy composed of uranium plus another element (the concentration of uranium being known) allows to obtain the range of the fragments in this alloy. By varying the concentration, the range of the fragments in uranium and in the other element can be deduced. (author)Fren. [French] La mesure du nombre total de fragments de fission sortant d'un alliage homogene epais d'uranium et d'un autre element, pour lequel la concentration en uranium est donnee, permet la mesure du parcours des fragments dans cet alliage. En faisant varier la concentration, on peut deduire de ces mesures le parcours des fragments dans l'uranium et dans l'autre element. (auteur)

Incidência de fumonisina B1, aflatoxinas B1, B2, G1 e G2, ocratoxina A e zearalenona em produtos de milho Occurrence of fumonisin B1, aflatoxins B1, B2, G1, and G2, ochratoxin A and zearalenone in corn products

Directory of Open Access Journals (Sweden)

Luciane Mie Kawashima

2006-09-01

Full Text Available Levantamentos de ocorrência de micotoxinas em alimentos foram realizados nas últimas duas décadas nas regiões Sudeste e Sul do Brasil. Levantamentos em alimentos comercializados em outras regiões têm-se limitado a aflatoxinas em amendoim e castanhas do Brasil. O presente trabalho pesquisou a presença de fumonisina B1, aflatoxinas B1, B2, G1 e G2, ocratoxina A e zearalenona em 74 amostras de produtos a base de milho adquiridas no comércio da cidade de Recife, PE, durante o período de 1999 a 2001. Fumonisina B1 foi determinada por cromatografia líquida de alta eficiência com detecção por fluorescência e as demais toxinas foram determinadas por cromatografia em camada delgada. Fumonisina B1 foi encontrada em 94,6% das amostras em concentrações variando de 20 a 8600 µg/kg. Apenas 5 amostras continham aflatoxina B1 e o teor máximo encontrado foi 20 µg/kg. Duas amostras ultrapassaram o limite de 20 µg/kg para a somatória das aflatoxinas B1, B2, G1 e G2 (farinha de milho pré-cozida com 21,5 µg/kg e quirera (xerém com 23,3 µg/kg. As aflatoxinas G1 e G2, ocratoxina A e zearalenona não foram detectadas em nenhuma das amostras. Todas as amostras contaminadas com aflatoxinas também apresentaram fumonisina B1.Research concerning the presence of mycotoxin in food has been conducted in the Southwest and South regions of Brazil over the last two decades. Research in other regions has been limited to aflatoxin in peanuts and Brazil nuts. The aim of this work is to study the presence of fumonisin B1, aflatoxins B1, B2, G1, and G2, ochratoxin A and zearalenone in 74 samples of corn products acquired in shops and food markets in the city of Recife (PE from 1999 to 2001. Fumonisin B1 was determined by high performance liquid chromatography and fluorescence was detected. The other toxins were determined by thin layer chromatography. Fumonisin B1 was found in 94.6% of the samples in levels from 20 to 8600 µg/kg. Only 5 samples contained

Hythane (H2 and CH4) production from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals via up-flow anaerobic self-separation gases reactor

International Nuclear Information System (INIS)

Mahmoud, Mohamed; Elreedy, Ahmed; Pascal, Peu; Sophie, Le Roux; Tawfik, Ahmed

2017-01-01

Highlights: • Bio-hythane production from polyester wastewater via UASG reactor was assessed. • Impacts of influent contamination by 1,4-dioxane and heavy metals were discussed. • Maximum volumetric H 2 and CH 4 productions of 0.12 and 1.06 L/L/d were achieved. • Significant drop in CH 4 production was resulted at OLR up to 1.07 ± 0.06 gCOD/L/d. • Bioenergy recovery through UASG economically achieved a net profit of 10,231 $/y. - Abstract: A long-term evaluation of hythane generation from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals was investigated in a continuous up-flow anaerobic self- separation gases (UASG) reactor inoculated with mixed culture. The reactor was operated at constant hydraulic retention time (HRT) of 96 h and different organic loading rates (OLRs) of 0.31 ± 0.04, 0.71 ± 0.08 and 1.07 ± 0.06 gCOD/L/d. Available data showed that volumetric hythane production rate was substantially increased from 0.093 ± 0.021 to 0.245 ± 0.016 L/L/d at increasing OLR from 0.31 ± 0.04 to 0.71 ± 0.08 gCOD/L/d. However, at OLR exceeding 1.07 ± 0.06 gCOD/L/d, it was dropped to 0.114 ± 0.016 L/L/d. The reactor achieved 1,4-dioxane removal efficiencies of 51.8 ± 2.8, 35.9 ± 1.6 and 26.3 ± 1.6% at initial 1,4-dioxane concentrations of 1.14 ± 0.28, 1.97 ± 0.41 and 4.21 ± 0.30 mg/L, respectively. Moreover, the effect and potential removal of the contaminated by heavy metals (i.e., Cu 2+ , Mn 2+ , Cr 3+ , Fe 3+ and Ni 2+ ) were highlighted. Kinetic modelling and microbial community dynamics were studied, according to each OLR, to carefully describe the UASG performance. The economic analysis showed a stable operation for the anaerobic digestion of unsaturated polyester resin wastewater using UASG, and the maximum net profit was achieved at OLR of 0.71 ± 0.08 gCOD/L/d.

Comparison of some characteristics of aerobic granules and sludge flocs from sequencing batch reactors.

Science.gov (United States)

Li, J; Garny, K; Neu, T; He, M; Lindenblatt, C; Horn, H

2007-01-01

Physical, chemical and biological characteristics were investigated for aerobic granules and sludge flocs from three laboratory-scale sequencing batch reactors (SBRs). One reactor was operated as normal SBR (N-SBR) and two reactors were operated as granular SBRs (G-SBR1 and G-SBR2). G-SBR1 was inoculated with activated sludge and G-SBR2 with granules from the municipal wastewater plant in Garching (Germany). The following major parameters and functions were measured and compared between the three reactors: morphology, settling velocity, specific gravity (SG), sludge volume index (SVI), specific oxygen uptake rate (SOUR), distribution of the volume fraction of extracellular polymeric substances (EPS) and bacteria, organic carbon and nitrogen removal. Compared with sludge flocs, granular sludge had excellent settling properties, good solid-liquid separation, high biomass concentration, simultaneous nitrification and denitrification. Aerobic granular sludge does not have a higher microbial activity and there are some problems including higher effluent suspended solids, lower ratio of VSS/SS and no nitrification at the beginning of cultivation. Measurement with CLSM and additional image analysis showed that EPS glycoconjugates build one main fraction inside the granules. The aerobic granules from G-SBR1 prove to be heavier, smaller and have a higher microbial activity compared with G-SBR2. Furthermore, the granules were more compact, with lower SVI and less filamentous bacteria.

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

The heavy water reactors

International Nuclear Information System (INIS)

Brudermueller, G.

1976-01-01

This is a survey of the development so far of this reactor line which is in operation all over the world in various types (e.g. BHWR, PHWR). MZFR and the CANDU-type reactors are discussed in more detail. (UA) [de

PAI-1 4G/5G polymorphism contributes to cancer susceptibility: evidence from meta-analysis.

Science.gov (United States)

Wang, Shangqian; Cao, Qiang; Wang, Xiaoxiang; Li, Bingjie; Tang, Min; Yuan, Wanqing; Fang, Jianzheng; Qian, Jian; Qin, Chao; Zhang, Wei

2013-01-01

The plasminogen activator inhibitor-1 (PAI-1) is expressed in many cancer cell types and allows the modulation of cancer growth, invasion and angiogenesis. To date, studies investigated the association between a functional polymorphism in PAI-1 (4G/5G) and risk of cancer have shown inclusive results. A meta-analysis based on 25 case-control studies was performed to address this issue. Odds ratios (OR) with corresponding 95% confidence intervals (CIs) were used to assess the association. The statistical heterogeneity across studies was examined with I(2) test. Overall, a significant increased risk of cancer was associated with the PAI-1 4G/4G polymorphism for the allele contrast (4G vs. 5G: OR = 1.10, CI = 1.03-1.18, I(2) = 49.5%), the additive genetic model (4G/4G vs. 5G/5G: OR = 1.21, CI = 1.06-1.39, I(2) = 51.9%), the recessive genetic model (4G/4G vs. 4G/5G+5G/5G: OR = 1.11, CI = 1.04-1.18, I(2) = 20.8%). In the subgroup analysis by ethnicity, the results indicated that individuals with 4G/4G genotype had a significantly higher cancer risk among Caucasians (4G/4G vs. 5G/5G: OR = 1.31, 95%CI = 1.09-1.59, I(2) = 59.6%; 4G/4G vs. 4G/5G: OR = 1.12, 95%CI = 1.04-1.21, I(2) = 3.6%; recessive model: OR = 1.12, 95%CI = 1.05-1.21, I(2) = 25.3%). The results of the present meta-analysis support an association between the PAI-1 4G/5G polymorphism and increasing cancer risk, especially among Caucasians, and those with 4G allele have a high risk to develop colorectal cancer and endometrial cancer.

Brayton rotating units for space reactor power systems

Energy Technology Data Exchange (ETDEWEB)

Gallo, Bruno M.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies and Chemical and Nuclear Engineering Dept., The Univ. of New Mexico, Albuquerque, NM 87131 (United States)

2009-09-15

Designs and analyses models of centrifugal-flow compressor and radial-inflow turbine of 40.8kW{sub e} Brayton Rotating Units (BRUs) are developed for 15 and 40 g/mole He-Xe working fluids. Also presented are the performance results of a space power system with segmented, gas cooled fission reactor heat source and three Closed Brayton Cycle loops, each with a separate BRU. The calculated performance parameters of the BRUs and the reactor power system are for shaft rotational speed of 30-55 krpm, reactor thermal power of 120-471kW{sub th}, and turbine inlet temperature of 900-1149 K. With 40 g/mole He-Xe, a power system peak thermal efficiency of 26% is achieved at rotation speed of 45 krpm, compressor and turbine inlet temperatures of 400 and 1149 K and 0.93 MPa at exit of the compressor. The corresponding system electric power is 122.4kW{sub e}, working fluid flow rate is 1.85 kg/s and the pressure ratio and polytropic efficiency are 1.5% and 86.3% for the compressor and 1.42% and 94.1% for the turbine. For the same nominal electrical power of 122.4kW{sub e}, decreasing the molecular weight of the working fluid (15 g/mole) decreases its flow rate to 1.03 kg/s and increases the system pressure to 1.2 MPa. (author)

Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

International Nuclear Information System (INIS)

Kuruc, J.; Matel, L.

2007-01-01

The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

[PAL-1 5G/4G polymorphism in patients with systemic lupus erythematosus].

Science.gov (United States)

Savov, A; Andonova, S; Tanev, D; Robeva, R; Marincheva, Ts; Tomova, A; Kumanov, Ph; Rashkov, R; Kolarov, Zl

2014-01-01

Systemic lupus erythematosus (SLE) is a connective tissue disease affecting predominantly women that has been widely associated with obstetric complications. Inherited thrombophilias are significant risk factors for pregnancy loss, but their role in patients with SLE, and especially in those without concomitant secondary antiphospholipid syndrome (APS) has not been clarified. The aim of the present study was to study PAI-1 5G/4G polymorphism in women with lupus. A total of 103 SLE patients as well as 69 healthy volunteers were genotyped for PAI-1 5G/4G (rs1799889). No significant differences in the PAI-1 5G/4G genotype prevalence between patients and controls were found. After exclusion of the women with secondary APS, the frequency of pregnancies and spontaneous abortions, as well as the number of live births were similar in the studied patients with different PAI-1 genotype (p> 0.05). PAI-1 5G/4G polymorphism was not significantly related to any of the lupus ACR criteria or disease activity (p > 0.05), but it could influence the platelet number in the studied patients (263.52 ± 91.10 [5G/5G genotype] versus 210.12 ± 71.79 [4G/4G genotype], p = 0.023). In conclusion, our results showed that PAI-1 4G/5G polymorphism did not worsen the reproductive outcome in SLE women without secondary APS.

Demolition of the FRJ-1 research reactor (MERLIN)

International Nuclear Information System (INIS)

Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2003-01-01

FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

Evaporation rate measurement in the pool of IEAR-1 reactor

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

2000-01-01

The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

Energy Technology Data Exchange (ETDEWEB)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

PAI-1 mRNA expression and plasma level in rheumatoid arthritis: relationship with 4G/5G PAI-1 polymorphism.

Science.gov (United States)

Muñoz-Valle, José Francisco; Ruiz-Quezada, Sandra Luz; Oregón-Romero, Edith; Navarro-Hernández, Rosa Elena; Castañeda-Saucedo, Eduardo; De la Cruz-Mosso, Ulises; Illades-Aguiar, Berenice; Leyva-Vázquez, Marco Antonio; Castro-Alarcón, Natividad; Parra-Rojas, Isela

2012-12-01

Rheumatoid arthritis (RA) is a chronic inflammatory disease affecting the synovial membrane, cartilage and bone. PAI-1 is a key regulator of the fibrinolytic system through which plasminogen is converted to plasmin. The plasmin activates the matrix metalloproteinase system, which is closely related with the joint damage and bone destruction in RA. The aim of this study was to investigate the relationship between 4G/5G PAI-1 polymorphism with mRNA expression and PAI-1 plasma protein levels in RA patients. 113 RA patients and 123 healthy subjects (HS) were included in the study. The 4G/5G PAI-1 polymorphism was determined by polymerase chain reaction-restriction fragment length polymorphism method; the PAI-1 mRNA expression was determined by real-time PCR; and the soluble PAI-1 (sPAI-1) levels were quantified using an ELISA kit. No significant differences in the genotype and allele frequencies of 4G/5G PAI-1 polymorphism were found between RA patients and HS. However, the 5G/5G genotype was the most frequent in both studied groups: RA (42%) and HS (44%). PAI-1 mRNA expression was slightly increased (0.67 fold) in RA patients with respect to HS (P = 0.0001). In addition, in RA patients, the 4G/4G genotype carriers showed increased PAI-1 mRNA expression (3.82 fold) versus 4G/5G and 5G/5G genotypes (P = 0.0001), whereas the sPAI-1 plasma levels did not show significant differences. Our results indicate that the 4G/5G PAI-1 polymorphism is not a marker of susceptibility in the Western Mexico. However, the 4G/4G genotype is associated with high PAI-1 mRNA expression but not with the sPAI-1 levels in RA patients.

The measurement of g{sub 1}{sup n} polarized structure of the neutron by the E154 experiment at SLAC; Mesure de la fonction de structure polarisee g{sub 1}{sup n} du neutron par l`experience E154 au SLAC

Energy Technology Data Exchange (ETDEWEB)

Incerti, Sebastien [Ecole Doctorale des Sciences Fondamentales, Universite Blaise Pascal, U.F.R. de Recherche Scientifique et Technique, F-63177 Aubiere (France)

1998-01-21

This thesis presents the precision measurement of the neutron polarized structure g{sub 1}{sup n} performed by the E154 collaboration at the Standford Linear Accelerator Center, USA, in autumn 1995, using a 48.3 GeV polarized electron beam scattered off a polarized Helium 3 target. The scattered electrons were detected using two spectrometer arms, covering the deep inelastic scattering range: 0.0134 < x < 0.7 and 1 GeV{sup 2} < Q{sup 2} < 17 GeV{sup 2} at an average value of Q{sup 2} = 5 GeV{sup 2}. Two electromagnetic calorimeters have been designed by the LPC in Clermont-Ferrand and the SphN-CEA in Saclay to measure the scattered electron energy and to eject the contaminating hadron background using, a cellular automaton and a neural network, widely described in this thesis. The analysis performed in Clermont-Ferrand and presented in this document led us to the integral on the measurement region of g{sub 1}{sup n} equaling: - 0.34 {+-} 0.003{sub STAT} {+-} 0.004{sub SYST} {+-} 0.001{sub EVOL} at Q{sup 2} = 5 GeV{sup 2}, where our data have been evolved to Q{sup 2} = 5 GeV{sup 2} using the next-to-leading order DGLAP evolution equations and a world parametrization of the polarized parton distributions. The Ellis and Jaffe sum rule is clearly violated. Using different low x extrapolations, our integral is compatible with the Bjorken sum rule. The quark contribution to the nucleon spin is {Delta}{Sigma} = 29 {+-} 6 % in the M S-bar scheme and {Delta}{Sigma} = 37 {+-} 7% in the AB scheme, at Q{sup 2} = 5 GeV{sup 2}. The gluon contribution seems to be positive and within the range: 0 < {Delta}G < 2. (author) 245 refs., 170 figs., 60 tabs.

Thermal flux flattering and increase of reactor output

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

It is worthwhile, when building power reactors, to have excess reactivity in order to increase rating by fitting closely together the heat sources and the cooling possibilities. The power per unit volume of a graphite reactor can then be increased, given the power of the most heavily loaded channel. The solutions adopted for G.1, G.2, and E.D.F.1 are described here, and also the improvements based on the actual neutron flux flattening, the introduction of several zones for the coolant, the variation of uranium rod and coolant channel diameters according to their location, and finally the change in lattice pitch. The perturbation of neutron flux due to variation of mean absorption in the lattice is also discussed. (author)

Temperature measurements in thermonuclear plasmas; Mesures des temperatures dans les plasmas thermonucleaires

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The temperatures needed to produce thermonuclear reactions are of the order of several million degrees Kelvin. Devising methods for measuring such temperatures has been the subject of research in many countries. In order to present the problem clearly and to demonstrate its importance, the author reviews the various conditions which must be fulfilled in order that reactions may be qualified as thermonuclear. The relationship between the temperature and the cross-section of the reactions is studied, and it is shown that the notion of temperature in the plasmas is complex, which leads to a consideration of the temperature of the ions and that of the electrons. None of the methods for the temperature measurements is completely satisfactory because of the hypotheses which must be made, and which are seldom fulfilled during high-intensity discharges in the plasmas. In practice it is necessary to use several methods simultaneously. (author) [French] Les temperatures necessaires pour produire des reactions thermonucleaires sont de l'ordre de plusieurs millions de degres Kelvin. Les methodes envisagees pour mesurer ces temperatures font l'objet de recherches dans de nombreux pays. Afin de preciser le probleme et de montrer son importance, l'auteur rappelle les conditions qui doivent etre reunies pour que des reactions puissent etre qualifiees thermonucleaires. Il etudie la relation entre la temperature et la section efficace des reactions et montre que la notion de temperature dans les plasmas est complexe, ce qui amene a considerer la temperature des ions et celle des electrons. Aucune des methodes de mesure des temperatures n'est completement satisfaisante en raison des hypotheses qu'elles exigent et qui sont rarement realisees lors des decharges a haute intensite dans les plasmas. En pratique, il est necessaire d'utiliser plusieurs methodes simultanement. (auteur)

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

Coincidence measurements with the use of detectors measuring the energy of the radiances (proportional meters and scintillation counter); Mesures de coincidences avec utilisation de detecteurs mesurant l'energie des rayonnements (compteurs proportionnels et compteur a scintillations)

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

In the setting of the realization of a set of installations permitting of the measures of coincidences between sorted radiances according to their energies, an installation understanding a proportional counter and a scintillation counter has been constructed and optimized. It has been used to do some measures of coincidences between X{sub K} photons and photons {gamma} issued at the time of the radioactive transformation of the selenium 75 (electronic capture). The efficiency of the proportional meter has been determined roughly. Besides, a proportional counter of solid angle neighboring of 4{pi} was able to achieve measures of coincidences while only doing one selection of amplitudes: indeed, the simultaneity of the detection of two radiances appear by an impulse whose amplitude is the sum of the amplitudes of the impulses resulting from each of the studied radiations. This method, applied to the coincidences between X-rays, permitted to bring the information on the diagram of decay of the arsenic 73. Besides, the coefficient of internal conversion of a consecutive transition to this decay has been valued. (author) [French] Dans le cadre de la realisation d'une serie de montages permettant des mesures de coincidences entre rayonnements tries d'apres leurs energies, un montage comprenant un compteur proportionnel et un compteur a scintillations a ete construit et mis au point. Il a ete utilise pour effectuer quelques mesures de coincidences entre photons X{sub K} et photons {gamma} emis lors de la transformation radioactive du selenium 75 (capture electronique). L'efficacite du compteur proportionnel a ete approximativement determinee. De plus, un compteur proportionnel d'angle solide voisin de 4{pi} a pu etre utilise pour realiser des mesures de coincidences en n'effectuant qu'une selection d'amplitudes: en effet, la simultaneite de la detection de deux rayonnements se manifeste par une impulsion dont l'amplitude est la somme des amplitudes des impulsions

The pressure of hot QCD up to $g^{6}$ ln(1/g)

CERN Document Server

Kajantie, Keijou; Rummukainen, K; Schröder, Y

2003-01-01

The free energy density, or pressure, of QCD has at high temperatures an expansion in the coupling constant g, known so far up to order g^5. We compute here the last contribution which can be determined perturbatively, g^6 ln(1/g), by summing together results for the 4-loop vacuum energy densities of two different three-dimensional effective field theories. We also demonstrate that the inclusion of the new perturbative g^6 ln(1/g) terms, together with the so far unknown perturbative and non-perturbative g^6 terms, could potentially extend the applicability of the resummed coupling constant series down to surprisingly low temperatures.

Determination of one spectral index at the argonaut reactor

International Nuclear Information System (INIS)

Klawa, R.

1973-01-01

One spectral index at the Argonauta Reactor was determined. The Westcott formalism was employed assuming two components: Maxwellian and 1/E. The values of g(T) and s(T) were obtained from the Westcott definitions by means of the Breit - Wigner formula for the cross section. The r and T were determined for one point at the core of Argonauta Reactor. (author)

Theoretical studies aiming at the IEA-R1 reactor core conversion from high U-235 enrichment to low U-235 enrichment

International Nuclear Information System (INIS)

Frajndlich, R.

1982-01-01

The research reactors, of which the fuel elements are of MTR type, functions presently, almost in their majority with high U-235 enrichment. The fear that those fuel elements might generate a considerabLe proliferation of nuclear weapons rendered almost mandatory the conversion of highly enriched fuel elements to a low U-235 enrichment. As the IEA-R1 reactor of IPEN is operating with highly enriched fuel elements a study aiming at this conversion was done. The problems related to the conversion and the results obtained, demonstrated the technical viabilty for its realization. (E.G.) [pt

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

RA reactor operation and maintenance in 1992, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

1992-01-01

During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

26 CFR 1.514(g)-1 - Business lease indebtedness.

Science.gov (United States)

2010-04-01

....514(g)-1 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES (CONTINUED) Taxation of Business Income of Certain Exempt Organizations § 1.514(g)-1... to cases such as the following: Example 1. A university pledges some of its investment securities...

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

Spent fuel management - two alternatives at the FiR 1 reactor

International Nuclear Information System (INIS)

Salmenhaara, S.E.J.

2001-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

Quantitative Analysis of Microbes in Water Tank of G.A. Siwabessy Reactor

International Nuclear Information System (INIS)

Itjeu Karliana; Diah Dwiana Lestiani

2003-01-01

The quality of water in reactor system has an important role because it could effect the function as a coolant and the operation of reactor indirectly. The study of microbe analyzes has been carried out to detect the existence of microbes in water tank and quantitative analyzes of microbes also has been applied as a continuation of the previous study. The samples is taken out from the end side of reactor GA Siwabessy's tank, inoculated in TSA (Tripcase Soy Agar) medium, put in incubator at 30 - 35 o C for 4 days. The results of experiment show the reconfirmation for the existence of bacteria and the un-existence of yield. The quantitative analysis with TPC method show the growth rate of bacteria is twice in 24 hours. (author)

Computer-aided testing and operational aids for PARR-1 nuclear reactor

International Nuclear Information System (INIS)

Ansari, S.A.

1990-01-01

The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

Properties of autoregressive model in reactor noise analysis, 1

International Nuclear Information System (INIS)

Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

1987-01-01

Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

Measurements of spectral indices in homogeneous multiplying media; Mesures d'indices de spectre dans les milieux multiplicateurs homogenes

Energy Technology Data Exchange (ETDEWEB)

Bruna, J G; Brunet, J P; Clouet D' Orval, Ch; Verriere, Ph; Kremser, J; Moret-Bailly, J; Tellier, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Methods for computation of spectra in light water are developed at Saclay and it is interesting to carry out at the same time experimental studies of simple media such as solutions of fissionable salts which allow quite direct comparisons with computed values. The spectral indices measurements were made with two small fission chambers, one containing deposited plutonium, the other deposited uranium 235. Their response, when neutron spectrum is modified, allows to study the epithermal part of the flux. The media studied with these chambers are fissionable solutions (of plutonium or 90 per cent enriched uranium) which were made critical in bare cylindrical geometry in the Alecto reactor. If the ratio of the chambers is normalized to unity in a Maxwell spectrum, then the noted variation of the ratio of the counts Pu chamber/ U{sup 235} chamber reaches 1,4 in the range of the studied concentrations. (authors) [French] Des calculs de spectres dans l'eau legere sont mis au point a Saclay et il est interessant de mener parallelement des etudes experimentale sur des milieux simples tels que des solutions de sels fissiles, qui permettent des comparaisons tres directes avec les valeurs calculees. On a choisi d'effectuer des mesures d' 'indices de spectres' a l'aide de de deux petites chambres a fission contenant des depots, l'une de plutonium, l'autre d'uranium 235. Leur reponse lorsque le spectre des neutrons est modifie permet d'etudier la partie epithermique du flux. Les milieux etudies a l'aide de ces chambres sont des solutions fissiles (plutonium ou uranium enrichi a 90 pour cent) rendus critiques, en geometrie cylindrique nue, dans le reacteur Alecto. Si le rapport des chambres est normalise a un dans un spectre de Maxwell, la variation constatee du rapport des comptages chambre Pu/ chambre U{sup 235} atteint, dans les gammes de concentrations etudiees, 1,4. (auteurs)

Peptidolytic microbial community of methanogenic reactors from two modified UASBs of brewery industries

Directory of Open Access Journals (Sweden)

C. Díaz

2010-10-01

Full Text Available We studied the peptide-degrading anaerobic communities of methanogenic reactors from two mesophilic full-scale modified upflow anaerobic sludge blanket (UASB reactors treating brewery wastewater in Colombia. Most probable number (MPN counts varied between 7.1 x 10(8 and 6.6 x 10(9 bacteria/g volatile suspended solids VSS (Methanogenic Reactor 1 and 7.2 x 10(6 and 6.4 x 10(7 bacteria/g (VSS (Methanogenic Reactor 2. Metabolites detected in the highest positive MPN dilutions in both reactors were mostly acetate, propionate, isovalerate and, in some cases, negligible concentrations of butyrate. Using the highest positive dilutions of MPN counts, 50 dominant strains were isolated from both reactors, and 12 strains were selected for sequencing their 16S rRNA gene based on their phenotypic characteristics. The small-subunit rRNA gene sequences indicated that these strains were affiliated to the families Propionibacteriaceae, Clostridiaceae and Syntrophomonadaceae in the low G + C gram-positive group and Desulfovibrio spp. in the class d-Proteobacteria. The main metabolites detected in the highest positive dilutions of MPN and the presence of Syntrophomonadaceae indicate the effect of the syntrophic associations on the bioconversion of these substrates in methanogenic reactors. Additionally, the potential utilization of external electron acceptors for the complete degradation of amino acids by Clostridium strains confirms the relevance of these acceptors in the transformation of peptides and amino acids in these systems.

Reactor technology

International Nuclear Information System (INIS)

Erdoes, P.

1977-01-01

This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

Dependence of neutron rate production with accelerator beam profile and energy range in an ADS-TRIGA RC1 reactor

Energy Technology Data Exchange (ETDEWEB)

Firoozabadi, M.M.; Karimi, J. [Birjand Univ. (Iran, Islamic Republic of). Dept. of Physics; Zangian, M. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Nuclear Engineering Dept.

2016-12-15

Lead, mercury, tantalum and tungsten were used as target material for calculation of spallation processes in an ADS-TRIGA RC1 reactor. The results show that tungsten has the highest neutron production rate. Therefore it was selected as target material for further calculations. The sensitivity of neutron parameters of the ADS reactor core relative to a change of beam profile and proton energy was determined. The core assembly and parameters of the TRIGA RC1 demonstration facility were used for the calculation model. By changing the proton energy from 115 to 1 400 MeV by using the intra-nuclear cascade model of Bertini (INC-Bertini), the quantity of the relative difference in % for energy gain (G) and spallation neutron yield (Y{sub n/p}), increases to 289.99 % and 5199.15 % respectively. These changes also reduce the amount of relative difference for the proton beam current (I{sub p}) and accelerator power (P{sub acc}), 99.81 % and 81.28 % respectively. In addition, the use of a Gaussian distribution instead of a uniform distribution in the accelerator beam profile increases the quantity of relative difference for energy gain (G), net neutron multiplication (M) and spallation neutron yield (Y{sub n/p}), up to 4.93 %, 4.9 % and 5.55 % respectively.

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

Spent fuel management plans for the FiR 1 Reactor

International Nuclear Information System (INIS)

Salmenhaara, S. E. J.

2002-01-01

The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

International Nuclear Information System (INIS)

Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

2001-01-01

Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)