Que peut-on voir avec des neutrons? Une introduction pour des non spécialistes

Science.gov (United States)

Schweizer, J.

2005-11-01

Le neutron est une particule élémentaire qui a été découverte en 1932 par James Chadwick. Ses caractéristiques principales sont résumées dans le tableau I. Il a été utilisé pour la première fois par Clifford Shull en 1946 comme outil pour des expériences de diffusion. Cette technique s'est depuis constamment développée pour concerner tous les aspects de la matière condensée: physique, chimie, matériaux, biologie. Il s'agit d'un outil tout à fait exceptionnel car le neutron possède des propriétés uniques et particulièrement adaptées pour ces études.

Propagation of thermal neutrons in mock-up screw-shaped steel elements with water protection; Propagation des neutrons thermiques dans des fausses cartouches d'acier en helice dans une protection d'eau. Programme tournesol 3

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Devillers, C L; Lanore, J M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-07-01

This report treats the streaming of thermal neutrons in a cylindrical duct in light water. The duct contains a spiral iron shield. Transmission and reflection matrices are used to describe the probabilities for the thermal neutrons to be absorbed or to be scattered on the surfaces. The neutron paths across the void are represented by geometrical matrices. The numerical resolution is performed by the Monte-Carlo method. (authors) [French] Dans ce rapport on traite un probleme de fuites de neutrons thermiques dans un canal cylindrique plonge dans l'eau et obture par un ecran helicoidal en acier. On utilise des matrices de transmission-reflexion pour decrire les probabilites d'absorption et de diffusion des neutrons sur les parois et l'helicoide et des matrices de correspondance geometrique pour representer la propagation dans le vide. La resolution numerique se fait par une methode de Monte-Carlo. (auteur)

Diffusion of graphite. The effect of cylindrical canals; Longueur de diffusion du graphite effet des canaux cylindriques

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Carle, R; Clouet d' Orval, C; Martelly, J; Mazancourt, T de; Sagot, M; Lattes, R; Teste du Bailler, A [Commissariat a l' Energie Atomique, Dir. Industrielle, Saclay (France). Centre d' Etudes Nucleaires; Robert, C [Ecole Normale Superieure, 75 - Paris (France)

1957-07-01

Experiments on thermal neutron diffusion in the graphite used as moderator in the pile G1 have been carried out. The object of these experiments is to determine: - the intrinsic quality of this graphite, characterised by its diffusion length L or its Laplacian 1/L{sup 2} - the effect of the canals, which modifies anisotropically the macroscopic diffusion equation and is characterized by two principal diffusion regions (or two principal Laplacian), valid respectively for the diffusion in the direction of the canals and in a perpendicular direction. In order to determine them two experiments are necessary, in which the second derivatives of the flux in relation to the space coordinates are very different. These experiments form the object of the first two parts. Part 1: Diffusion along the axis of a flux coming from the pile source, and limited radially by a quasi cylindrical screen of cadmium bars. This screen, or Faraday cage is designed to give to the thermal flux produced the same radius of extrapolation to zero as that of the pile source. The determination of L (with the graphite full) has been made under the same conditions. The measurements have been interpreted in two ways. The influence of the brackets holding the detectors is discussed. Part 2: Radial diffusion in the graphite surrounding the 'long' cylindrical pile. This is well described by a sum of Bessel functions. Part 3: Results (valid for d = 1.61 t = 17 deg. C). For the graphite without cavity L = 52.7 {+-} 0.4 cm. The effect of the canals on the diffusion area and its anisotropy are in excellent agreement with the theory of Behrens: L(parallel) = 64.6 cm and L(perpendicular) 62.2 cm. Appendix: Theory of the Faraday cage. (author) [French] Des experiences de diffusion des neutrons thermiques dans le graphite constituant le moderateur de la pile G1 ont ete effectuees. Elles ont pour objet de determiner: - la qualite intrinseque de ce graphite, caracterisee par sa longueur de diffusion L ou son

The Scattering of Slow Neutrons from Hydrogen and Ethylene; Diffusion des neutrons lents par l'hydrogene et par l'ethylene; O rasseyanii medlennykh nejtronov vodorodom i ehtilenom; Dispersion de neutrones lentos por hidrogeno y etileno

Energy Technology Data Exchange (ETDEWEB)

Bally, D; Tarina, V; Todireanu, S [Institute of Atomic Physics, Bucharest (Romania)

1963-01-15

The angular distribution of 1.61 A wavelength neutrons scattered from hydrogen and ethylene has been measured. For ethylene the pressure was varied from 15 to 60 atm and for hydrogen from 60 to 110 atm. The investigated angular range was between 5''o and 90{sup o}. The experimental results obtained have been compared with the differential cross-section calculated using the Krieger-Nelkin formulae. (author) [French] Les auteurs ont mesure la distribution angulaire des neutrons ayant une longueur d'onde de 1,61 A diffuses par de l'hydrogene et par l'ethylene. Ils ont fait varier la pression de l'ethylene de 15 a 60 atm et celle de l'hydrogene de 60 a 110 atm. Le domaine angulaire etudie etait compris entre 5{sup o} et 90{sup o}. Les resultats des experiences ont ete compares aux sections efficaces differentielles calculees au moyen des formules de Krieger-Nelkin. (author) [Spanish] Los autores han medido la distribucion angular de neutrones de longitud de onda igual a 1,61 A, despersados por hidrogeno y etileno. En el caso del etileno, la presion vario de 15 a 60 atm y en el del higrogeno, de 60 a 110 atm. Los autores han explorado angulos comprendidos entre 5{sup o} y 90{sup o}. comparan los resultados experimentales obtenidos con el valor de la seccion eficaz diferencial calculado con ayuda de las formulas de Kriger-Nelkin. (author)

Measure of the slowing down area of thermal state 1,45 eV neutrons in the graphite; Mesure de l'aire de ralentissement des neutrons dans le graphite de 1,45 eV a l'etat thermique

Energy Technology Data Exchange (ETDEWEB)

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

1956-12-15

The life of a neutron emitted as a fast one within a moderating medium may be divided into two parts; at first it is slowed down colliding with the nuclei of the medium; later on, as soon as its energy is of the order of the thermal agitation energy, it diffuses until captured by some nucleus. As a matter of fact, dividing the evolution of a neutron into two phases, a slowing down phase and a diffusing phase is arbitrary, since the thermal equilibrium is attained stepwise. This report is intended to examine the means of inferring the distribution of thermal neutrons from the distribution of fast neutrons on the basis of the distribution of nascent thermal neutrons and of the law of scattering. (author) [French] La vie d'un neutron emis a l'etat rapide dans un milieu moderateur comprend 2 parties: au debut il se ralentit par chocs sur les noyaux du milieu; ensuite, lorsque son energie est de l'ordre de l'energie d'agitation thermique, il diffuse jusqu'a sa capture par un noyau. La separation de l'evolution d'un neutron en deux phases, phase de ralentissement, puis phase de diffusion, est, en realite, arbitraire; un neutron doit atteindre l'equilibre thermique progressivement. L'objet de ce rapport est de voir comment on peut deduire la repartition des thermiques de la repartition des rapides connaissant la repartition des thermiques naissants et la loi de diffusion. (au0010te.

Study by {alpha}-particle scattering, of the collective states of closed shell nuclei with 28 neutrons, 28 and 50 protons; Etude des etats a caractere collectif des noyaux au voisinage des couches fermees par diffusion inelastique des particules {alpha}

Energy Technology Data Exchange (ETDEWEB)

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

1967-01-01

This report gives the results from a systematic study by 44 MeV {alpha} particle scattering on the collective states of closed shell nuclei with 28 neutrons, 28 and 50 protons especially in the energy region corresponding to the two phonon vibrational triplet. The nuclei studied are: {sup 46}Ti, {sup 48}Ti, {sup 50}Ti, {sup 52}Cr, {sup 54}Fe, {sup 56}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 62}Ni, {sup 64}Ni, {sup 112}Sn, {sup 114}Sn, {sup 116}Sn, {sup 118}Sn, {sup 120}Sn, {sup 122}Sn and {sup 124}Sn. The theoretical analysis has been made with the Austern and Blair model using a phase shift analysis of the elastic scattering. Deformation lengths {delta}{sub I} = {beta}{sub I}R and transition probabilities obtained are compared with several experimental and theoretical data. (author) [French] Ce rapport donne les resultats d'une etude systematique par diffusion de particules {alpha} de 44 MeV des etats collectifs des noyaux au voisinage des couches fermees a 28 neutrons et 28 et 50 protons, specialement dans la region correspondant au triplet vibrationnel a deux phonons. Les noyaux etudies sont: {sup 46}Ti, {sup 48}Ti, {sup 50}Ti, {sup 52}Cr, {sup 54}Fe, {sup 56}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 62}Ni, {sup 64}Ni, {sup 112}Sn, {sup 114}Sn, {sup 116}Sn, {sup 118}Sn, {sup 120}Sn, {sup 122}Sn et {sup 124}Sn. L'analyse theorique a ete faite a l'aide du modele de Austern et Blair utilisant une analyse en dephasages de la diffusion elastique. Les longueurs de deformation {delta}I = {beta}{sub I}R obtenues pour chaque niveau sont comparees a plusieurs donnees experimentales et theoriques. (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)

Measurement of the diffusion length of thermal neutrons inside graphite; Mesure de la longueur de diffusion des neutrons thermiques dans le graphite

Energy Technology Data Exchange (ETDEWEB)

Ertaud, A; Beauge, R; Fauquez, H; De Laboulay, H; Mercier, C; Vautrey, L

1948-11-01

The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra {alpha} {yields} Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm {+-} 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

Application des techniques de diffusion de la lumière des rayons x et des neutrons à l'étude des systèmes colloïdaux. Deuxième partie : étude des différents systèmes : polymères en solution à l'état solide, solutions micellaires, systèmes fractals Application of Light, X-Ray and Neutron Diffusion Techniques to the Study of Colloidal Systems. Part Two: Research on Different Systems: Polymers in Solution in the Solid State, Micellar Solutions, Fractals Systems

Directory of Open Access Journals (Sweden)

Espinat D.

2006-11-01

Full Text Available Cet article fait suite à la première partie (Revue Inst. Franç. du Pétrole, vol. 45, n°6, novembre-décembre 1990 concernant l'application des techniques de diffusion de la lumière, des rayons X et des neutrons à l'étude des systèmes colloïdaux et plus précisément à la présentation théorique des trois méthodes. L'objet de cette deuxième partie est la présentation non exhaustive de quelques domaines d'applications. Nous nous attacherons tout particulièrement à présenter les potentialités des méthodes pour la caractérisation de systèmes colloïdaux ou divisés rencontrés dans de nombreuses branches d'activité de l'industrie pétrolière. Nous aborderons dans une première partie les solutions polymériques et colloïdales. En particulier nous nous attarderons sur l'importance des techniques de diffusion pour la caractérisation des polymères en solution et des solutions micellaires. Nous verrons également quelles informations la diffusion centrale peut apporter sur la macrostructure des polymères cristallisés ou amorphes à l'état solide. De nombreux systèmes présentent une structure de type fractal ; après présentation de quelques exemples, nous montrerons que les méthodes de diffusion peuvent apporter certaines informations sur ces matériaux, notamment la dimension fractale. This article is the second one (the first one was published in Revue de l'Institut Français du Pétrole No. 6, NovemberDecember 1990 concerning the application of techniques of light scattering, X rays and neutrons to the analysis of colloidal systems and more specifically to the theoretical description of the three methods. The aim of this second part is to make a nonexhaustive description of several fields of applications. A special effort is made to describe the potential of these methods for characterizing colloidal or divided systems encountered in a great many activities involving the petroleum industry. The first part of this

Theory of the diffusion coefficient of neutrons in a lattice containing cavities; Theorie du coefficient de diffusion des neutrons dans un reseau comportant des cavites

Energy Technology Data Exchange (ETDEWEB)

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

1964-01-15

In an previous publication, a simple and general formulation of the diffusion coefficient, which defines the mode of weighting of the mean free paths of the various media, in introducing the collision probabilities in each medium, was established. This expression is demonstrated again here through a more direct method, and the velocity is introduced; new terms are emphasised, the existence of which implies that the representation of the diffusion area as the mean square of the straight line distance from source to absorption is not correct in a lattice. However these terms are of small enough an order of magnitude to he treated as a correction. The general expression also shows the existence, for the radial coefficient, of the series of angular correlation terms, which is seen to converge very slowly for large channels. The term by term computation which was initiated in the first work was then interrupted and a global formulation, which emphasize a resemblance with the problem of the thermal utilisation factor, was adopted. An integral method, analogous to that use for the computation of this factor, gives the possibility to establish new and simple practical formulae, which require the use of a few basic functions only. These formulae are very accurate, as seen from the results of a variational method which was studied as a reference. Various correction effects are reviewed. Expressions which allow the exact treatment of fuel rod clusters are presented. The theory is confronted with various experimental results, and a new method of measuring the radial coefficient is proposed. (author) [French] Dans une publication anterieure, on a etablie une formulation simple et generale du coefficient de diffusion, qui definit le mode de ponderation des libres parcours des differents milieux constituants en faisant apparaitre les probabilites de collision dans chaque milieu. On redemontre ici cette expression d'une maniere plus directe, tout en introduisant la variable

Angular distributions of fast neutrons scattered by Al, Si, P, S and Zn; Distributions angulaires des neutrons rapides diffuses par Al, Si, P, S et Zn; Usloviya raspredeleniya bystrykh nejtronov, rasseyannykh alyuminiem, kremniem, fosforom i tsinkom; Distribuciones angulares de neutrones rapidos dispersados por Al, Si, P, S y Zn

Energy Technology Data Exchange (ETDEWEB)

Tstjkaija, K; Tanaka, S; Maeuyama, M; Tomita, Y [Japan Atomic Energy Research Institute, Tokai-Mtjea (Japan)

1962-03-15

Differential scattering cross-sections of Al, Si, P, S and Zn for fast neutrons have been measured in an energy range of 3.4 to 4.6 MeV by using the time-of-flight method. Angular distributions of the inelastically scattered neutrons are nearly isotropic in all cases. These results are discussed on the basis of the Hauser-Feshbach theory. (author) [French] Les sections efficaces differentielles de diffusion de Al, Si, P, S et Zn pour des neutrons rapides ont ete mesurees dans la gamme d'energies de 3,4 a 4,6 MeV, en employant la methode du temps de vol. Les distributions angulaires des neutrons diffuses inelastiquement sont presque isotropes dans tous les cas. Les auteurs analysent ces resultats en se fondant sur la theorie de Hauser-Feshbach. (author) [Spanish] Los autores han medido por el metodo del tiempo de vuelo las secciones eficaces diferenciales de dispersion del Al, Si, P, S y Zn para neutrones rapidos de energia comprendida entre 3,4 y 4,6 MeV. Las distribuciones angulares de los neutrones dispersados inelasticamente son casi isotropicas en todos los casos. Los autores analizan los resultados obtenidos basandose en la teoria de Hauser-Feshbach . (author) [Russian] Differentsial'no e sechenie rasseyaniya alyuminiya, kremniya, fosfora, sery i tsinka dlya bystrykh nejtronov izmereno v diapazone ehnergii ot 3,4 do 4,6 Megaehlektronvol't ispol'zovanie m metoda vremeni proleta. Uglovye raspredeleniya neuprugo rasseyannykh nejtronov yavlyayutsya pochti vo vsekh sluchayakh izotropnymi. Jeti rezul'taty obsuzhdayutsya na osnove teorii Hauzera-Feshbakha. (author)

Theory of the diffusion coefficient of neutrons in a lattice containing cavities; Theorie du coefficient de diffusion des neutrons dans un reseau comportant des cavites

Energy Technology Data Exchange (ETDEWEB)

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

1964-01-15

In an previous publication, a simple and general formulation of the diffusion coefficient, which defines the mode of weighting of the mean free paths of the various media, in introducing the collision probabilities in each medium, was established. This expression is demonstrated again here through a more direct method, and the velocity is introduced; new terms are emphasised, the existence of which implies that the representation of the diffusion area as the mean square of the straight line distance from source to absorption is not correct in a lattice. However these terms are of small enough an order of magnitude to he treated as a correction. The general expression also shows the existence, for the radial coefficient, of the series of angular correlation terms, which is seen to converge very slowly for large channels. The term by term computation which was initiated in the first work was then interrupted and a global formulation, which emphasize a resemblance with the problem of the thermal utilisation factor, was adopted. An integral method, analogous to that use for the computation of this factor, gives the possibility to establish new and simple practical formulae, which require the use of a few basic functions only. These formulae are very accurate, as seen from the results of a variational method which was studied as a reference. Various correction effects are reviewed. Expressions which allow the exact treatment of fuel rod clusters are presented. The theory is confronted with various experimental results, and a new method of measuring the radial coefficient is proposed. (author) [French] Dans une publication anterieure, on a etablie une formulation simple et generale du coefficient de diffusion, qui definit le mode de ponderation des libres parcours des differents milieux constituants en faisant apparaitre les probabilites de collision dans chaque milieu. On redemontre ici cette expression d'une maniere plus directe, tout en introduisant la variable

Vibrations et relaxations dans les molécules biologiques. Apports de la diffusion incohérente inélastique de neutrons

Science.gov (United States)

Zanotti, J.-M.

2005-11-01

Le présent document ne se veut pas un article de revue mais plutôt un élément d'initiation à une technique encore marginale en Biologie. Le lecteur est supposé être un non spécialiste de la diffusion de neutrons poursuivant une thématique à connotation biologique ou biophysique mettant en jeu des phénomènes dynamiques. En raison de la forte section de diffusion incohérente de l'atome d'hydrogène et de l'abondance de cet élément dans les protéines, la diffusion incohérente inélastique de neutrons est une technique irremplaçable pour sonder la dynamique interne des macromolécules biologiques. Après un rappel succinct des éléments théoriques de base, nous décrivons le fonctionnement de différents types de spectromètres inélastiques par temps de vol sur source continue ou pulsée et discutons leurs mérites respectifs. Les deux alternatives utilisées pour décrire la dynamique des protéines sont abordées: (i)l'une en termes de physique statistique, issue de la physique des verres, (ii) la seconde est une interprétation mécanistique. Nous montrons dans ce cas, comment mettre à profit les complémentarités de domaines en vecteur de diffusion et de résolution en énergie de différents spectromètres inélastiques de neutrons (temps de vol, backscattering et spin-écho) pour accéder, à l'aide d'un modèle physique simple, à la dynamique des protéines sur une échelle de temps allant d'une fraction de picoseconde à quelques nanosecondes.

Measurement of Diffusion Parameters and of Anisotropy of Graphite with a Pulsed Source of Neutrons; Mesure des parametres de diffusion et de l'anisotropie du graphite a l'aide d'une source pulsee 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)

1963-07-01

The diffusion coefficient, cooling coefficient, and anisotropy of graphite were determined to be (2.19 {+-} 0.03) x 10{sup 5} cm{sup 2} sec{sup -1}, (37.9 {+-} 4) x 10{sup 5} cm{sup 4} sec{sup -1}, and 1.017 {+-} 0.008, respectively. The range of geometrical buckling was from 7 to 155 m{sup -2}. The values obtained are compared with published values. (authors) [French] Un programme experimental utilisant la methode de la source pulsee de neutrons a ete realise sur le graphite. La gamme des laplaciens couverte est de 7 m{sup -2} a 155 m{sup -2}. Les resultats en sont presentes dans ce rapport: - coefficient de diffusion D{sub 0} = (2,19 {+-} 0,03) x 10{sup 5} cm{sup 2} s{sup -1} - coefficient de refroidissement C = (37,9 {+-} 4) x 10{sup 5} cm{sup 4} s{sup -1} - anisotropie du graphite (D parall./D perp.) = 1,017 {+-} 0,008. Ils sont compares aux valeurs deja publiees. (auteurs)

X-ray diffraction studies of neutron irradiated beryllium oxide; Etude par diffraction des rayons X de glucine irradiee aux neutrons

Energy Technology Data Exchange (ETDEWEB)

Belbeoch, B; Rodot, J; Roulliay, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1966-07-01

X-ray diagrams of neutron irradiated sintered BeO (E > 1 MeV) can be classified into 3 types depending on the value of the c/a ratio, independently of the characteristics of the sinters, (hot pressed or cold pressed, material, variable density and grain size, pure BeO or BeO with additions). The anisotropic expansion of the BeO lattice caused by irradiation depends not only on the dose of irradiation, but also on the intergranular stresses of the samples. The sintered samples which resist most satisfactorily to irradiation consist of cold pressed high density and small grain size material for which the c/a ratio varies little with increasing dosage. The single crystal diagrams show the same classification as the polycrystalline material in respect with the c/a ratio. Furthermore, in addition to the Bragg reflections, X-ray diffusions are observed; the distribution of the defects is no longer isotropic; a description is given in particular of the scattering which occurs when 1.625 < c/a < 1.645. It appears that the defects start to congregate and cause pronounced distortion of the lattice. The defects are stabilized in the form of large aggregates only when a later irradiation stage is reached. (author) [French] Les diagrammes de rayons X obtenus a l'aide d'echantillons frittes de BeO irradies aux neutrons (E > 1 MeV) peuvent etre classes en trois types qui dependent de la valeur du rapport c/a, quelles que soient les caracteristiques du fritte (fritte sous charge ou fritte naturel, densite et taille de grains variables, BeO pur ou avec additions). La dilatation anisotrope de la maille de BeO observee par irradiation ne depend pas seulement de la dose recue, mais aussi des etats de contraintes intergranulaires des echantillons; ainsi les frittes qui resistent le mieux sous irradiation sont les frittes naturels a forte densite et a petits grains pour lesquels le rapport c/a varie peu en fonction de la dose d'irradiation. Les diagrammes de monocristaux presentent la

Fast neutron irradiation effects on diffusion processes in the aluminum-magnesium system; Effets de l'irradiation aux neutrons rapides sur les phenomenes lies a la diffusion dans le systeme aluminium-magnesium

Energy Technology Data Exchange (ETDEWEB)

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

1969-06-01

Examination of bulky diffusion couples Al (Mg) - Al and Mg (Al) - Mg handled in same thermal conditions (between 200 and 440 C) out of pile and under fast neutron irradiation show, in the latter case: 1 - An increase of the growth kinetics of {beta} phase which can be explained with KIDSON' s formula. 2 - An apparent increase of solubility caused by migration of a part of excess vacancies as complexes (vacancy - solute atom) to sinks (stacking faults, grain boundaries) or to sub-microscopical clusters. 3 - An enhancement of chemical diffusion at low temperature. At infinite dilution, chemical diffusion coefficient of Mg in Al can be expressed in normal conditions as: D = 1 exp(- 31000/RT {+-} 1200/RT cal/mole) cm{sup 2}.s{sup -1} and under irradiation as: D = 8.10{sup -3} exp(-24500/RT {+-} 1200/RT cal/mole) cm{sup 2}.s{sup -1}. Interpretation can be carried out by DIENES and Damask's theory. Excess defects (vacancies and interstitials generated in equal numbers by radiation) annihilate by migration to sinks and by direct recombination. Sinks density varies with temperature and irradiation time. The part of complexes (vacancy-solute atom) is important in the vacancies annealing kinetics. (author) [French] L'examen de couples de diffusion massifs Al (Mg) - Al et Mg (Al) - Mg traites dans les memes conditions thermiques (entre 200 et 440 C) hors pile et sous flux de neutrons rapides montre dans le dernier cas: 1 - Une acceleration de la cinetique de croissance de la phase {beta} a basse temperature dont on peut rendre compte a l'aide de la formule de KIDSON. 2 - Une augmentation apparente de la solubilite due a l'elimination d'une partie des lacunes en exces sous forme de complexes (lacune -solute) sur des pieges (dislocations, joints) ou sous forme d'amas sub-microscopiques. 3 - Une acceleration de la diffusion a basse temperature. A dilution infinie la diffusion (en cm{sup 2}/s) du Mg dans l'Al passe de: 1 exp(- 31000/RT {+-} 1200/RT cal/mole) a 8.10{sup -3} exp

Fast neutron spectrometry and dosimetry; Spectrometrie et dosimetrie des neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

We have studied fast neutron spectrometry and dosimetry through the recoil protons they produce in hydrogenated samples. In spectrometric, we used nuclear emulsions, in dosimetric, we used polyethylene coated with zinc sulphide and placed before a photomultiplier. (author)Fren. [French] Nous avons etudie la spectrometrie et la dosimetrie des neutrons rapides en utilisant les protons de recul qu'ils produisent dans une matiere hydrogenee. En spectrometrie, nous avons employe des emulsions nucleaires, en dosimetrie, du polyethylene recouvert de sulfure de zinc place devant un photomultiplicateur. (auteur)

The fortran programme for the calculation of the absorption and double scattering corrections in cross-section measurements with fast neutrons using the monte Carlo method (1963); Programme fortran pour le calcul des corrections d'absorption et de double diffusion dans les mesures de sections efficaces pour les neutrons rapides par la methode de monte-carlo (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

A calculation for double scattering and absorption corrections in fast neutron scattering experiments using Monte-Carlo method is given. Application to cylindrical target is presented in FORTRAN symbolic language. (author) [French] Un calcul des corrections de double diffusion et d'absorption dans les experiences de diffusion de neutrons rapides par la methode de Monte-Carlo est presente. L'application au cas d'une cible cylindrique est traitee en langage symbolique FORTRAN. (auteur)

Study of the propagation of fast neutrons in water, by Monte-Carlo methods; Etude de la propagation des neutrons rapides dans l'eau par des methodes de Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

We have studied the propagation in water of neutrons from mono-directional plane sources with energies ranging from 300 keV to 19,66 MeV, placed in an infinite water medium. The exact paths of a number of neutrons are determined, taking into account the microscopic sections, assuming that inelastic collisions of the neutrons on oxygen are absorptions, and neglecting the loss of energy by elastic collisions on oxygen. The neutron lifetimes have been made use of to study the propagation of neutrons from fission sources, Po-Be, Po-B and Ra-Be, as well as the reflection of fast neutrons on a semi-infinite water medium. We have taken complete account of the first collision in order to improve the precision of the results. The calculations were carried out by Mrs J. VASSEUR and Mr A. GUILLOU. (author)Fren. [French] Nous etudions la propagation dans l'eau des neutrons a partir de sources planes monodirectionnelles dont les energies sont repartis de 300 keV a 19,66 MeV, placees dans un milieu infini d'eau. Nous determinons les trajectoires exactes d'un certain nombre de neutrons en tenant compte des sections microscopiques, en supposant que les chocs inelastiques des neutrons sur l'oxygene sont des absorptions, et en negligeant la perte d'energie par chocs elastiques sur l'oxygene. Les vies de neutrons ont ete exploitees pour etudier la propagation des neutrons a partir de sources de fission, Po-Be, Po-B et Ra-Be, ainsi que la reflexion des neutrons rapides sur un milieu semi-infini d'eau. On a tenu compte integralement du premier choc pour ameliorer la precision des resultats. Les calculs ont ete effectues par Mme J. VASSEUR et M.A. GUILLOU. (auteur)

Contribution to the radio crystallographic study of the effect of neutrons on LiF monocrystals (1962); Contribution a l'etude radiocristallographique de l'effet des neutrons sur les monocristaux de LiF (1962)

Energy Technology Data Exchange (ETDEWEB)

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

1962-07-01

We have studied the changes in the 200 and 400 reflections in LiF monocrystals irradiated by neutrons, for doses of between 10{sup 17} and 10{sup 18} n/cm{sup 2}. The reflections are distinctly dissymmetric and in the case of doses under 5.10{sup 17} appear to consist of a fine peak associated with a wide and diffuse base. These appear to exist organised groups of point defects which lead to perturbed zones whose number varies very little with the irradiation and which exist perhaps beforehand. (authors) [French] Nous avons etudie l'evolution des reflexions 200 et 400 de monocristaux de LiF irradie aux neutrons pour des doses variant de 10{sup 17} a 10{sup 18} n/cm{sup 2}. Les reflexions presentent une nette dissymetrie et pour des doses inferieures a 5.10{sup 17} semblent constituees d'un pic fin associe a une base large et diffuse. Il semble qu'il y ait des rassemblements organises de defauts ponctuels pour donner des zones perturbees dans le nombre varierait peu avec l'irradiation et lui serait peut-etre preexistant. (auteurs)

Structure of simple liquids; Structure des liquides simples

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

The results obtained by application to argon and sodium of the two important methods of studying the structure of liquids: scattering of X-rays and neutrons, are presented on one hand. On the other hand the principal models employed for reconstituting the structure of simple liquids are exposed: mathematical models, lattice models and their derived models, experimental models. (author) [French] On presente d'une part les resultats obtenus par application a l'argon et au sodium des deux principales methodes d'etude de la structure des liquides: la diffusion des rayons X et la diffusion des neutrons; d'autre part, les principaux modeles employes pour reconstituer la structure des liquides simples sont exposes: modeles mathematiques, modeles des reseaux et modeles derives, modeles experimentaux. (auteur)

Contribution to the study of magnetic diffusion of neutrons; Contribution a l'etude de la diffusion magnetique des neutrons

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

Certain statistical aspects of a large collection of electronic spins coupled by exchange forces are examined. The treatment is limited to substances where the orbital magnetic moment can be considered fixed, and where the effect of thermal agitation of the ions can be neglected. A system of this kind can be followed experimentally by elastic and inelastic diffusion of neutrons. At high temperatures, inelastic diffusion allows the microscopic aspect, reversible, and the macroscopic aspect, irreversible, to be studied simultaneously and these 2 fields to be linked. At temperatures around the Curie point, the average phenomenon is the appearance of a critical opalescence. At low temperatures, collective spin excitations can be observed. In the neighborhood of the Curie point, the spin coefficient {lambda}, which governs the relaxation of fluctuations of magnetization, is calculated. An intrinsic factor is discussed in {lambda}, bound to the microscopic frequency of the spin exchanges, and to a factor due to the damping of the diffusion by the magnetic field. At the transition point, {lambda} is cancelled. The spectrum of the spin excitations in metals is discussed. (author)

Single Crystal Diffuse Neutron Scattering

Directory of Open Access Journals (Sweden)

Richard Welberry

2018-01-01

Full Text Available Diffuse neutron scattering has become a valuable tool for investigating local structure in materials ranging from organic molecular crystals containing only light atoms to piezo-ceramics that frequently contain heavy elements. Although neutron sources will never be able to compete with X-rays in terms of the available flux the special properties of neutrons, viz. the ability to explore inelastic scattering events, the fact that scattering lengths do not vary systematically with atomic number and their ability to scatter from magnetic moments, provides strong motivation for developing neutron diffuse scattering methods. In this paper, we compare three different instruments that have been used by us to collect neutron diffuse scattering data. Two of these are on a spallation source and one on a reactor source.

Homogenization of neutronic diffusion models; Homogeneisation des modeles de diffusion en neutronique

Energy Technology Data Exchange (ETDEWEB)

Capdebosq, Y

1999-09-01

In order to study and simulate nuclear reactor cores, one needs to access the neutron distribution in the core. In practice, the description of this density of neutrons is given by a system of diffusion equations, coupled by non differential exchange terms. The strong heterogeneity of the medium constitutes a major obstacle to the numerical computation of this models at reasonable cost. Homogenization appears as compulsory. Heuristic methods have been developed since the origin by nuclear physicists, under a periodicity assumption on the coefficients. They consist in doing a fine computation one a single periodicity cell, to solve the system on the whole domain with homogeneous coefficients, and to reconstruct the neutron density by multiplying the solutions of the two computations. The objectives of this work are to provide mathematically rigorous basis to this factorization method, to obtain the exact formulas of the homogenized coefficients, and to start on geometries where two periodical medium are placed side by side. The first result of this thesis concerns eigenvalue problem models which are used to characterize the state of criticality of the reactor, under a symmetry assumption on the coefficients. The convergence of the homogenization process is proved, and formulas of the homogenized coefficients are given. We then show that without symmetry assumptions, a drift phenomenon appears. It is characterized by the mean of a real Bloch wave method, which gives the homogenized limit in the general case. These results for the critical problem are then adapted to the evolution model. Finally, the homogenization of the critical problem in the case of two side by side periodic medium is studied on a one dimensional on equation model. (authors)

Frequency Spectrum of Liquids and Cold Neutron Scattering; Spectre de frequences des liquides et diffusion de neutrons froids; Chastotnyj spektr zhidkostej i rasseyanie kholodnykh nejtronov; Espectro de frecuencias de los liquidos y dispersion de neutrones frios

Energy Technology Data Exchange (ETDEWEB)

Singwi, K S; Sjolander, A; Rahman, A [Argonne National Laboratory, Argonne, IL (United States)

1963-01-15

An important question which arises in connection with slow neutron scattering by liquids is: does there exist a frequency spectrum tor liquids analogous to that in solids? The answer to this question is given by showing that the width function {gamma}(t) of the Gaussian space-time self-correlation function G{sub s}(r,t) can be expressed, by using the frequency spectrum of the velocity auto-correlation function, in a form which is formally identical with that of a harmonic solid. Thus a knowledge of the frequency spectrum of the velocity auto-correlation function should enable one to calculate slow neutron scattering by liquids as has been emphasized by Egelstaff and co-workers. Using a stochastic model of a liquid, the frequency spectrum f({omega}) of the velocity auto-correlation function is calculated for water at 300{sup o}K and for liquid lead at 620{sup o}K. In the case of water a maximum in f({omega}) corresponding to {Dirac_h} 75{sup o}K is predicted. For lead the maximum also occurs at nearly the same {omega}-value. There also occurs a minimum in f({omega})in both cases. Larsson and Dahlborg have observed a maximum in f({omega}) at the above predicted {omega}-value. A recent observation in liquid lead of Cotter et al. probably confirms our prediction too. (author) [French] Dans le domaine de la diffusion des neutrons lents par des liquides une question importante se pose: existe-t-il un spectre de frequences pour les liquides analogue au spectre pour les solides? On peut repondre d cette quesnon en montrant que la fonction de largeur {gamma}(t) de la fonction gaussienne spatio-temporelle d'auto-correlation G{sub s}(r,t) peut etre exprimee, en employant le spectre de frequences de la fonction d'auto-correlation des vitesses, sous uune forme qui presente le membe aspect formel que celle d'un solide harmonique. par consequent, la connaissance du spectre de frequences de la fonction d'auto-correlation des vitesses devrait permettre de calculer la diffusion des

A Cold Neutron Monochromator and Scattering Apparatus; Monochromateur et appareillage pour la diffusion de neutrons lents; Monokhromator dlya ''kholodnykh'' nejtronov i pribor dlya rasseyaniya; Monocromador y aparato de dispersion para neutrones frios

Energy Technology Data Exchange (ETDEWEB)

Harris, D; Cocking, S J; Egelstaff, P A; Webb, F J [Nuclear Physics Division, Aere, Harwell, Didcot, Berks (United Kingdom)

1963-01-15

A narrow band of neutron wavelengths (4 A and greater) is selected from a collimated neutron beam obtained from the Dido reactor at Harwell. These neutrqps are scattered by various samples and the energy transfer of the scattered neutrons measured using time-of-flight techniques. The neutrons, moderated by a liquid hydrogen source in the reactor pass through first a liquid nitrogen- cooled filter, then a single crystal of bismuth and finally they are ''chopped'' by a magnesium-cadmium high- speed curved slot rotor. In this apparatus the wavelength spread of 0. 3 A at 4 . 1 A is determined primarily by the Be-Bi filter, while the time spread (8 {mu}s) is determined by the rotor. The monochromated neutron bursts from this rotor are scattered by a sample and detected in one of two counter arrays. When studying liquid or polycrystalline samples an array of six BF{sub 3}, counter assemblies (each 2 inches x 24 inches in area)are used covering scatter angles from 20{sup o} to 90{sup o}. This array is placed below the neutron beam. Above the line of the neutron beam is a second array consisting of three scintillators 2 inches in diameter, which is used for the study of single crystal samples. The output of each counter is fed into a tape recording system which has 500 time channels available for each counter. This apparatus has been used to study neutron scattering from several gaseous, liquid and crystalline samples and the most recent measurements are presented in other papers in these proceedings. [French] Les auteurs extraient une bande etroite de neutrons ( 4 A et plus) d'un faisceau collimate de neutrons produits par le reacteur Dido de Harwell. On fait diffuser ces neutrons au moyen de divers echantillons et on mesure le transfert d'energie des neutrons diffuses par la methode du temps de vol. Les neutrons ralentis par de l'hydrogene liquide place dans le reacteur passent d'abord dans un filtre refroidi a l'azote liquide, puis dans un monocristal de bismuth

Huang diffuse scattering of neutrons

International Nuclear Information System (INIS)

Burkel, E.; Guerard, B. v.; Metzger, H.; Peisl, J.

1979-01-01

Huang diffuse neutron scattering was measured for the first time on niobium with interstitially dissolved deuterium as well as on MgO after neutron irradiation and Li 7 F after γ-irradiation. With Huang diffuse scattering the strength and symmetry of the distortion field around lattice defects can be determined. Our results clearly demonstrate that this method is feasible with neutrons. The present results are compared with X-ray experiments and the advantages of using neutrons is discussed in some detail. (orig.)

Neutrons moderation theory; Theorie du ralentissement des neutrons

Energy Technology Data Exchange (ETDEWEB)

Vigier, J P

1949-07-01

This report gives a summarized presentation of the theory of fast neutrons diffusion and moderation in a given environment as elaborated by M. Langevin, E. Fermi, R. Marshak and others. This statistical theory is based on three assumptions: there is no inelastic diffusion, the elastic diffusion has a spherical symmetry with respect to the center of gravity of the neutron-nucleus system (s-scattering), and the effects of chemical bonds and thermal agitation of nuclei are neglected. The first chapter analyzes the Boltzmann equation of moderation, its first approximate solution (age-velocity equation) and its domain of validity, the extension of the age-velocity theory (general solution) and the boundary conditions, the upper order approximation (spherical harmonics method and Laplace transformation), the asymptotic solutions, and the theory of spatial momenta. The second chapter analyzes the energy distribution of delayed neutrons (stationary and non-stationary cases). (J.S.)

Studies of magnetism with inelastic scattering of cold neutrons; Etudes de magnetisme realisees a l'aide de la diffusion inelastique de neutrons froids

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

Inelastic scattering of cold neutrons can be used to study some aspects of magnetism: spins waves, exchange integrals, vicinity of Curie point. After description of the experimental set-up, several experiments, in the fields mentioned above, are analysed. (author) [French] La technique de diffusion inelastique des neutrons froids est utilisee pour etudier certains aspects du magnetisme: ondes de spins, integrales d'echange, etude au voisinage du point de Curie, etc. Apres une description de l'appareillage, on analyse diverses experiences effectuees dans les domaines enumeres plus haut. (auteur)

Neutron spin echo spectroscopy. Its application to the study of the dynamics of polymers in solution; La spectrometrie par echos de spins de neutrons. Application a l'etude de la dynamique des polymeres en solution

Energy Technology Data Exchange (ETDEWEB)

Papoular, Robert

1992-06-15

This work focuses on Neutron Spin Echo (NSE) spectroscopy and on the NSE spectrometer MESS, which we have built at the L.L.B. (CE Saclay). After analyzing in detail the classical and quantum principles of this type of instrument, and illustrated them with optical analogies, we expound a simple formalism for the interpretation of polarized neutron experiments of the most general type. In a second part, we describe the MESS spectrometer extensively; its characteristics and performances as well as the first results obtained with this instrument. In particular, we include two papers showing how the neutron depolarization, spin rotation and echoes can be used to investigate high-Tc superconductors. The last part deals with the dynamics of Polymer-Polymer-Solvent ternary solutions and demonstrates how the Neutron Spin Echo technique becomes a privileged tool for such physico-chemical studies thanks to the joint use of NSE and contrast variation methods, coupled with the adequate ranges of time and scattering vectors accessible. Finally, we describe the specific case of partially deuterated polydimethyl-siloxane (PDMS) in semi-dilute solution in Toluene. We have experimentally and separately evidenced the cooperative and inter-diffusive diffusion modes predicted by the theory of Akcasu, Benoit, Benmouna et al. These results, obtained at the L.L.B. (CE Saclay) are the subject matter of the last paper included in this work. (author) [French] Ce memoire est centre sur la spectroscopie par echos de spins de neutrons, et plus particulierement, sur le spectrometre a echos de spins MESS que nous avons construit au L.L.B (CE/Saclay). Apres avoir detaille les principes classique et quantique de ce type d'instrument et les avoir illustres par des analogies optiques, nous detaillons un formalisme simple permettant d'interpreter les experiences utilisant les neutrons polarises dans le cas le plus general. Une seconde partie decrit de maniere approfondie le spectrometre MESS de Saclay

Contribution to the study of 14 MeV neutron scattering by {sup 12}C using a time-of-flight spectrometer (1963); Contribution a l'etude de la diffusion des neutrons de 14 MeV par {sup 12}C, a l'aide d'un spectrometre a temps de vol (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-15

Experimental measurements of 14 MeV fast neutrons scattering by {sup 12}C by time-flight spectrometer, with a 1.7 n. sec over-all resolution. The excitation of the 7.65 MeV level is observed. (author) [French] Etude experimentate de la diffusion par {sup 12}C des neutrons rapides de 14 MeV a l'aide d'un spectrometre de resolution totale de 17 n/s. Observation de l'excitation du niveau 0+ de 7,65 MeV. (auteur)

Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

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

1965-06-01

Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

Theory of the Influence of Phonon-Phonon and Electron-Phonon Interactions on the Scattering of Neutrons by Crystals; Theorie de l'influence des interactions phonon-phonon et electron-phonon sur la diffusion des neutrons par des cristaux; Teoriya vliyaniya vzaimodejstvij fonon-fonon iehlvktron-fonon na rasseyanie nejtronov kristalla-; Teoria de la influencia de las interacciones fonon-fonon y electron-fonon en la dispersion de neutrones por cristales

Energy Technology Data Exchange (ETDEWEB)

Kokkedee, J J.J. [Institute for Theoretical Physics of the University of Utrecht (Netherlands)

1963-01-15

interatomic or interionic distance in the crystal). Approximate calculations are performed to give some insight into the orders of magnitude of the effects under study. (author) [French] Comme le predit la theorie harmonique, le spectre de la diffusion inelastique coherente de neutrons par un monocristal non conducteur, pour un angle de diffusion donne, se compose d'une serie de pics de fonction delta, qui se superposent a un bruit de fond continu. Les pics correspondent a des phenomenes a un phonon, dans lesquels un phonon est absorbe ou emis par le neutron; le bruit de fond provient de phenomenes a plusieurs phonons. Lorsqu'il existe des forces anharmoniques (interaction phonon-phonon) les pics de fonction delta s'elargissent pour former des pics finis et leur frequence centrale est dephasee par rapport aux valeurs harmoniques. Dans le cas d'un metal il y a, en plus des interactions phonon-phonon, une interaction entre les phonons et les electrons de conduction, laquelle contribue a dephaser et elargir encore davantage les pics a un phonon. Continuant les travaux de Van Hove (qui avait considere le cas relativement simple d'un cristal non conducteur a l'etat fondamental, soit T = 0{sup o}K) l'auteut a etudie les deplacements et les largeurs des pics de diffusion resultant des interactions indiquees plus haut, a l'aide de la theorie de la perturbation a plusieurs particules, en ayant largement recours a la methode des diagrammes. Il admet, avant tout examen du probleme, que quelle que soit la force des interactions, la largeur de chaque pic est petite par rapport a la valeur de la frequence en son centres dans ces conditions seulement, on peut considerer que les pics sont bien definis par rapport au bruit de fond, si les calculs d'interaction sont pousses jusqu'aux ordres superieurs. On estime que cette hipothbse est realisee tant que les temperatures ne sont pas trop elevees et que les valeurs du vecteur d'onde des phonons ne sont pas trop considerables. La methode

Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

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

1965-06-01

Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

Neutron transport equation - indications on homogenization and neutron diffusion

International Nuclear Information System (INIS)

Argaud, J.P.

1992-06-01

In PWR nuclear reactor, the practical study of the neutrons in the core uses diffusion equation to describe the problem. On the other hand, the most correct method to describe these neutrons is to use the Boltzmann equation, or neutron transport equation. In this paper, we give some theoretical indications to obtain a diffusion equation from the general transport equation, with some simplifying hypothesis. The work is organised as follows: (a) the most general formulations of the transport equation are presented: integro-differential equation and integral equation; (b) the theoretical approximation of this Boltzmann equation by a diffusion equation is introduced, by the way of asymptotic developments; (c) practical homogenization methods of transport equation is then presented. In particular, the relationships with some general and useful methods in neutronic are shown, and some homogenization methods in energy and space are indicated. A lot of other points of view or complements are detailed in the text or the remarks

Theoretical Interpretation of Pulsed Neutron Phenomena; Interpretation Theorique des Phenomenes dus aux Neutrons Pulses; Teoreticheskaya interpretatsiya yavlenij, svyazannykh s impul'snymi nejtronami; Interpretacion Teorica de los Fenomenos Debidos a Neutrones Pulsados

Energy Technology Data Exchange (ETDEWEB)

Corngold, N. [Brookhaven National Laboratory, Upton, NY (United States)

1965-08-15

The author discusses the present understanding of transient effects in the variation of neutron distributions in space, time and energy. In the analysis, particular attention is given to the pulsed neutron experiment in non-multiplying media, and to its interpretation in terms of space-energy modes and (time) decay constants. Theoretical research in this field, much of which has come from the Brookhaven National Laboratory, has been built upon the exact solution of several special problems, and approximate, numerical solutions in more realistic situations. Recently, a unified point of view, based on the Van Hove-Glauber theory of the scattering of slow neutrons, has been proposed. It has been found that the long-time behaviour of the pair correlation function, which enables one to distinguish between different types of moderating materials, also controls the type of decay-constant spectrum that one finds. In his discussion the author emphasizes the case of infinite medium solutions that are proportional to exp(iB * r), in media having 1/v absorption. The analytical work is concerned with modal solutions, proportional to exp({lambda}t) as well, while the numerical work is directed at the full, ''life-history'' solution, as well as the modal solutions. The author also remarks on phenomena encountered when the absorption is not 1/v and when the moderator sample is so small that the long-time solutions are no longer separable in space and energy. Finally, the paper discusses the bearing of these recent developments upon the interpretation of pulsed experiments. Particular reference is made to the measurement of diffusion and diffusion cooling coefficients, and to the nature of the approach to the asymptotic distribution. (author) [French] L'auteur examine la conception actuelle des effets transitoires dans la variation des distributions des neutrons dans l'espace, dans le temps et selon l'energie. Dans son etude, il accorde une attention particuliere aux experiences

Measurement of the Anisotropy of Diffusion Constant in Media with Empty Channels; Mesure de l'Anisotropie de la Constante de Diffusion dans des Milieux a Canaux Vides; Izmerenie anizotropii postoyannoj diffuzii v srede s pustymi kanalami; Medicion de la Anisotropia de la Constante de Difusion en Varios Sistemas con Canales Vacios

Energy Technology Data Exchange (ETDEWEB)

Copic, M.; Kalin, T.; Pregl, G.; Zerdin, F. [Nuclear Institute Jozef Stefan, Ljubljana, Yugoslavia (Slovenia)

1964-04-15

Using the pulsed-neutron source technique, the diffusion constant was measured in systems with empty channels. Plexiglas was used as the neutron diffusing material. From separate sets of measurements on rectangular blocks the diffusion constants parallel and perpendicular to channels were determined. The average value of the diffusion constant was also obtained experimentally from measurements on cubes. The difference between both diffusion constants, D{sub Double-Up-Tack} - D{sub Up-Tack }, agrees with theoretical predictions inside the limits of experimental errors, yet the average diffusion constant lies systematically below the predictions of Behrens' theory. (author) [French] En se servant de la methode de la source des neutrons puises, les auteurs ont mesure la constante de diffusion dans des systemes a canaux vides. Comme matiere diffusant les neutrons, ils ont utilise du plexiglas. A partir de series de mesures differentes faites sur des blocs rectangulaires, ils ont determine les constantes de diffusion parallele et perpendiculaire aux canaux. Ils ont egalement obtenu experimentalement la valeiir moyenne de la constante de diffusion a i'aide de mesures faites sur des cubes. La difference entre les deux constantes de diffusion, D{sub Double-Up-Tack} - D{sub Up-Tack }, concorde avec les previsions theoriques dans les limites des erreurs d'experience; cependant, la valeur moyenne de la constante de diffusion reste systematiquement inferieure aux previsions etablies par la theorie de Behrens. (author) [Spanish] Utilizando la tecnica de la fuente neutronica pulsada, los autores midieron la constante de difusion en varios sistemas con canales vacios. Como material difusor de neutrones emplearon polimetacrilato de metilo. Basandose en varias series de mediciones efectuadas en bloques rectangulares, los autores determinaron las constantes de difusion Inverted-Question-Mark n sentido paralelo y perpendicular a los canales. Obtuvieron experimentalmente el valor

Diffuse scattering of neutrons

International Nuclear Information System (INIS)

Novion, C.H. de.

1981-02-01

The use of neutron scattering to study atomic disorder in metals and alloys is described. The diffuse elastic scattering of neutrons by a perfect crystal lattice leads to a diffraction spectrum with only Bragg spreads. the existence of disorder in the crystal results in intensity and position modifications to these spreads, and above all, to the appearance of a low intensity scatter between Bragg peaks. The elastic scattering of neutrons is treated in this text, i.e. by measuring the number of scattered neutrons having the same energy as the incident neutrons. Such measurements yield information on the static disorder in the crystal and time average fluctuations in composition and atomic displacements [fr

The study of neutron transport by oscillation method; Etude du transport des neutrons par la methode de modulation

Energy Technology Data Exchange (ETDEWEB)

Raievski, V

1959-07-01

The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [French] La methode de modulation est un procede tres general d'etude des proprietes neutroniques des milieux contenant des neutrons thermiques. Le present rapport a pour but de decrire les principales de ces experiences et d'en donner une theorie generale. Cette theorie, exposee dans la premiere partie, est etablie dons le cadre de l'approximation a deux groupes de vitesse qui a prouve son efficacite dons le cas des piles a neutrons thermiques. Le domaine de validite de l'approximation a deux groupes est rappele au debut, ce qui permet de definir avec precision la signification des parametres qui entrent dons la theorie et qui font l'objet de ces mesures. La deuxieme partie decrit les experiences realisees, en particulier celles effectuees au Centre d'Etudes Nucleaires de Saclay ou la methode a ete considerablement developpee. Ces experiences sont interpretees dans le cadre de la theorie generale exposee precedemment. On prouve ainsi l'identite des parametres mesures par cette methode et de ceux figurant

Total cross-section measurements on aluminium, carbon, fluorine and hydrogen for d, d reaction neutrons using the coincidence method (1963); Mesure de la section efficace totale de l'aluminium, du carbone, du fluor et de l'hydrogene pour des neutrons de la reaction d, d par la methode des coincidences (1963)

Energy Technology Data Exchange (ETDEWEB)

Ambrosino, G; Sorriaux, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experiment described consists in the measurement of the total cross-section of various materials: aluminium, carbon, fluorine and hydrogen, for mono-energetic 2.77 MeV neutrons obtained from the d,d reaction. The measurement is carried out by transmission. The neutrons are detected by means of a plastic scintillator mounted on a 56 AVP photomultiplier, and are isolated from all secondary phenomena (background noise, scattered neutrons) by coincidence with helium 3. which particles are associated to the neutrons from the reaction {sup 2}{sub 1}D ({sup 2}{sub 1}D, n) {sup 3}{sub 2}H The helium 3 particles are detected by a PN junction diode used with inverted polarisation. An absorption exponential has been traced out using measurements made on seven aluminium bars. The accuracy of the total cross-section measurements is about 10{sup -2}. (authors) [French] L'experience exposee dans ce rapport consiste en la mesure des sections efficaces totales de differents materiaux: aluminium, carbone, fluor et hydrogene, pour des neutrons monoenergetiques de 2,77 MeV, obtenus par la reaction d,d. La mesure est faite par transmission. Les neutrons sont detectes par un scintillateur plastique monte sur un photomultiplicateur 56 AVP, et sont separes de tout phenomene secondaire (bruit de fond, neutrons diffuses) par coincidence avec les helium 3, particules associees aux neutrons de la reaction {sup 2}{sub 1}D ({sup 2}{sub 1}D, n) {sup 3}{sub 2}H Les helium 3 sont detectes par une diode a jonction PN utilisee en polarisation inverse. Une exponentielle d'absorption a ete tracee a partir de mesures faites sur sept barreaux d'aluminium. La precision des mesures des sections efficaces totales est de l'ordre de 10{sup -2}. (auteurs)

Determination of the separation factor of uranium isotopes by gaseous diffusion; Determination des facteurs de separation des isotopes de l'uranium par diffusion gazeuse

Energy Technology Data Exchange (ETDEWEB)

Bilous, O; Counas, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

A 12-stage pilot separation cascade with a low output has been constructed to measure the separation factor of uranium isotopes by gaseous diffusion. The report describes some of the separation results obtained, and also provides information on the time necessary for equilibrium to be established and on the influence of various perturbations on the pressure profile in the cascade. (author) [French] Une cascade pilote de 12 etages de separation a faible debit a ete construite pour mesurer le facteur de separation des isotopes de l'uranium par diffusion gazeuse. Le rapport decrit certains des resultats de separation obtenus et fournit egalement des donnees sur les temps de mise en equilibre et l'influence de diverses perturbations sur le profil des pressions dans la cascade. (auteur)

Theory of Pulsed Neutron Experiments in Highly Heterogeneous Multiplying Media; Theorie des Experiences au Moyen des Neutrons Pulses, dans les Milieux Multiplicateurs Tres Heterogenes; O teorii ehksperimentov s impul'snymi neitronami v geterogennykh razmnozhayushchikh sredakh; Aspectos Teoricos de los Experimentos con Neutrones Pulsados en Medios Multiplicadores Muy Heterogeneos

Energy Technology Data Exchange (ETDEWEB)

Corno, S. E. [Instituto di Fisica ' ' A. Volta' ' (Italy); Unversity of Pavia, Pavia (Italy); SNAM, Milan [Italy

1965-10-15

In this work we investigate the time and space dependence of the neutron flux within a highly heterogeneous assembly, in which pulsed or sinusoidally modulated neutrons are injected. We consider, for the sake of simplicity, a device consisting of a cylindrical block of heavy moderator, along the axis of which a line-shaped region of fissionable material is located. The driving neutron source is assumed to be located on one of the end faces of the cylinder. The extent of the fissionable region allows us to deal with it as with an absorbing and multiplying singularity of the neutron field. As our attention is mostly concentrated on space and time variation of the neutron flux, rather crude approximations are assumed as far as the energy dependence of the neutron population is concerned. Within the limits of the age-diffusion theory, the response of the device to any neutron excitation may be found in closed form. For a sinusoidally modulated source of given frequency, it may easily be shown that, if the axial singularity were a purely absorbing one, the neutron waves being propagated along the device would possess a phase shift; a wavelength and an attenuation constant depending on the absorbing properties of the singularity. This picture becomes more and more complicated when neutron multiplication occurs. For this general case the solution derived in our paper obviously turns out to be dependent on both absorption and multiplication properties of the singularity. This circumstance suggests, among others, the idea of using a device of the type described above for testing fuel elements of heterogeneous reactors. (author) [French] L'auteur etudie les variations du flux de neutrons en fonction du temps et du lieu dans un assemblage tres heterogene ou l'on injecte des neutrons puises ou modules sinusoidalement. Il envisage, pour des raisons de simplicite, un dispositif constitue par un ralentisseur lourd en forme de bloc cylindrique sur l'axe duquel est placee une zone

Relations de Dispersion et Diffusion des Glueballs et des Mesons dans la Theorie de Jauge U(1)(2+1) Compacte

Science.gov (United States)

Ahmed, Chaara El Mouez

Nous avons etudie les relations de dispersion et la diffusion des glueballs et des mesons dans le modele U(1)_{2+1} compact. Ce modele a ete souvent utilise comme un simple modele de la chromodynamique quantique (QCD), parce qu'il possede le confinement ainsi que les etats de glueballs. Par contre, sa structure mathematique est beaucoup plus simple que la QCD. Notre methode consiste a diagonaliser l'Hamiltonien de ce modele dans une base appropriee de graphes et sur reseau impulsion, afin de generer les relations de dispersion des glueballs et des mesons. Pour la diffusion, nous avons utilise la methode dependante du temps pour calculer la matrice S et la section efficace de diffusion des glueballs et des mesons. Les divers resultats obtenus semblent etre en accord avec les travaux anterieurs de Hakim, Alessandrini et al., Irving et al., qui eux, utilisent plutot la theorie des perturbations en couplage fort, et travaillent sur un reseau espace-temps.

Use of the helium-3 proportional counter for neutron spectrometry; Utilisation du compteur proportionnel a helium 3 pour la spectrometrie des neutrons

Energy Technology Data Exchange (ETDEWEB)

Vialettes, H; Le Thanh, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

Up to now, two methods have been mainly used for neutron spectrometry near nuclear installations: - photographic emulsion spectrometry - the so-called, 'multisphere' technique spectrometry. The first method, which is fairly difficult to apply, has a threshold energy of about 500 keV; this is a big disadvantage for an apparatus which has to be used for spectrometry around nuclear installations where the neutron radiation is very much degraded energetically. The second method does not suffer from this disadvantage but the results which it yields are only approximate. In order to extend the energy range of the neutron spectra studied with sufficient accuracy the use of a helium-3 proportional counter has been considered. This report presents the principles of operation of the helium-3 spectrometer, and the calculation methods which make it possible to take into account the two main effects tending to deform the spectra obtained: - energy absorption by the walls of the counter, - energy loss of the incident neutrons due to elastic collisions with helium-3 nuclei. As an example of the application, the shape of the neutron spectrum emitted by a polonium-lithium source is given; the results obtained are in excellent agreement with theoretical predictions. (authors) [French] Jusqu'ici deux methodes ont ete utilisees principalement pour la spectrometrie des neutrons autour des installations nucleaires: - la spectrometrie par emulsions photographiques - la spectrometrie par la technique dite des multispheres. La premiere methode, d'un emploi assez delicat presente un seuil en energie d'environ 500 keV qui est un obstacle serieux a la spectrometrie autour des installations nucleaires ou le rayonnement neutronique est tres degrade en energie. La deuxieme methode ne presente pas cet inconvenient mais les resultats qu'elle permet d'obtenir ne sont qu'approches. Pour etendre la gamme d'energie des spectres de neutrons etudies avec une precision suffisante, l'utilisation du

Linear extended neutron diffusion theory for semi-in finites homogeneous means

International Nuclear Information System (INIS)

Vazquez R, R.; Vazquez R, A.; Espinosa P, G.

2009-10-01

Originally developed for heterogeneous means, the linear extended neutron diffusion theory is applied to the limit case of monoenergetic neutron diffusion in a semi-infinite homogeneous mean with a neutron source, located in the coordinate origin situated in the frontier of dispersive material. The monoenergetic neutron diffusion is studied taking into account the spatial deviations in the neutron flux to the interfacial current caused by the neutron source, as well as the influence of the spatial deviations in the absorption rate. The developed pattern is an unidimensional model for an energy group obtained of application of volumetric average diffusion equation in the moderator. The obtained results are compared against the classic diffusion theory and qualitatively against the neutron transport theory. (Author)

Diffuse neutron scattering signatures of rough films

International Nuclear Information System (INIS)

Pynn, R.; Lujan, M. Jr.

1992-01-01

Patterns of diffuse neutron scattering from thin films are calculated from a perturbation expansion based on the distorted-wave Born approximation. Diffuse fringes can be categorised into three types: those that occur at constant values of the incident or scattered neutron wavevectors, and those for which the neutron wavevector transfer perpendicular to the film is constant. The variation of intensity along these fringes can be used to deduce the spectrum of surface roughness for the film and the degree of correlation between the film's rough surfaces

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

Energy Technology Data Exchange (ETDEWEB)

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

1963-11-15

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

Etude de la dynamique en régimes quasiharmonique et anharmonique de la phase basse température de la triéthylènediamine par diffusion cohérente inélastique des neutrons

OpenAIRE

Hedoux , A.; Sauvajol , J.L.; More , M.

1988-01-01

La triéthylènediamine N(CH2CH2)3N cristallise jusqu'à Tc = 351 K dans une phase solide ordonnée de symétrie hexagonale compacte. Les expériences de diffusion cohérente inélastique des neutrons réalisées dans la phase basse température s'inscrivent dans le cadre de l'étude du processus de la transition martensitique à Tc à partir de l'évolution avec la température des paramètres dynamiques. Ainsi les courbes de dispersion ont été déterminées à 190 K, 245 K et 295 K. Les mesures effectuées à te...

Contribution to the study of neutron propagation in cavities; Contribution a l'etude de la propagation des neutrons dans les cavites

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

In large size cavities where the dimensions of the holes are greater than the mean free path of the radiations, the neutron propagation calculations are carried out by taking into account the effect of the medium surrounding the hole using a reflection coefficient or albedo. In this work the fast neutron albedos are obtained for various materials and these results are applied for a Monte-Carlo propagation calculation. A comparison of this calculation with experimental results shows the validity of the method. (author) [French] Dans les cavites de grandes dimensions, ou les dimensions des vides sont superieures au libre parcours moyen des rayonnements, le calcul de la propagation des neutrons se fait en essayant de rendre compte de l'effet du milieu entourant le vide, par un coefficient de reflexion ou albedo. Dans cette etude, sont d'une part obtenus des albedo en neutrons rapides sur divers materiaux, d'autre part ces resultats sont appliques pour un calcul de MONTE-CARLO de propagation. La comparaison entre le calcul et l'experience montre la validite de la methode. (auteur)

Fast neutron dosimetry and spectrometry using radioactivation (1963); Dosimetrie et spectrometrie des neutrons rapides par radioactivation (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

The author first recalls rapidly a few generalities concerning induced radioactivity detectors and gives, in an appendix, tables summarizing the properties of detector elements which may be used in radioprotection. The excitation functions found in the literature and also given. The technological characteristics of the detectors used are given, together with the counting methods. The many advantages of activation dosimetry for strong or periodic neutron fluxes and for those in the presence of {gamma}-radiation are stressed. The main problem in activation dosimetry is, however, the calculation of the dose absorbed from the results of the measurements. It is shown how the dose is expressed, fairly accurately, as a function of the radioactivities induced in a series of detectors. As an example, the spectrometry and the dosimetry of the neutron flux emitted by a Po-Be source are presented. (author) [French] L'auteur fait d'abord un bref rappel des generalites sur les detecteurs a radioactivite induite, accompagne, en annexe, des tableaux resumant les proprietes d'elements detecteurs utilisables en radioprotection. Les fonctions d'excitation trouvees dans la litterature y sont egalement annexees. On donne ensuite les caracteristiques technologiques des detecteurs employes ainsi que les methodes de comptage utilisees. On souligne les nombreux avantages de la dosimetrie par activation dans les flux de neutrons intenses ou periodiques et en presence de rayonnement {gamma}. Il reste que le probleme central de la dosimetrie par activation est le calcul de la dose absorbee a partir des resultats de mesure. On montre comment la dose s'exprime, de maniere approchee, en fonction des radioactivites induites dans une serie de detecteurs. A titre d'exemple, la spectrometrie et la dosimetrie du flux de neutrons emis par une source de Po-Be sont presentees. (auteur)

Thermal neutron diffusion parameters in homogeneous mixtures

Energy Technology Data Exchange (ETDEWEB)

Drozdowicz, K.; Krynicka, E. [Institute of Nuclear Physics, Cracow (Poland)

1995-12-31

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs.

Thermal neutron diffusion parameters in homogeneous mixtures

International Nuclear Information System (INIS)

Drozdowicz, K.; Krynicka, E.

1995-01-01

A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs

A Double Slow Neutron Spectrometer; Spectrometre double pour neutrons lents; Dvojnoj spektrometr medlennykh nejtronov; Espectrometro doble para neutrones lentos

Energy Technology Data Exchange (ETDEWEB)

Bondarenko, I I; Liforov, V G; Nikolaev, M N; Orlov, V V; Parfenov, V A; Semenov, V A; Smirnov, V I; Turchin, V F [Fehi, Moscow, SSSR (Russian Federation)

1963-01-15

The neutron spectrometer described in the paper is intended for measurements of the angular and energy distribution of monochromatic slow neutrons, inelasticaily scattered by liquid and solid bodies. Experiments of this type permit detailed information to be obtained concerning the dynamics of the atoms in various aggregate states of a substance. The spectromeeter is based on the time-of-flight method. The pulse source of neutrons is the IBR (1) reactor. A mechanical interrupter, rotating synchronously with the disc of the IBR and having a prescribed phase shift, serves as the monochromator. A special phasing system ensures a phasee stability better than 0.5{sup o}. The neutrons scattered by the sample are recorded by a scintillation detector set at a given angle to the neutron beam. The resolving power of the spectrometer is - 15 {mu}s/m. The paper gives a detailed description of the construction of the spectroscope and its characteristics. (author) [French] Le spectrometre neutronique decrit dans le memoire est destine a mesurer la distribution en energie, la distribution angulaire et la diffusion inelastique des neutrons lents monochromatiques dans des liquides et des solides. Les experiences de ce genre permettent d'obtenir des renseignements detailles sur la dynamique des atomes dans de divers etats de la matiere. Le spectrometre est fonde sur la mesure du temps de vol. Comme source puisee de neutrons on a utilise le reacteur IBR-1. Le systeme monochromateur etait constitue par un interrupteur mecanique en rotation synchronisee avec celle du disque du reacteur IBR, avec un dephasage determine. Un dispositif special de dephasage assure la stabilite de la phase a 0,5{sup o} pres. La diffusion des neutrons par l'echantillon est enregistree a l'aide d'un detecteur a scintillations dispose sous un angle determine par rapport au faisceau de neutrons. Le pouvoir de resolution du spectrometre est d'environ 15 {mu}s/m. Les auteurs decrivent en detail la construction du

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-03-15

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

Contribution to the study of neutron propagation in cavities; Contribution a l'etude de la propagation des neutrons dans les cavites

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

In large size cavities where the dimensions of the holes are greater than the mean free path of the radiations, the neutron propagation calculations are carried out by taking into account the effect of the medium surrounding the hole using a reflection coefficient or albedo. In this work the fast neutron albedos are obtained for various materials and these results are applied for a Monte-Carlo propagation calculation. A comparison of this calculation with experimental results shows the validity of the method. (author) [French] Dans les cavites de grandes dimensions, ou les dimensions des vides sont superieures au libre parcours moyen des rayonnements, le calcul de la propagation des neutrons se fait en essayant de rendre compte de l'effet du milieu entourant le vide, par un coefficient de reflexion ou albedo. Dans cette etude, sont d'une part obtenus des albedo en neutrons rapides sur divers materiaux, d'autre part ces resultats sont appliques pour un calcul de MONTE-CARLO de propagation. La comparaison entre le calcul et l'experience montre la validite de la methode. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

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

1952-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1952-07-01

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

Derivation of a volume-averaged neutron diffusion equation; Atomos para el desarrollo de Mexico

Energy Technology Data Exchange (ETDEWEB)

Vazquez R, R.; Espinosa P, G. [UAM-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico D.F. 09340 (Mexico); Morales S, Jaime B. [UNAM, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: rvr@xanum.uam.mx

2008-07-01

This paper presents a general theoretical analysis of the problem of neutron motion in a nuclear reactor, where large variations on neutron cross sections normally preclude the use of the classical neutron diffusion equation. A volume-averaged neutron diffusion equation is derived which includes correction terms to diffusion and nuclear reaction effects. A method is presented to determine closure-relationships for the volume-averaged neutron diffusion equation (e.g., effective neutron diffusivity). In order to describe the distribution of neutrons in a highly heterogeneous configuration, it was necessary to extend the classical neutron diffusion equation. Thus, the volume averaged diffusion equation include two corrections factor: the first correction is related with the absorption process of the neutron and the second correction is a contribution to the neutron diffusion, both parameters are related to neutron effects on the interface of a heterogeneous configuration. (Author)

Delayed neutron detection in canning burst detection studies (1961); Etude sur la detection des neutrons differes en vue de la detection des ruptures de gaines (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

This paper describes a theoretical and experimental study on the detection of neutrons present in the primary cooling circuit of a reactor cooled by heavy or light water, with a view to the installation of a canning burst detection unit. The concentration of background neutrons is first calculated, taking into account the neutrons from nitrogen 17 decay, and the photoneutrons produced by the decay of nitrogen 16 and sodium 24. The emission of delayed fission neutrons, originating at a given crack in the canning, has been estimated. Using the D{sub 2}O circuit of the pile EL-3, three units have been developed by means of which the following three types of detector may be compared: 1) BF{sub 3} proportional counter 2) Boron scintillator 3) Fission chamber Under the present experimental conditions the BF{sub 3} counter gave the best results. The influence on these detectors of the {gamma} flux, which in certain cases reaches 200 R/h, is analysed. Finally a calibration is carried out with an experimental crack of 30 mm{sup 2} of uranium exposed to a flux of 5.8 x 10{sup 13} n.cm{sup -2}.s{sup -1}. The sensitivity obtained with the BF{sub 3} counter during this test is 2 counts/s per mm{sup 2} of exposed uranium. (author) [French] Le present rapport est une etude theorique et experimentale sur la detection des neutrons presents dans le circuit primaire de refroidissement d'un reacteur refrigere par l'eau lourde ou l'eau legere, en vue d'une installation de detection de ruptures de gaines. On fait d'abord un calcul sur la concentration des neutrons de bruit de fond en tenant compte: des neutrons de decroissance de l'azote 17 et des photoneutrons produits par les decroissances de l'azote 16 et du sodium 24. L'emission des neutrons differes de fission, qui ont pour origine une fissure de gaine donnee, a ete evaluee. Utilisant le circuit D{sub 2}O de la pile EL3, trois installations ont ete mises au point permettant de comparer les trois types de detecteurs suivants: 1

Domain decomposition methods for the mixed dual formulation of the critical neutron diffusion problem; Methodes de decomposition de domaine pour la formulation mixte duale du probleme critique de la diffusion des neutrons

Energy Technology Data Exchange (ETDEWEB)

Guerin, P

2007-12-15

The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, diffusion approximation is often used. For this problem, the MINOS solver based on a mixed dual finite element method has shown his efficiency. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose in this dissertation two domain decomposition methods for the resolution of the mixed dual form of the eigenvalue neutron diffusion problem. The first approach is a component mode synthesis method on overlapping sub-domains. Several Eigenmodes solutions of a local problem solved by MINOS on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is a modified iterative Schwarz algorithm based on non-overlapping domain decomposition with Robin interface conditions. At each iteration, the problem is solved on each sub domain by MINOS with the interface conditions deduced from the solutions on the adjacent sub-domains at the previous iteration. The iterations allow the simultaneous convergence of the domain decomposition and the eigenvalue problem. We demonstrate the accuracy and the efficiency in parallel of these two methods with numerical results for the diffusion model on realistic 2- and 3-dimensional cores. (author)

Diffuse scattering of neutrons and X-rays

International Nuclear Information System (INIS)

Novion, C.H. de

1978-01-01

Diffuse scattering is used to study defect concentrations of about 10 -4 in the case of X-rays and 10 -3 in the case of neutrons. The foundations of diffuse scattering formalism are given, some experimental devices described and a few applications discussed: study by diffraction on powders of defects in CeOsub(2-x); short-range order study by X-rays on Cusub(0.75) Ausub(0.25); short-range order study by neutrons on Cusub(0.435)Nisub(0.565); short-range order study by electrons TiOx; study of irradiation-induced self-interstitials in Al; study of holes created by neutrons in Al [fr

SNAP - a three dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1993-02-01

This report describes a one- two- three-dimensional multi-group diffusion code, SNAP, which is primarily intended for neutron diffusion calculations but can also carry out gamma calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP can solve the multi-group neutron diffusion equations using finite difference methods. The one-dimensional slab, cylindrical and spherical geometries and the two-dimensional case are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. (Author)

Neutron spin-echo spectroscopy for diffusion in crystalline solids

International Nuclear Information System (INIS)

Kaisermayr, M.; Rennhofer, M.; Vogl, G.; Pappas, C.; Longeville, S.

2002-01-01

Neutron spin-echo spectroscopy (NSE) offers unprecedented opportunities in the investigation of diffusion in crystalline systems due to its outstanding energy resolution. NSE not only enables measurements at lower diffusivities than the established techniques of neutron spectroscopy, but it also gives a very immediate access to the different time scales involved in the diffusion process. This is demonstrated in detail on the example of the binary alloy NiGa where the Ni atoms hop between regular sites on the Ni sublattice and anti-sites on the Ga sublattice. Experiments on two different NSE instruments are compared to measurements using neutron backscattering spectroscopy. The potential of NSE for the investigation of jump diffusion and experimental requirements are discussed

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-06-15

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

Development of neutron diffuse scattering analysis code by thin film and multilayer film

International Nuclear Information System (INIS)

Soyama, Kazuhiko

2004-01-01

To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering by thin film, roughness of surface of thin film, correlation function, neutron propagation by thin film, diffuse scattering by DWBA theory, measurement model, SDIFFF (neutron diffuse scattering analysis program by thin film) and simulation results are explained. On neutron diffuse scattering by multilayer film, roughness of multilayer film, principle of diffuse scattering, measurement method and simulation examples by MDIFF (neutron diffuse scattering analysis program by multilayer film) are explained. (S.Y.)To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering

Dosimetry techniques of thermal neutrons and {gamma} radiation in reactor cores; Techniques de dosimetrie des neutrons thermiques et du rayonnement {gamma} dans les piles

Energy Technology Data Exchange (ETDEWEB)

Sutton, J; Draganic, I; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Chemical studies under radiation done in the reactor cores require to be followed by dosimetry. When the irradiations are done in the reflector, one can limit to the measure of the {gamma} and the neutron radiation. For the dosimetry of the {gamma} radiation, a dosimeter of ferrous sulfate is convenient until doses of about 10{sup 6} rep. The use of aired oxalic acid solutions permits to reach 10{sup 7} rep. The dosimetry of thermal neutrons has been made with solutions of cobalt sulphate or paper filter impregnated with this salt. The total chemical effect of the {gamma} and of the slow neutrons radiation is obtained with solutions of ferrous sulfate added with lithium sulphate. (M.B.) [French] Les etudes de chimie sous radiation faites dans les piles exigent d'etre suivies par dosimetrie. Lorsque les irradiations sont effectues dans le reflecteur, on peut se limiter a doser le rayonnement {gamma} et les neutrons. Pour la dosimetrie du rayonnement {gamma}, un dosimetre a sulfate ferreux convient jusqu'a des doses d'environ 10{sup 6} rep. L'emploi de solutions aerees d'acide oxalique permet d'atteindre 10{sup 7} rep. La dosimetrie des neutrons thermiques a ete faite avec des solutions de sulfate de cotalt ou du papier filtre impregne de ce sel. L'effet chimique total du rayonnement {gamma} et des neutrons lents est obtenu avec des solutions de sulfate ferreux additionne de sulfate de lithium. (M.B.)

Neutron moderation at very low temperatures (1691); Moderation des neutrons aux tres basses temperatures (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-04-15

Starting from Harwell experiment carried out inside a low-power reactor, we intended to maintain a liquid hydrogen cell in a channel of the EL3 reactor (at Saclay) whose thermal neutrons flux is 10{sup 14} neutrons/cm{sup 2}/s. We tried to work out a device giving off an important beam of cold neutrons and able to operate in a way as automatic as possible during many consecutive day without a stop. Several circuits have already been achieved at very low temperatures but they brought out volumes and fluxes much lower than those we used this time. The difficulties we have met in carrying out such a device arose on the one hand from the very high energy release to which any kind of experiment is inevitably submitted when placed near the core of the reactor, on the other, hand from the very little room which is available in experimental channels of reactors. In such condition, it is necessary to use a moderator as effective as possible. This study is divided into three parts ; in the first part, we try to determine: a) conditions in which moderation takes place, hence the volume of the cell; b) materials likely to be used at low temperature and in pile; c) cooling system; hence we had to study fluid flow conditions at very low temperatures in very long ducts. The second part is devoted to the description of the device. The third part ventilates the results we have obtained. (author) [French] Partant de l'experience de Harwell faite dans une pile de faible puissance, nous nous sommes propose de maintenir une cellule d'hydrogene liquide dans un canal de la pile EL3 de Saclay dont le flux de neutrons thermiques est de 10{sup 14} neutrons par seconde et par cm{sup 2}. Nous avons cherche a realiser une installation donnant un faisceau de neutrons froids important, et pouvant fonctionner d'une maniere aussi automatique que possible, pendant des periodes de plusieurs jours sans arret. Plusieurs circuits aux tres basses temperatures ont deja ete realises, mais ils ne mettaient

Partial radiative capture of resonance neutrons; Capture radiative partielle des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

The radiative capture of resonance neutrons has been studied near the Saclay linac between 0.5 and 700 eV with time-of-flight method and a Ge(Li) detector. {sup 195}Pt + n and {sup 183}W + n allow the study of the distribution of partial radiative widths and their eventual correlation and also the variation of < {gamma}{sub {gamma}{sub i}} > with E{sub {gamma}}. The mean values of Ml and El transition intensities are compared in several tin isotopes. Interference effects, either between resonances or between direct capture and resonant capture are found in {sup 195}Pt + n, {sup 197}Au + n and {sup 59}Co + n. The excited level schemes of a great deal of nuclei are obtained and compared with theoretical predictions. This study has been completed by an analysis of thermal spectrum. (author) [French] La capture radiative des neutrons de resonance a ete etudiee pres de l'accelerateur lineaire de Saclay entre 0,5 et 700 eV a l'aide de la methode du temps-de-vol et d'un detecteur Ge(Li). Les noyaux {sup 195}Pt + n et {sup 183}W + n permettent l'analyse de la distribution de resonance en resonance des largeurs radiatives partielles {gamma}{sub {gamma}{sub i}} et de leur eventuelle correlation, ainsi que l'etude de la variation de < {gamma}{sub {gamma}{sub i}} > en fonction de E{sub {gamma}}. Les intensites moyennes des transitions Ml et El sont comparees pour quelques isotopes de l'etain. Des effets d'interference, soit entre resonances, soit entre capture directe et capture resonnante sont mis en evidence dans {sup 195}Pt + n, {sup 197}Au + n et {sup 59}Co + n. Enfin les schemas des etats excites d'un grand nombre de noyaux sont obtenus et compares avec les predictions theoriques. Cette etude a ete completee par une analyse des spectres thermiques. (auteur)

Partial radiative capture of resonance neutrons; Capture radiative partielle des neutrons de resonance

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

The radiative capture of resonance neutrons has been studied near the Saclay linac between 0.5 and 700 eV with time-of-flight method and a Ge(Li) detector. {sup 195}Pt + n and {sup 183}W + n allow the study of the distribution of partial radiative widths and their eventual correlation and also the variation of < {gamma}{sub {gamma}{sub i}} > with E{sub {gamma}}. The mean values of Ml and El transition intensities are compared in several tin isotopes. Interference effects, either between resonances or between direct capture and resonant capture are found in {sup 195}Pt + n, {sup 197}Au + n and {sup 59}Co + n. The excited level schemes of a great deal of nuclei are obtained and compared with theoretical predictions. This study has been completed by an analysis of thermal spectrum. (author) [French] La capture radiative des neutrons de resonance a ete etudiee pres de l'accelerateur lineaire de Saclay entre 0,5 et 700 eV a l'aide de la methode du temps-de-vol et d'un detecteur Ge(Li). Les noyaux {sup 195}Pt + n et {sup 183}W + n permettent l'analyse de la distribution de resonance en resonance des largeurs radiatives partielles {gamma}{sub {gamma}{sub i}} et de leur eventuelle correlation, ainsi que l'etude de la variation de < {gamma}{sub {gamma}{sub i}} > en fonction de E{sub {gamma}}. Les intensites moyennes des transitions Ml et El sont comparees pour quelques isotopes de l'etain. Des effets d'interference, soit entre resonances, soit entre capture directe et capture resonnante sont mis en evidence dans {sup 195}Pt + n, {sup 197}Au + n et {sup 59}Co + n. Enfin les schemas des etats excites d'un grand nombre de noyaux sont obtenus et compares avec les predictions theoriques. Cette etude a ete completee par une analyse des spectres thermiques. (auteur)

Diffusion along and around dislocations; Diffusion le long et autour des dislocations

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

We have gathered together in this text some solutions for Fick's equations applicable to diffusion in dislocations. The problems is fairly similar to that of grain boundaries but in this case a further difficulty arises, purely mathematical in fact, due to the supposedly cylindrical shape of the perturbed region around a dislocation. It follows that Fick's equation is used in the form: {partial_derivative}C/{partial_derivative}t=D[{partial_derivative}{sup 2}C/{partial_derivative}r{sup 2}+1/r{partial_derivative}C/{partial_derivative} r + {partial_derivative}{sup 2}C/{partial_derivative}z{sup 2}] in order to express simply the limiting conditions and so that the solution takes into account the symmetry of revolution of the system. This very much complicates the final form of the results. We give here as an illustration a solution obtained using the same hypotheses and making the same approximations as those employed by WHIPPLE for his grain boundary work. Unfortunately the final form is not suitable for a numerical calculation. By making grosser approximations, such a those used by FISHER, it is possible to determine the parameter D/(a{sup 2}D'); the same result as for grain boundary is found i.e that the logarithm of the mean concentration varies linearly with penetration, the slope of this straight line is proportional to {radical}(D/(a{sup 2}D')) Finally we give the exact solution for a platelet of finite thickness assuming that the diffusion in the defect less crystal is negligible and that the surface diffusion is infinitely fast. This is the problem dealt with by HENDRIGKSON and MACHLIN. We arrive at conclusions different to those obtained by these two authors. (author) [French] Nous avons groupe dans ce texte quelques solutions des equations de Fick applicables a la diffusion dans les dislocations. Le probleme est assez analogue a celui des joints de grains mais il s'introduit ici une difficulte supplementaire, d'ailleurs purement mathematique, due au fait de

La diffusion spatiale des technologies de l'information géographique en France

Directory of Open Access Journals (Sweden)

Stéphane Roche

1999-03-01

Full Text Available La diffusion des technologies de l'information géographique sur le territoire français s'inscrit dans les modèles classiques «centre-périphérie» de diffusion des innovations technologiques. Pourtant, au-delà de ces grandes dynamiques nationales, la distribution spatiale des TIG reflète les caractéristiques institutionnelles, culturelles et spatiales spécifiques d'un secteur géomatique en plein développement.

Chemical order-disorder in alloys. Study by neutrons diffuse diffusion

International Nuclear Information System (INIS)

Novion, C. de; Beuneu, B.

1993-01-01

Applications of neutrons diffuse diffusion for short distance chemical order in FCC transition metals solid solutions (Pd-V, Ni-V, Ni-Cr) and understoichiometric carbides or nitrides of transition metals (TiC 1-x , NbC 1-x , TiN 1-x ) are shortly presented with theoretical and experimental aspects. (A.B.)

The study of neutron propagation in natural uranium; Etude de la propagation des neutrons dans l'uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Campan, J L; Clauzon, P P; Kirchner, B; Ribon, P; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

Fast exponential experiment. Experimental facilities are described; the experimental results obtained with the block of natural uranium are then presented. Detectors used were fission chambers, activation detectors and nuclear emulsions. One tries to explain the results, which are, then compared with theoretical values obtained by calculations based on diffusion theory. Measured spectrum (pseudo equilibrium) may be characterised by the following values: diffusion length: 95,25 {+-} 0,5; {sigma}{sub f}U{sub 235}/{sigma}{sub f}U{sub 238} : 230 {+-} 10, {sigma}{sub f}U{sub 235}/{sigma}{sub f}Np{sub 237} : 14 {+-} 0.5, {sigma}{sub f}U{sub 235}/{sigma}{sub f}U{sub 233} : 0.67 {+-} 0.03, {sigma}{sub f}U{sub 235}/{sigma}{sub f}Pu{sub 239} : 0.90 {+-} 0.05, {sigma}{sub f}U{sub 235}/{sigma}(ny) In : 230 {+-}10. (author) [French] Experience exponentielle rapide. Apres une description du dispositif experimental, on presente des resultats de mesures effectuees dans un massif d'uranium naturel. Les detecteurs utilises sont des chambres a fission, des detecteurs d'activation et des emulsions nucleaires. Une interpretation des resultats est donnee; ceux-ci sont compares a des valeurs theoriques obtenues par des calculs utilisant la theorie de la diffusion. Le spectre mesure (pseudo equilibre) peut etre caracterise par les valeurs suivantes: Longueur de diffusion: 95,25 {+-} 0,5; {sigma}{sub f}U{sub 235}/{sigma}{sub f}U{sub 238} : 230 {+-} 10, {sigma}{sub f}U{sub 235}/{sigma}{sub f}Np{sub 237} : 14 {+-} 0.5, {sigma}{sub f}U{sub 235}/{sigma}{sub f}U{sub 233} : 0.67 {+-} 0.03, {sigma}{sub f}U{sub 235}/{sigma}{sub f}Pu{sub 239} : 0.90 {+-} 0.05, {sigma}{sub f}U{sub 235}/{sigma}(ny) In : 230 {+-}10. (auteur)

Experiments on neutron-attenuation by ordinary concrete; Etude experimentale de l'attenuation des neutrons dans le beton ordinaire

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

The penetration of fission-neutrons in water and glass mixtures has been investigated in the NAIADE facility. The slopes of the thermal and fast fluxes (the latter measured with a dosimeter) remain similar when the volume proportion of water is greater than 15 per cent. For smaller water-contents the measurements show the evidence of streaming, presumably due to slowing-down neutrons of energy smaller than 300 keV. Since a volume proportion of about 15 per cent of water in glass corresponds to the composition of ordinary concrete, the present work enables to predict the effects of a desiccation of concrete. Moreover, it is seen that there is no interest in increasing the proportion of water in ordinary concrete beyond the usual values (15-20 per cent). (author) [French] La propagation des neutrons de fission a ete etudiee dans les melanges verre-eau au moyen du dispositif NAIADE. La pente du flux thermique et la pente du flux rapide (obtenue au moyen d'un dosimetre) restent voisines lorsque la proportion d'eau dans le melange depasse 15 pour cent en volume. Pour des proportions d'eau inferieures les mesures mettent en evidence un 'streaming' de neutrons en ralentissement d'energie probablement inferieure a 300 keV. La composition des melanges verre-eau aux environs de 15 pour cent d'eau en volume correspondant au beton ordinaire, la presente etude permet de prevoir les effets d'une dessication du beton, et montre, en outre, qu'il n'y a pas interet a essayer d'augmenter la proportion d'eau dans le beton ordinaire au dela des valeurs usuelles (15 a 20 pour cent). (auteur)

Some Results of the Research Work on the Biological Effect of Neutrons and Protons; Quelques Resultats des Etudes sur les Effets Biologiques des Neutrons et des Protons; Nekotorye itogi izucheniya biologicheskogo dejstviya nejtronov i protonov; Algunos Resultados del Estudio de los Efectos Biologicos de los Neutrones y Protones

Energy Technology Data Exchange (ETDEWEB)

Moskalev, Ju. I. [Institut Biofiziki AMN SSSR, SSSR (Russian Federation)

1964-03-15

The author correlates and analyses the experimental data obtained over the past two or three years regarding the biological effects of neutrons and high-energy protons, and shows that it is a matter of prime concern for present-day radiobiology to investigate the relative biological effectiveness of the various types of radiation under various irradiation conditions and the reasons for the qualitative differences in their effect on the animal organism. Attention is drawn to the need for more research into the combined effect of various types of radiation and other agents. Specific examples are cited to demonstrate the main progress achieved in studying the prophylaxis of radiation injury induced by neutrons or high-energy protons and to show the contribution such research can make towards understanding the specific way in which various types of radiation act. (author) [French] L'auteur generalise les resultats des recherches experimentales qui ont ete faites au cours des deux ou trois dernieres annees sur les effets biologiques des neutrons et protons de haute energie. Il montre qu'une des taches importantes de la radiobiologie moderne est d'etudier a la fois l'efficacite biologique relative des divers types de rayonnements dans des conditions d'irradiation differentes et les differences qualitatives que presente leur action sur l'organisme vivant. Le memoire souligne qu'il est indispensable de developper les recherches sur l'effet combine des divers facteurs: rayonnements et autres. En partant d'exemples concrets, l'auteur expose les principaux resultats des recherches-sur la prophylaxie des radiolesions provoquees par les neutrons et les protons de haute energie; il montre egalement a quel point ces travaux sont necessaires pour mieux comprendre le caractere specifique de l'e ffet des divers types de rayonnements. (author) [Spanish] En esta memoria se generalizan los resultados de los estudios experimentales efectuados en los ultimos dos o tres anos en materia de

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

Derivation of the neutron diffusion equation

International Nuclear Information System (INIS)

Mika, J.R.; Banasiak, J.

1994-01-01

We discuss the diffusion equation as an asymptotic limit of the neutron transport equation for large scattering cross sections. We show that the classical asymptotic expansion procedure does not lead to the diffusion equation and present two modified approaches to overcome this difficulty. The effect of the initial layer is also discussed. (authors). 9 refs

Albedo-adjusted fast-neutron diffusion coefficients in reactor reflectors

International Nuclear Information System (INIS)

Terney, W.B.

1975-01-01

In the newer, larger pressurized-water reactor cores, the calculated power distributions are fairly sensitive to the number of neutron groups used and to the treatment of the reflector cross sections. Comparisons between transport and diffusion calculations show that the latter substantially underpredict the reflector albedos in the fast (top) group and that the power distribution is shifted toward the core center when compared to 4-group transport theory results. When the fast-neutron diffusion coefficients are altered to make the transport- and diffusion-theory albedos agree, the power distributions are also brought into agreement. An expression for the fast-neutron diffusion coefficients in reflector regions has been derived such that the diffusion calculation reproduces the albedo obtained from a transport solution. In addition, a correction factor for mesh effects applicable to coarse mesh problems is presented. The use of the formalism gives the correct albedos and improved power distributions. (U.S.)

Prompt neutron emission; Emission des neutrons prompts de fission

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

It is shown that Ramanna and Rao's tentative conclusion that prompt fission neutrons are emitted (in the fragment system) preferentially in the direction of fragment motion is not necessitated by their angular distribution measurements, which are well explained by the usual assumptions of isotropic emission with a Maxwell (or Maxwell-like) emission spectrum. The energy distribution (Watt spectrum) and the angular distribution, both including the effects of anisotropic emission, are given. (author) [French] On montre que la conclusion experimentale de Ramanna et Rao selon laquelle les neutrons prompts de fission sont emis (dans le systeme de reference des fragments) preferentiellement dans la direction du mouvement du fragment, ne decoule pas necessairement de leurs mesures de distribution angulaire. Celles-ci sont bien expliquees par l'hypothese classique de l'emission isotrope et d'un spectre d'emission maxwellien (ou quasi-maxwellien). On donne la distribution en energie (ou spectre de Watt) et la distribution angulaire, comprenant toutes les deux les effets d'emission anisotrope. (auteur)

Application of microhardness measurements to the study of diffusion zones; Application des mesures de microdurete a l'etude des zones de diffusion

Energy Technology Data Exchange (ETDEWEB)

Levy, V; Bouchet, P; Siouffi, J; Adda, Y [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

It is well known that the concentration fluctuations in diffusion zones can be determined by microhardness tests. It was therefore interesting to define the experimental conditions leading to the best accuracy. To this end, we have studied the influence of the kind of penetrator used, and of the applied load on the reproducibility of the curves plotting the variation of microhardness values against penetration. This work carried out for diffusion zones of various widths, has enabled us to determine in each case the best experimental procedure. (author) [French] On sait que les variations de concentration existant dans les zones de diffusion peuvent etre determinees au moyen de mesures de microdurete. Il etait donc interessant de definir les conditions experimentales conduisant a la meilleure precision. Dans ce but, nous avons etudie l'influence de la nature du penetrateur et celle de la charge appliquee sur la reproductibilite des courbes representant les variations de durete en fonction de la penetration. Ce travail, effectue pour des zones de diffusion de differentes largeurs, nous a permis de determiner dans chaque cas les conditions experimentales les plus favorables. (auteur)

Reflector modelization for neutronic diffusion and parameters identification

International Nuclear Information System (INIS)

Argaud, J.P.

1993-04-01

Physical parameters of neutronic diffusion equations can be adjusted to decrease calculations-measurements errors. The reflector being always difficult to modelize, we choose to elaborate a new reflector model and to use the parameters of this model as adjustment coefficients in the identification procedure. Using theoretical results, and also the physical behaviour of neutronic flux solutions, the reflector model consists then in its replacement by boundary conditions for the diffusion equations on the core only. This theoretical result of non-local operator relations leads then to some discrete approximations by taking into account the multiscaled behaviour, on the core-reflector interface, of neutronic diffusion solutions. The resulting model of this approach is then compared with previous reflector modelizations, and first results indicate that this new model gives the same representation of reflector for the core than previous. (author). 12 refs

Apparatus for studying the diffusion of rare gases in stainless steel; Appareil pour etude de la diffusion des gaz rares dans l'acier inoxydable

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

This apparatus enables measurements to be carried out on the diffusion of gaseous fission products and of gases in general across thin metallic walls at high temperatures. This work was initially intended to solve the problems involved in systems for detecting the rupture of a fuel element can (D.R.G.) by the diffusion of fission products through the cans at high temperatures. The extension of the work to other fields is envisaged. (author) [French] Cet appareil permet d'effectuer des mesures sur la diffusion des produits de fission gazeux, et des gaz en general, au travers de parois metalliques minces a haute temperature. Au depart, ce procede devait contribuer a resoudre les problemes poses aux systemes detecteurs de rupture de gaine (D.R.G.), par la diffusion des produits de fission au travers des gaines de cartouches a haute temperature. Son extension a d'autres etudes est envisagee. (auteur)

Use of the Neutron Die-Away Technique to Test Control Rod Effectiveness Theories; Emploi de la Methode d'Absorption des Neutrons pour Verifier les Theories sur l'Efficacite des Barres de Commande; Ispol'zovanie metoda spada potoka nejtronov dlya proverki teorij ehffektivnosti reguliruyushchikh sterzhnej; Aplicacion de la Tecnica de Extincion Neutronica a la Verificacion de las Teorias sobre la Eficacia de las Barras de Control

Energy Technology Data Exchange (ETDEWEB)

Perez, R. B. [University of Florida, Gainesville, FL (United States); De Saussure, G.; Silver, E. G. [University of Florida, Gainesville, FL (United States); Oak Ridge National Laboratory, Oak Ridge, TN (United States)

1964-04-15

extremely accurate. The disagreement found for the thick rods is to be expected when diffusion theory is used to describe the effect of absorbers with cross-sectional dimensions comparable to the neutron mean free path in the moderator. Measurements with rods of intermediate sizes are being carried out to determine the point at which diffusion theory becomes inadequate. (author) [French] Le calcul de l'efficacite des barres de commande se trouve complique par le fait qu'elle depend a la fois de la repartition energetique des neutrons et de la geometrie de l'ensemble. La comparaison entre la theorie et les resultats experimentaux obtenus sur des reacteurs ou des systemes sous- critiques souleve des difficultes, en raison de la complexite intrinseque de ces systemes. La methode d'absorption des neutrons permet d'utiliser un modele a neutrons exclusivement thermiques, dans lequel on peut dissocier la repartition energetique des neutrons.des effets spatiaux. Il s'ensuit que l'on peut etudier ie facteur geometrique de l'efficacite de la barre de commande sans ten ir compte des details du spectre des neutrons, et comparer les resultats a un dispositif experimental simple non empoisonne. La methode est fondee sur le fait que, dans une experience d'absorption des neutrons du genre de celle que decrivent les autetits, le laplacien de l'ensemble est fonction de la constante de decroissance du mode fondamental. B{sup 2} = ({lambda} - {lambda}{sub a})/D {lambda}{sub a} = inverse de la periode des neutrons dans le moderateur (s{sup -1}) D = constante de diffusion (cm{sup 2}/s) Le milieu ralentisseur utilise pour ces experiences etait constitue par des prismes rectangulaires de. beryllium, de dimensions diverses, dont les plus petits etaient des blocs de 2,54 cm de haut et 7,3 de cote. Trois types de barres de cadmium ont ete utilises: des barres minces de 0,476 cm de diametre, une barre de section cruciforme, et des barres creuses et 'epaisses' ayant une section carree de 7,3 de ctfte

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

Energy Technology Data Exchange (ETDEWEB)

Costa da Silva, Adilson; Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil); Senra Martinez, Aquilino, E-mail: aquilino@lmp.ufrj.br [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil)

2011-07-15

Highlights: > We proposed a new neutron diffusion hybrid equation with external neutron source. > A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. > 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

International Nuclear Information System (INIS)

Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

2011-01-01

Highlights: → We proposed a new neutron diffusion hybrid equation with external neutron source. → A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. → 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

Continuous and discreet methods in the aggregation and des fuzzy stages of a diffuse controller of neutron power

International Nuclear Information System (INIS)

Najera H, M.C.; Benitez R, J.S.

2003-01-01

The results of a comparative study are presented of: to) A denominated diffuse controller 'exact', designed by means of an innovative method that determines analytically so much the group of exit resultant in the aggregation stage like the de fuzzy process, and b) a diffuse controller denominated 'discreet' based on the discretization of the variable of having left as much for the aggregation as for the de fuzzy. These stages incorporated to the control algorithms whose objective is the ascent and regulation of the neutron power, carrying out an analysis of its performance. (Author)

Experiments on neutron-attenuation by water-lead mixtures; Etude experimentale de l'attenuation des neutrons dans les melanges plomb-eau

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

The penetration of fission neutrons in water-lead mixtures has been investigated in the NAIADE facility. The slopes of the thermal and fast fluxes (the latter measured with a dosimeter) remain similar when the volume proportion of water is greater than 65 per cent. For smaller water contents, the measurements show the evidence of streaming presumably due to slowing-down neutrons of energy smaller than 300 keV. (author) [French] La propagation des neutrons de fission a ete etudiee dans les melanges plomb-eau au moyen du dispositif NAIADE. La pente du flux thermique et la pente du flux rapide (obtenue au moyen d'un dosimetre) restent voisines lorsque la proportion d'eau dans le melange depasse 65 pour cent en volume. Pour des proportions inferieures les mesures mettent en evidence un 'streaming' de neutrons en ralentissement d'energie probablement inferieure a 300 keV. (auteur)

SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion

International Nuclear Information System (INIS)

Daneri, A.; Toselli, G.

1976-01-01

1 - Nature of physical problem solved: SHREDI is a removal - diffusion neutron shielding code. The program computes neutron fluxes and activations in bidimensional sections (x,y or r,z) of the shield. It is also possible to consider shielding points with the same y or z coordinate (mono-dimensional problems). 2 - Method of solution: The integrals which define the removal fluxes are computed in some shield points by means of a particular algorithm based on the Simpson's and trapezoidal rules. For the diffusion calculation the finite difference method is used. The removal sources are interpolated in all diffusion points by Chebyshev polynomials. 3 - Restrictions on the complexity of the problem: Maxima: number of removal energy groups NGR = 40; number of diffusion energy groups NGD = 40; number of the reactor core and shield materials NCMP = 50; number of core mesh points in r (or x) direction for integral calculation = 75; number of core mesh points in z (or y) direction for integral calculation = 75; number of core mesh points in theta (or z) direction for integral calculation = 75; number of shield mesh points for the neutron flux calculation in r (or x) direction NPX = 200; number of shield mesh points for the neutron flux calculation in z (or y) direction NPY = 200; n.b. (NPX * NPY) le 12000

Neutron diffusion: connection with the theory of browniam motion

International Nuclear Information System (INIS)

Dellagi, Mohamed

1977-01-01

The displacement of the neutron projection on an axis Ox and its density of probability are introduced instead of describing the diffusion theory with neutron density, as is usual. If the point source O is isotropic and neutron monoenergetic, the brownian particle described by Langevin's equation and neutron have the same time correlation of velocity [fr

The structure of para-toluidine by X-ray and neutron diffraction; Etude de la structure de la para-toluidine par la diffraction des rayons X et des neutrons

Energy Technology Data Exchange (ETDEWEB)

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

1965-12-01

The crystal and molecular structure of para-toluidine has been solved by X-ray and neutron diffraction counter techniques. The molecules are arranged in the form of infinite chains in the crystal, each molecule being linked to two neighbours by hydrogen bonds. The presence of the H bonds makes clear the difference in the melting points between para-toluidine and benzene hydrocarbons of related symmetry and molecular weight. Their direction accounts for the (001) cleavage and the growth anisotropy of crystals from supersaturated vapour phase. A structure-seeking method by computer has been elaborated, using lattice energy calculations applied to molecules treated as rigid bodies and making use of a simplex method for function minimization without calculation of derivatives. The way the available information is handled allows to increase the range of convergence, as shown in the case of para-toluidine. (author) [French] La structure cristalline et moleculaire de la para-toluidine a ete resolue a l'aide de la diffraction des rayons X et des neutrons. L'analyse a revele que la structure presentait un encha ement particulier des molecules, liees entre elles par des ponts hydrogene. La presence des liaisons rend compte de l'elevation du point de fusion de la para-toluidine par rapport a celui des carbures benzeniques de symetrie et de poids moleculaires voisins. La direction des liaisons explique le clivage facile (001) des cristaux obtenus par fusion et l'anisotropie de croissance des cristaux en phase vapeur. Une methode de recherche de structures par ordinateur a ete elaboree, faisant usage du concept d'energie reticulaire applique aux molecules considerees comme rigides et faisant appel a une methode de simplexe pour la minimisation des fonctions sans calcul de derivees. La maniere dont est exploitee l'information disponible permet d'accro re, comme cela est montre dans le cas de la para-toluidine, l'etendue du domaine de

The structure of para-toluidine by X-ray and neutron diffraction; Etude de la structure de la para-toluidine par la diffraction des rayons X et des neutrons

Energy Technology Data Exchange (ETDEWEB)

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

1965-12-01

The crystal and molecular structure of para-toluidine has been solved by X-ray and neutron diffraction counter techniques. The molecules are arranged in the form of infinite chains in the crystal, each molecule being linked to two neighbours by hydrogen bonds. The presence of the H bonds makes clear the difference in the melting points between para-toluidine and benzene hydrocarbons of related symmetry and molecular weight. Their direction accounts for the (001) cleavage and the growth anisotropy of crystals from supersaturated vapour phase. A structure-seeking method by computer has been elaborated, using lattice energy calculations applied to molecules treated as rigid bodies and making use of a simplex method for function minimization without calculation of derivatives. The way the available information is handled allows to increase the range of convergence, as shown in the case of para-toluidine. (author) [French] La structure cristalline et moleculaire de la para-toluidine a ete resolue a l'aide de la diffraction des rayons X et des neutrons. L'analyse a revele que la structure presentait un encha ement particulier des molecules, liees entre elles par des ponts hydrogene. La presence des liaisons rend compte de l'elevation du point de fusion de la para-toluidine par rapport a celui des carbures benzeniques de symetrie et de poids moleculaires voisins. La direction des liaisons explique le clivage facile (001) des cristaux obtenus par fusion et l'anisotropie de croissance des cristaux en phase vapeur. Une methode de recherche de structures par ordinateur a ete elaboree, faisant usage du concept d'energie reticulaire applique aux molecules considerees comme rigides et faisant appel a une methode de simplexe pour la minimisation des fonctions sans calcul de derivees. La maniere dont est exploitee l'information disponible permet d'accro re, comme cela est montre dans le cas de la para-toluidine, l'etendue du domaine de convergence. (auteur)

Diffraction des neutrons : principe, dispositifs expérimentaux et applications

Science.gov (United States)

Muller, C.

2003-02-01

La diffraction de neutrons, sur monocristal ou sur échantillon polycristallin (ou poudre), est une technique très largement utilisée, en science des matériaux comme en biologie, lorsque l'on souhaite déterminer la structure cristalline d'un composé ou d'une molécule. Toutefois, le degré de précision de la détermination structurale est très corrélé au choix de l'instrument utilisé. Il s'en suit que la question “comment choisir l'instrument le mieux adapté au composé et à la problématique ?" apparaît comme fondamentale. L'objectif de ce cours est de tenter de répondre à cette question en décrivant brièvement les caractéristiques instrumentales de différents diffractomètres, en exposant les avantages spécifiques des expériences de diffraction de neutrons et en donnant quelques exemples d'application.

Development of a new time-amplitude converter with tunnel diodes for improving fast neutron spectrometry by time of flight; Realisation d'un nouveau convertisseur temps-amplitude a diodes ''tunnel'' ameliorant la spectrometrie des neutrons rapides par temps de vol

Energy Technology Data Exchange (ETDEWEB)

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

1963-11-15

New time-amplitude converter with Esaki diodes, the events being preselected before analysis, allows for realizing a fast neutron spectrometer by time-of-flight with an 1.5 * 10{sup -9} s overall time resolution for {sup 12}C (n,n') at 14 MeV. (author) [French] Realisation d'un convertisseur temps-amplitude a diodes 'tunnel', avec preselection des impulsions; l'application pour un spectrometre de neutrons rapides a temps-de-vol permet d'obtenir une resolution totale en temps de 1,5 nanoseconde dans la diffusion {sup 12}C (n,n') a 14 MeV. (auteur)

Calculation of the power factor using the neutron diffusion hybrid equation

International Nuclear Information System (INIS)

Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

2013-01-01

Highlights: ► A neutron diffusion hybrid equation with an external neutron source was used. ► Nodal expansion method to obtain the neutron flux was used. ► Nuclear power factors in each fuel element in the reactor core were calculated. ► The results obtained were very accurate. -- Abstract: In this paper, we used a neutron diffusion hybrid equation with an external neutron source to calculate nuclear power factors in each fuel element in the reactor core. We used the nodal expansion method to obtain the neutron flux for a given control rods bank position. The results were compared with results obtained for eigenvalue problem near criticality condition and fixed source problem during the start-up of the reactor, where external neutron sources are extremely important for the stabilization of external neutron detectors.

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)

The Biological Effect of Fast Neutrons and High-Energy Protons; Effets Biologiques des Neutrons Rapides et des Protons de Haute Energie; Biologicheskoe dejstvie bystrykh nejtronov i protonov vysokikh ehnergii; Efectos Biologicos de los Neutrones Rapidos y de los Protones de Elevada Energia

Energy Technology Data Exchange (ETDEWEB)

Moskalev, Ju. I.; Petrovich, I. K.; Strel' cova, V. N.

1964-03-15

tumour of the thyroid gland also appear. With proton irradiation, the minimum tumour dose for haemopoietic tissue and the thyroid gland is in the neighbourhood of 250 rad, for the testicles 300 rad, for the prostate, the pancreas and the subcutaneous tissue 100 rad. After irradiation with fast neutrons the incidence of tumours of the testicles, the large intestine, the kidneys and the liver is increased at doses between 42.5 and 85 rad. The same is true for tumours of the skin and the subcutaneous tissue at a dose of 85 rad. With fast neutron irradiation tumours appeared at approximately three to five times lower radiation doses than with proton irradiation. (author) [French] Le memoire expose les resultats d'experiences visant a comparer l'e ffe t des neutrons rapides et celui des protons de haute energie (500 MeV) sur la duree de vie, le sang peripherique, la frequence et la rapidite d'apparition des tumeurs chez le rat, en tenant compte de la dose efficace et du delai d'observation. Les travaux ont porte sur 573 rats blancs pour les neutrons et sur 490 rats blancs pour les protons. Les animaux etaient exposes a des doses de neutrons rapides de 8,5 a 510 rad et a des doses de protons de 28 a 1008 rad. Pour les neutrons rapides et les protons, les auteurs ont determine les valeurs des doses effectives aiguees, sub-aiguees et chroniques. La DL{sup 30}{sub 50} est de 408 rad pour les neutrons et de 600 rad pour les protons; la DL{sup 120}{sub 50} s'eleve a 380 et 600 rad, respectivement. On a procede a une etude critique des lois qui regissent la mortalite des rats a breve echeance et apres des delais prolonges. D'apres les constatations, la survie moyenne des rats irradies par des neutrons rapides est independante de leur sexe. En ce qui concerne les differents indices du sang peripherique, la forme de la courbe dose/effet depend dans une forte mesure, non seulement de la radiosensibilite des cellules correspondantes du sang, mais aussi de la duree d'observation. Elle

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

Energy Technology Data Exchange (ETDEWEB)

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

1955-07-01

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

Discrete formulation for two-dimensional multigroup neutron diffusion equations

Energy Technology Data Exchange (ETDEWEB)

Vosoughi, Naser E-mail: vosoughi@mehr.sharif.edu; Salehi, Ali A.; Shahriari, Majid

2003-02-01

The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method.

Discrete formulation for two-dimensional multigroup neutron diffusion equations

International Nuclear Information System (INIS)

Vosoughi, Naser; Salehi, Ali A.; Shahriari, Majid

2003-01-01

The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method

Experimental measurement of neutron spectrum in the reflector of a light water reactor; Determination experimentale du spectre des neutrons dans le reflecteur d'une pile a eau legere

Energy Technology Data Exchange (ETDEWEB)

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

1963-09-15

augmente, a ete mise en evidence. 3. Neutrons rapides: 3 detecteurs a seuil (Ni, Al, Mg) ont ete utilises. On a suppose au-dessus de 3 MeV une forme en E{sup 1/2} e{sup -{beta}}{sup E}. Les valeurs de {beta} varient de 0,775 au centre du coeur et dans un creneau, a 0,64 a environ 30 cm du coeur. 4. Trace continu du spectre d'ensemble: Les interpolations suivantes permettent un raccordement valable entre le domaine ou des mesures ont ete faites. entre 340 eV et 10 keV forme en 1/E. entre 10 keV et 330 keV forme en 1/(E {sigma}{sub S}(E)) ({sigma}{sub S}(E) = section de diffusion elastique sur l'hydrogene) entre 330 keV et 3 MeV: forme calculee par la methode des moments (ref. BSR). (auteur)

3-D anisotropic neutron diffusion in optically thick media with optically thin channels

International Nuclear Information System (INIS)

Trahan, Travis J.; Larsen, Edward W.

2011-01-01

Standard neutron diffusion theory accurately approximates the neutron transport process for optically thick, scattering-dominated systems in which the angular neutron flux is a weak (nearly linear) function of angle. Therefore, standard diffusion theory is not directly applicable for Very High Temperature Reactor (VHTR) cores, which contain numerous narrow, axially-oriented, nearly-voided coolant channels. However, we have derived a new, accurate diffusion equation for such problems, which contains nonstandard anisotropic diffusion coefficients near and within the channels, but which reduces to the standard diffusion approximation away from the channels. The new diffusion approximation significantly improves the accuracy of VHTR diffusion simulations, while having lower computational cost than higher-order transport methods. (author)

The neutronic method for measuring soil moisture; La methode neutronique de mesure de l'humidite des sols

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

The three group diffusion theory being chosen as the most adequate method for determining the response of the neutron soil moisture probe, a mathematical model is worked out using a numerical calculation programme with Fortran IV coding. This model is fitted to the experimental conditions by determining the effect of different parameters of measuring device: channel, fast neutron source, detector, as also the soil behaviour under neutron irradiation: absorbers, chemical binding of elements. The adequacy of the model is tested by fitting a line through the image points corresponding to the couples of experimental and theoretical values, for seven media having different chemical composition: sand, alumina, line stone, dolomite, kaolin, sandy loam, calcareous clay. The model chosen gives a good expression of the dry density influence and allows {alpha}, {beta}, {gamma} and {delta} constants to be calculated for a definite soil according to the following relation which gives the count rate of the soil moisture probe: N = ({alpha} {rho}{sub s} +{beta}) H{sub v} +{gamma} {rho}{sub s} + {delta}. (author) [French] La theorie a trois groupes etant choisie comme le moyen le plus adequat pour determiner la reponse de l'humidimetre, une mise en forme d'un modele mathematique est effectuee grace a un programme de calcul numerique ecrit en code Fortran IV. Cette representation est adaptee a l'experience par l'etude de l'influence de parametres du systeme de mesure: tubage, source, detecteur, et du comportement du sol vis-a-vis des neutrons: presence d'absorbeurs, liaison chimique des elements. La validite du modele est testee par ajustement d'une droite aux points images des couples de valeurs experimentales et theoriques, ceci pour sept milieux distincts definis par leur composition chimique: sable, alumine, calcaire, dolomie, kaolin, limon sableux, argile calcaire. Le modele prevoit, bien l'influence de la densite seche et permet de calculer pour un sol donne les constantes

Measurement of the temperature of the neutrons in reactor G1; Mesure de la temperature des neutrons dans la pile G1

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

A precise experimental method has been adapted to the analysis of the spectrum of neutrons in the thermal region. This method uses the technique of modulation applied to a beam of neutrons issuing from a characteristic point in the pile. The analysis of the spectrum is made by adjusting, by the method of least squares, an analytical form to the experimental results. In this report are given the results obtained with a beam from the centre of the moderator of G1. The spectrum of this beam essentially represents the spectrum of the neutrons in the moderator. The most probable velocity was determined by means of Maxwell's functions. The measurements were made of different moderator temperatures between 304 deg. K and 435 deg. K. (author) [French] Une methode experimentale precise a ete mise au point pour l'analyse du spectre des neutrons dans le domaine thermique. Cette methode utilise la technique de la modulation appliquee a un faisceau de neutrons issu d'un point caracteristique de la pile. L'analyse du spectre est faite en ajustant par la methode des moindres carres une forme analytique aux resultats experimentaux. Dans ce rapport, on donne les resultats obtenus sur un faisceau du centre du moderateur de G1. Le spectre de ce faisceau represente convenablement le spectre des neutrons dans le moderateur. On s'est limite ici a une fonction de Maxwell dont on a recherche la vitesse la plus probable. Les mesures ont ete faites avec une temperature du moderateur variant entre 304 deg. K et 435 deg. K. (auteur)

Pulsed Neutron Studies of BeO-Natural Uranium Lattices; Etudes de Reseaux Uranium Naturel-Glucine, par la Methode des Neutrons Pulses; Issledovanie reshetok BeO-prirodnyj uran s pomoshch'yu impul'snykh nejtronov; Estudio de Reticulados de Beo-Uranio Natural con Ayuda de Neutrones Pulsados

Energy Technology Data Exchange (ETDEWEB)

Joshi, B. V.; Nargundkar, V. R.; Subbarao, K. [Atomic Energy Establishment Trombay, Bombay (India)

1965-10-15

The use of the pulsed neutron method for determination of the diffusion parameters and multiplication constants of lattices is described. The diffusion parameters and multiplication constants of BeO-natural uranium lattices are obtained by this method. The uranium rods used were 2.92 cm diameter clad in aluminium 0.072 cm thick and were arranged vertically in channels of square cross-section 5 cm x 5 cm, in a square lattice of pitch 15 cm. The neutron bursts were produced from a cascade accelerator by pulsing the ion source and using the Be(d,n) reaction. The detectors were enriched boron trifluorfde proportional counters. The space and time harmonics were eliminated as described by the authors in another paper at this Symposium. The decay constant of the fundamental mode was determined for several bucklings, far from criticality. In these calculations the effect of diffusion cooling and anisotropy was taken into account. All the diffusion and multiplication constants were determined by the method of least squares fit by three different approaches. The prompt critical buckling was determined by solving the decay constant equation for {lambda} = 0. The final results are: Diffusion length L = 15.2 cm Absorption constant {lambda}{sub 0} = 762.5 s{sup -1}. Diffusion coefficient D = 1.761 x 10{sup 5} cm{sup 2}/s Diffusion cooling constant C = -8.766 x 10{sup 5} cm{sup 4}/s Age {tau} = 150.7 cm{sup -2} Infinite multiplication factor k{sub {infinity}} = 1.126 Prompt critical buckling B{sup 2} = 3.00 x 10{sup -4} cm{sup -2}. (author) [French] Les auteurs decrivent l'utilisation de la methode des neutrons puises pour la determination des parametres de diffusion et des constantes de multiplication des reseaux. Cette methode permet d'obtenir les parametres de diffusion et les constantes de multiplication du reseau uranium naturel-glucine. Les barres d'uranium utilisees avaient un diametre de 2,92 cm et un gainage'd'aluminium d'une epaisseur de 0,072 cm elles etaient disposees

Pulsed neutron method for diffusion, slowing down, and reactivity measurements

International Nuclear Information System (INIS)

Sjoestrand, N.G.

1985-01-01

An outline is given on the principles of the pulsed neutron method for the determination of thermal neutron diffusion parameters, for slowing-down time measurements, and for reactivity determinations. The historical development is sketched from the breakthrough in the middle of the nineteen fifties and the usefulness and limitations of the method are discussed. The importance for the present understanding of neutron slowing-down, thermalization and diffusion are point out. Examples are given of its recent use for e.g. absorption cross section measurements and for the study of the properties of heterogeneous systems

Size dependent diffusive parameters and tensorial diffusion equations in neutronic models for optically small nuclear systems

International Nuclear Information System (INIS)

Premuda, F.

1983-01-01

Two lines in improved neutron diffusion theory extending the efficiency of finite-difference diffusion codes to the field of optically small systems, are here reviewed. The firs involves the nodal solution for tensorial diffusion equation in slab geometry and tensorial formulation in parallelepiped and cylindrical gemometry; the dependence of critical eigenvalue from small slab thicknesses is also analitically investigated and finally a regularized tensorial diffusion equation is derived for slab. The other line refer to diffusion models formally unchanged with respect to the classical one, but where new size-dependent RTGB definitions for diffusion parameters are adopted, requiring that they allow to reproduce, in diffusion approach, the terms of neutron transport global balance; the trascendental equation for the buckling, arising in slab, sphere and parallelepiped geometry from the above requirement, are reported and the sizedependence of the new diffusion coefficient and extrapolated end point is investigated

Measurement of neutron diffusion length in heavy concrete

International Nuclear Information System (INIS)

Krejci, D.

2007-04-01

Using an aluminium sampler filled with heavy concrete the neutron diffusion length was determined, measuring thermal and fast neutrons over the whole beam hole with various threshold detectors using gold samples. These calculations should describe the neutron distribution in the whole concrete shield of the reactor and contribute to the investigation of the activation of the concrete shield using reactor parameters like operating time, power and neutron flux. Instrumentation, activation and positioning of the samples in the beam hole of the TRIGA Mark II reactor are described. (nevyjel)

Experiments on the thermalization of slow neutrons by liquid hydrogen (1962); Experience de thermalisation de neutrons lents par de l'hydrogene liquide (1962)

Energy Technology Data Exchange (ETDEWEB)

Cribier, D; Jacrot, B; Lacaze, A; Roubeau, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Institut Fourier, 38 - Grenoble (France)

1962-07-01

In order to increase the flux of neutrons of long wave-length ({lambda} > 4 A) emerging from a channel in the EL-3, a liquid hydrogen device was introduced into a channel of the reactor (Channel H{sub 1}). The principle of the device is simple. A volume of liquid hydrogen is introduced as close as possible to the reactor core into a region of intense isotropic flux. This hydrogen slows down the slow neutrons; because of the very small mean free diffusion path of slow in hydrogen, this slowing down is considerable even in a small volume of liquid hydrogen, and the spectrum temperature of neutrons emerging from the volume of liquid hydrogen can therefore be shifted. The intensity gain for neutrons with a wave length {lambda}, is a G ({lambda}) function which, for perfect thermalization and ignoring capture, is expressed by: G ({lambda}) = 225 exp (- 45.3/{lambda}{sup 2}), assuming a temperature of 300 deg. K for the neutrons before cooling and is 20 deg. K after cooling. For a wave-length of 5 A, the theoretical maximum gain of thus about 37. (authors) [French] Dans le but d'accroitre le flux des neutrons de grande longueur d'onde ({lambda} > 4 A) sortant d'un canal de la pile EL-3, un dispositif a hydrogene liquide a ete introduit dans un canal de la pile (canal H{sub 1}). Le principe du dispositif est simple. Un volume d'hydrogene liquide est introduit le plus pres possible du coeur de ia pile dans une region de flux intense et isotrope. Les neutrons lents sont ralentis par cet hydrogene; a cause du tres faible libre parcours moyen de diffusion des neutrons lents dans l'hydrogene, ce ralentissement est important meme dans un faible volume d'hydrogene liquide et l'on peut ainsi deplacer la temperature du spectre des neutrons sortant du volume d'hydrogene liquide. Le gain en intensite des neutrons de longueur d'onde {lambda} est une fonction G ({lambda}) qui pour une thermalisation parfaite et en negligeant la capture, s'exprime par: G ({lambda}) = 225 exp (- 45

Free diffusion of translation of macromolecules in solution with the rayleigh interferometer; Diffusion libre de translation des macromolecules en solution, par interferometrie de rayleigh

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

The aim of this study is to develop a rapid and accurate measurement, with the Rayleigh interferometer, of the free diffusion coefficient of translation of macromolecules in solution. After having explained the choice of a diffusion cell with laminar lateral flow, and explained the principle of the Rayleigh interferometer, a semi-automatic technique of free diffusion are then introduced. Solutions are proposed for systems composed of two or three components, such as biopolymers. The paper ends by drafting the possible treatment of recorded experimental data by means of electronic computer. (author) [French] Cette etude a ete entreprise pour mettre au point une methode precise et rapide de mesure, par interferometre de Rayleigh, du coefficient de diffusion libre de translation des macromolecules en solution. Apres avoir justifie le choix d'une cellule de diffusion a ecoulement laminaire lateral et explique le principe de l'interferometre de Rayleigh, l'auteur decrit une technique semi-automatique d'enregistrement des cliches d'interference. Il introduit ensuite les equations differentielles de diffusion libre et propose des solutions pour les systemes a deux et trois composants applicables aux biopolymeres. L'article se termine par une esquisse concernant le traitement des donnees experimentales enregistrees au moyen du calcul electronique. (auteur)

On the numerical solution of the neutron fractional diffusion equation

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2014-01-01

Highlights: • The new version of neutron diffusion equation which established on the fractional derivatives is presented. • The Neutron Fractional Diffusion Equation (NFDE) is solved in the finite differences frame. • NFDE is solved using shifted Grünwald-Letnikov definition of fractional operators. • The results show that “K eff ” strongly depends on the order of fractional derivative. - Abstract: In order to core calculation in the nuclear reactors there is a new version of neutron diffusion equation which is established on the fractional partial derivatives, named Neutron Fractional Diffusion Equation (NFDE). In the NFDE model, neutron flux in each zone depends directly on the all previous zones (not only on the nearest neighbors). Under this circumstance, it can be said that the NFDE has the space history. We have developed a one-dimension code, NFDE-1D, which can simulate the reactor core using arbitrary exponent of differential operators. In this work a numerical solution of the NFDE is presented using shifted Grünwald-Letnikov definition of fractional derivative in finite differences frame. The model is validated with some numerical experiments where different orders of fractional derivative are considered (e.g. 0.999, 0.98, 0.96, and 0.94). The results show that the effective multiplication factor (K eff ) depends strongly on the order of fractional derivative

Time-of-flight and vector polarization analysis for diffuse neutron scattering

International Nuclear Information System (INIS)

Schweika, W.

2003-01-01

The potential of pulsed neutron sources for diffuse scattering including time-of-flight (TOF) and polarization analysis is discussed in comparison to the capabilities of the present instrument diffuse neutron scattering at the research center Juelich. We present first results of a new method for full polarization analysis using precessing neutron polarization. A proposal is made for a new type of instrument at pulsed sources, which allows for vector polarization analysis in TOF instruments with multi-detectors

Neutron distribution in the central cell and a peripheral cell of the Fontenay-aux-Roses pile; Repartition des neutrons dans la cellule centrale et une cellule peripherique de la pile de Fontenay-aux-Roses

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The distribution of the neutron density has been determined in the central cell and a peripheral cell of the pile at Fontenay-aux-Roses. This measurement was carried out by the autoradiographic method with manganese detectors. The neutron density distribution in the uranium rod has already been studied. The measurement was completed by the study of the neutron density in the complete cell by means of detectors placed in the uranium and in the heavy water. (author) [French] La repartition de la densite des neutrons a ete determinee dans la cellule centrale et une cellule peripherique de la pile de Fontenay-aux-Roses. Cette mesure a ete effectuee par la methode d'autoradiographie avec des detecteurs de manganese. La repartition de la densite des neutrons dans la barre d'uranium a deja ete etudiee. La mesure a ete completee par l'etude de la densite des neutrons dans la cellule complete a l'aide de detecteurs places dans l'uranium et dans l'eau lourde. (auteur)

Fission ionisation chamber for the measurement of low fluxes of slow neutrons; Chambre d'ionisation a fission pour la mesure des faibles flux de neutrons lents

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The ionisation chamber described is designed for the measurement of slow neutron fluxes of average or low intensity, in the presence, eventually, of very high gamma fluxes. The capture of a slow neutron by a fissile material, in this case {sup 235}U, gives rise to fission fragments, high-energy particles which ionise the gas contained in the chamber. The neutrons are detected by virtue of the potential pulses, on the collecting electrode of the chamber, deriving from the collection of the ions produced by the fission fragments. The pulses are counted by means of a measuring system consisting of a preamplifier, a 2 Mc amplifier, a discriminator and an electronic scale with numerator or integrator. The general characteristics are as follows: sensitivity to neutrons: 0.07 kicks/n/cm{sup 2}.s, sensitivity to {gamma} rays: zero up to 3.10{sup 4} R/H, a background noise at the normal discrimination voltage: 0.01 kicks/s, working H.T.: -500 V, capacity: 40 {mu}{mu}F, average height of pulse: 8 mV, limits of use: from several neutrons to 10{sup 6} n/cm{sup 2}.s. This chamber may be used in all cases where low fluxes of slow neutrons must be measured, especially in the presence of high gamma fluxes, for example in the checking of Pu concentrations in an extraction plant or for the starting up of reactors. (author) [French] La chambre d'ionisation decrite est destinee a la mesure des flux de neutrons lents d'intensite moyenne ou faible, en presence eventuelle de flux gamma tres importants. La capture d'un neutron lent par un materiau fissile, en l'occurrence {sup 235}U, donne naissance a des fragments de fission, particules de grande energie qui ionisent le gaz contenu dans la chambre. Les neutrons sont detectes grace aux impulsions de potentiel, sur l'electrode collectrice de la chambre, provenant de la collection des ions produits par les fragments de fission. Une chaine de mesure comprenant un preamplificateur, un amplificateur 2 Mc, un discriminateur, une echelle

Research on GPU-accelerated algorithm in 3D finite difference neutron diffusion calculation method

International Nuclear Information System (INIS)

Xu Qi; Yu Ganglin; Wang Kan; Sun Jialong

2014-01-01

In this paper, the adaptability of the neutron diffusion numerical algorithm on GPUs was studied, and a GPU-accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. The IAEA 3D PWR benchmark problem was calculated in the numerical test. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. (authors)

Determination of neutron buildup factor using analytical solution of one-dimensional neutron diffusion equation in cylindrical geometry

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Julio Cesar L.; Vilhena, Marco Tullio, E-mail: julio.lombaldo@ufrgs.b, E-mail: vilhena@pq.cnpq.b [Universidade Federal do Rio Grande do Sul (DMPA/UFRGS), Porto Alegre, RS (Brazil). Dept. de Matematica Pura e Aplicada. Programa de Pos Graduacao em Matematica Aplicada; Borges, Volnei; Bodmann, Bardo Ernest, E-mail: bardo.bodmann@ufrgs.b, E-mail: borges@ufrgs.b [Universidade Federal do Rio Grande do Sul (PROMEC/UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

2011-07-01

The principal idea of this work, consist on formulate an analytical method to solved problems for diffusion of neutrons with isotropic scattering in one-dimensional cylindrical geometry. In this area were develop many works that study the same problem in different system of coordinates as well as cartesian system, nevertheless using numerical methods to solve the shielding problem. In view of good results in this works, we starting with the idea that we can represent a source in the origin of the cylindrical system by a Delta Dirac distribution, we describe the physical modeling and solved the neutron diffusion equation inside of cylinder of radius R. For the case of transport equation, the formulation of discrete ordinates S{sub N} consists in discretize the angular variables in N directions and in using a quadrature angular set for approximate the sources of scattering, where the Diffusion equation consist on S{sub 2} approximated transport equation in discrete ordinates. We solved the neutron diffusion equation with an analytical form by the finite Hankel transform. Was presented also the build-up factor for the case that we have neutron flux inside the cylinder. (author)

Determination of neutron buildup factor using analytical solution of one-dimensional neutron diffusion equation in cylindrical geometry

International Nuclear Information System (INIS)

Fernandes, Julio Cesar L.; Vilhena, Marco Tullio; Borges, Volnei; Bodmann, Bardo Ernest

2011-01-01

The principal idea of this work, consist on formulate an analytical method to solved problems for diffusion of neutrons with isotropic scattering in one-dimensional cylindrical geometry. In this area were develop many works that study the same problem in different system of coordinates as well as cartesian system, nevertheless using numerical methods to solve the shielding problem. In view of good results in this works, we starting with the idea that we can represent a source in the origin of the cylindrical system by a Delta Dirac distribution, we describe the physical modeling and solved the neutron diffusion equation inside of cylinder of radius R. For the case of transport equation, the formulation of discrete ordinates S N consists in discretize the angular variables in N directions and in using a quadrature angular set for approximate the sources of scattering, where the Diffusion equation consist on S 2 approximated transport equation in discrete ordinates. We solved the neutron diffusion equation with an analytical form by the finite Hankel transform. Was presented also the build-up factor for the case that we have neutron flux inside the cylinder. (author)

Comparison of fission probabilities with emission of long range particles under the action of slow and fast neutrons on various materials; Probabilites comparees de fission avec emission de particules de long parcours pour divers materiaux sous l'action des neutrons lents et rapides

Energy Technology Data Exchange (ETDEWEB)

Netter, F; Faraggi, H; Garin-Bonnet, A; Julien, J; Corge, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Turkiewicz, J [Institut de Recherches Nucleaire de Varsovie (Poland)

1958-07-01

The authors describe relative cross-section measurements of fission of the isotopes of uranium and plutonium (more particularly {sup 235}U, {sup 238}U, {sup 239}Pu), with emission of long range particles, under the action of neutrons of various energies: thermal neutrons, pile neutrons, neutrons produced with the Van de Graaff accelerator by reaction of protons on tritium. The measurements are carried out: 1) with the aid of photographic plates, by submitting to the action of the neutrons a layer of fissile material coupled with an Ilford nuclear emulsion of 200 microns; a tin sheet laying between the plate and the layer stops the {alpha} particles and the fission fragments. By an appropriate development the tracks of the long range particles can be distinguished in the emulsion, from the tracks of the recoil protons resulting of fission neutrons, or of the last primary neutrons. For neutrons of energy under 1 MeV, the compared frequency of the tracks of long range particles and of the recoils caused by the fission neutrons gives a measurement of the fission cross-section with emission of long range particles relative to the product of the fission cross-section by the mean number of neutrons emitted by fission. For neutrons of higher energy, one measures only the frequency of the tracks of long range particles, comparatively with the flux of primary neutrons. Some precautions are taken to eliminate the action of thermal neutrons in the measurements with fast neutrons. 2) with the aid of a system of ionization chamber and proportional counter, the rate of coincidence between the impulsions caused by the long range particles and the impulsions provided by one of the fission fragments is measured comparatively with the counting rate of fission fragme (author) [French] Les auteurs decrivent des mesures relatives a la section efficace de fission des isotopes de l'uranium et du plutonium (notamment {sup 235}U, {sup 238}U, {sup 239}Pu) avec emission de particules de long

Dynamics of water and ions in clays of type montmorillonite by microscopic simulation and quasi-elastic neutron scattering; Dynamique de l'eau et des ions dans des argiles de type montmorillonite par simulation microscopique et diffusion quasi-elastique des neutrons

Energy Technology Data Exchange (ETDEWEB)

Malikova, N

2005-09-15

Montmorillonite clays in low hydration states, with Na{sup +} and Cs{sup +} compensating counter ions, are investigated by a combination of microscopic simulation and quasi-elastic neutron scattering to obtain information on the local structure and dynamics of water and ions in the interlayer. At first predictions of simulation into the dynamics of water and ions at elevate temperatures are shown (0 deg C 80 deg C, pertinent for the radioactive waste disposal scenario) Marked difference is observed between the modes of diffusion of the Na{sup +} and C{sup +} counter ions. In water dynamics, a significant step towards bulk water behaviour is seen on transition from the mono- to bilayer states. Secondly, a detailed comparison between simulation and quasi-elastic neutron scattering (Neutron Spin Echo and Time-of-Flight) regarding ambient temperature water dynamics is presented. Overall, the approaches are found to be in good agreement with each other and limitations of each of the methods are clearly shown. (author)

Calculation of correlation between spins in a magnetic substance; Calcul des correlations entre spins dans une substance magnetique

Energy Technology Data Exchange (ETDEWEB)

Gennes, P.G. de [Commissariat a l' Energie Atomique, Saclay (France)

1959-07-01

The report states an elementary calculation of the correlation between spins in a magnetic substance, and particularly of their asymptotic form with regard to relatively wide-spaced spins. This permits the determination of the phenomenological parameters introduced by Var Hove to describe the magnetic scatter of neutrons in the critical opalescent area. (author) [French] On donne un calcul elementaire des correlations entre spins dans une substance magnetique, et notamment de leur forme asymptotique pour des spins assez eloignes. Ceci permet de determiner les parametres phenomenologiques introduits par Van Hove pour decrire la diffusion magnetique des neutrons dans la region d'opalescence critique. (auteur)

Fast solution of neutron diffusion problem by reduced basis finite element method

International Nuclear Information System (INIS)

Chunyu, Zhang; Gong, Chen

2018-01-01

Highlights: •An extremely efficient method is proposed to solve the neutron diffusion equation with varying the cross sections. •Three orders of speedup is achieved for IAEA benchmark problems. •The method may open a new possibility of efficient high-fidelity modeling of large scale problems in nuclear engineering. -- Abstract: For the important applications which need carry out many times of neutron diffusion calculations such as the fuel depletion analysis and the neutronics-thermohydraulics coupling analysis, fast and accurate solutions of the neutron diffusion equation are demanding but necessary. In the present work, the certified reduced basis finite element method is proposed and implemented to solve the generalized eigenvalue problems of neutron diffusion with variable cross sections. The order reduced model is built upon high-fidelity finite element approximations during the offline stage. During the online stage, both the k eff and the spatical distribution of neutron flux can be obtained very efficiently for any given set of cross sections. Numerical tests show that a speedup of around 1100 is achieved for the IAEA two-dimensional PWR benchmark problem and a speedup of around 3400 is achieved for the three-dimensional counterpart with the fission cross-sections, the absorption cross-sections and the scattering cross-sections treated as parameters.

Pulsed neutron determination of anisotropic diffusion constants in multi-layered slabs

International Nuclear Information System (INIS)

Sri Ram, K.

1978-01-01

Anisotropic neutron diffusion parameters for graphite and plexiglas slab assemblies were calculated using one-dimensional discrete ordinates code ANISN, and also Case's eigenfunction expansion technique as suggested by Leonard. These calculated values were checked with the pulsed neutron experimental results as well as simple diffusion theory calculations of Spinrad. Relatively little experimental work has been done with heterogeneous assemblies which do not contain voids. The present comparison shows that the experimental results agree well with transport theory calculations. It appears from the results and inter-comparison of this work in simple geometries, that the pulsed neutron method can yield accurate experimental anisotropic diffusion constants, and can therefore be applied to more complicated geometries which may be difficult to calculate. (author)

Enhanced finite difference scheme for the neutron diffusion equation using the importance function

International Nuclear Information System (INIS)

Vagheian, Mehran; Vosoughi, Naser; Gharib, Morteza

2016-01-01

Highlights: • An enhanced finite difference scheme for the neutron diffusion equation is proposed. • A seven-step algorithm is considered based on the importance function. • Mesh points are distributed through entire reactor core with respect to the importance function. • The results all proved that the proposed algorithm is highly efficient. - Abstract: Mesh point positions in Finite Difference Method (FDM) of discretization for the neutron diffusion equation can remarkably affect the averaged neutron fluxes as well as the effective multiplication factor. In this study, by aid of improving the mesh point positions, an enhanced finite difference scheme for the neutron diffusion equation is proposed based on the neutron importance function. In order to determine the neutron importance function, the adjoint (backward) neutron diffusion calculations are performed in the same procedure as for the forward calculations. Considering the neutron importance function, the mesh points can be improved through the entire reactor core. Accordingly, in regions with greater neutron importance, density of mesh elements is higher than that in regions with less importance. The forward calculations are then performed for both of the uniform and improved non-uniform mesh point distributions and the results (the neutron fluxes along with the corresponding eigenvalues) for the two cases are compared with each other. The results are benchmarked against the reference values (with fine meshes) for Kang and Rod Bundle BWR benchmark problems. These benchmark cases revealed that the improved non-uniform mesh point distribution is highly efficient.

Thermal, intermediate and fast neutron flux measurements using activation detectors; Mesure des flux de neutrons thermiques, intermediaires et rapides au moyen de detecteurs par activation

Energy Technology Data Exchange (ETDEWEB)

Brisbois, J; Lott, M; Manent, G [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The problem of neutron flux measurements using activation detectors is studied in the particular case of protection research. It is shown how it possible, it is possible, using a known thermal flux, to organise a coherent calibration system for all the detectors. The rapid neutron detectors are calibrated with respect to a reference detector (phosphorus) in a natural uranium converter; the intermediate neutron detectors with respect to gold in the axial channel of ZOE. This method makes it possible to minimise the errors due to the activation cross-sections. A brief description is given of the counting room of the Pile Safety Study Service, as well of the practical utilisation characteristics of the counters employed. (authors) [French] Le probleme de la mesure des flux de neutrons au moyen de detecteurs par activation est etudie dans le cas particulier des etudes de protections. On montre comment, a partir d'un flux thermique connu, on peut organiser un systeme coherent d'etalonnage de tous les detecteurs. Les detecteurs de neutrons rapides sont etalonnes par rapport a un detecteur de reference (phosphore) dans un convertisseur en uranium naturel; les detecteurs de neutrons intermediaires, par rapport a l'or dans le canal axial de ZOE, Cette methode permet de minimiser les erreurs dues aux sections efficaces d'activation. On decrit sommairement la salle de comptage du Service d'Etudes de Protections de Piles et on indique les caracteristiques d'emploi pratique des detecteurs utilises. (auteurs)

Statistical fluctuations of the number of neutrons in a pile; Fluctuations statistiques du nombre de neutrons dans une pile

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The theory of the statistical fluctuations in a pile is extended to the space dependent case, and gives the fluctuations of the number of neutrons in a cell of the core or reflector of the pile. This number changes through elementary processes occurring at random, which are, capture, source, fission and scattering. Of all these processes, fission is the only one which changes more than one neutron at a time and so is responsible of the deviation of the fluctuations from a Poisson law. The importance of this deviation depends on the dimensions of the cell compared to the slowing down length. When the dimensions are small, the fluctuations close to a Poisson law. (author) [French] La theorie des fluctuations statistiques est etendue au cas local et donne les fluctuations du nombre de neutrons dans une cellule situee dans le coeur ou le reflecteur de la pile. Ce nombre evolue au cours du temps sous l'influence de phenomenes aleatoires qui sont la capture, la diffusion, les sources et les neutrons secondaires de fission. L'emission simultanee de plusieurs neutrons distingue ce phenomene des precedents qui n'affectent qu'un neutron individuellement. L'importance de ce phenomene sur la loi de fluctuation depend des dimensions de la cellule par rapport a la longueur de ralentissement. Quand ces dimensions sont petites, le caractere particulier de ce phenomene disparait. (auteur)

Thermal Neutron Spectral and Spatial Distributions in Light-Water-Moderated Uranium Lattices; Distributions Spectrale et Spatiale des Neutrons Thermiques dans des Reseaux a Uranium et a Eau Leger; Spektral'noe i prostranstvennoe raspredelenie teplovykh nejtronov v uranovykh reshetkakh s vodnym zamedlitelem; Distribuciones Espectral y Espacial de los Neutrones Termicos en los Reticulados de Uranio Moderados por Agua Ligera

Energy Technology Data Exchange (ETDEWEB)

Hardy, J.; Volpe, J. J.; Klein, D.; Schmidt, E.; Gelbard, E. [Bettis Atomic Power Laboratory, Pittsburgh, PA (United States)

1964-02-15

distorted by cadmium poison. Calculations with each kernel were done, for this experiment, in a 12-energy group, P{sub 3} approximation. The slowing-down source shape was obtained by means of an eigenvalue calculation of the spatial modal shapes, decaying exponentially in lethargy, which are sustained by the system, supplemented by measured epicadmium activation shapes. Between the poisoned and unpoisoned cases, calculated relative activation rates predicted by the two kernels differed by as much as 10%. Comparisons of the experimental results with calculations using the Radkowsky kernel and the Nelkin kernel were made. Results strongly favour the harder spectrum predicted by the Nelkin kernel. (author) [French] Les auteurs passent en revue certaines experiences sur le comportement des neutrons thermiques dans les reseaux a uranium et a eau legere. Les experiences sont de deux types principaux: mesures de l'utilisation thermique et mesures par activation de la distribution spatiale des neutrons thermiques dans un milieu aqueux uniformement empoisonne. Ces experiences ont ete concues pour verifier la validite des modeles theoriques actuels en separant les effets spectraux des effets spatiaux a l'interieur d'un reacteur. On a procede a des comparaisons avec des modeles theoriques impliquant differentes approximations touchant la theorie du transport et le noyau de diffusion des neutrons thermiques. La premiere serie d'experiences a consiste en mesures par activation des facteurs de desavantage pour le flux thermique dans differentes cellules du reseau de l'assemblage TRX, qui est un reacteur a uranium faiblement enrichi et a eau legere et dont les barreaux de combustible cylindriques sont disposes en un faisceau hexagonal. Les auteurs ont compare les mesures de l'utilisation thermique avec les resultats de calculs faits par une methode de Monte-Carlo. Ils ont fait usage du noyau de Radkowsky et du noyau de Nelkin. Le noyau de Radkowsky est une approximation de premier ordre

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Solution of two energy-group neutron diffusion equation by triangular elements

International Nuclear Information System (INIS)

Correia Filho, A.

1981-01-01

The application of the triangular finite elements of first order in the solution of two energy-group neutron diffusion equation in steady-state conditions is aimed at. The EFTDN (triangular finite elements in neutrons diffusion) computer code in FORTRAN IV language is developed. The discrete formulation of the diffusion equation is obtained applying the Galerkin method. The power method is used to solve the eigenvalues' problem and the convergence is accelerated through the use of Chebshev polynomials. For the equation systems solution the Gauss method is applied. The results of the analysis of two test-problems are presented. (Author) [pt

Procedure for obtaining neutron diffusion coefficients from neutron transport Monte Carlo calculations (AWBA Development Program)

International Nuclear Information System (INIS)

Gast, R.C.

1981-08-01

A procedure for defining diffusion coefficients from Monte Carlo calculations that results in suitable ones for use in neutron diffusion theory calculations is not readily obtained. This study provides a survey of the methods used to define diffusion coefficients from deterministic calculations and provides a discussion as to why such traditional methods cannot be used in Monte Carlo. This study further provides the empirical procedure used for defining diffusion coefficients from the RCP01 Monte Carlo program

Three-group albedo method applied to the diffusion phenomenon with up-scattering of neutrons

International Nuclear Information System (INIS)

Terra, Andre M. Barge Pontes Torres; Silva, Jorge A. Valle da; Cabral, Ronaldo G.

2007-01-01

The main objective of this research is to develop a three-group neutron Albedo algorithm considering the up-scattering of neutrons in order to analyse the diffusion phenomenon in nonmultiplying media. The neutron Albedo method is an analytical method that does not try to solve describing explicit equations for the neutron fluxes. Thus the neutron Albedo methodology is very different from the conventional methodology, as the neutron diffusion theory model. Graphite is analyzed as a model case. One major application is in the determination of the nonleakage probabilities with more understandable results in physical terms than conventional radiation transport method calculations. (author)

Determination of thermal neutrons diffusion length in graphite

International Nuclear Information System (INIS)

Garcia Fite, J.

1959-01-01

The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs

Application of the pulsed neutron technique on the reactors ALIZE - AQUILON (1963); Application de la methode des neutrons pulses sur les piles ALIZE et AQUILON (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

Different methods of measuring the ratio effective delayed fraction / prompt neutron lifetime, {alpha}{sub c}, are described. According to the classic pulsed neutron technique the negative reactivity due to a localized absorber is given by {rho} / {beta}{sub eff} = {alpha} / {alpha}{sub c} -1 Experiments are reported which show that in this case {alpha}{sub c} can not be considered constant for large reactivities. The absorber element distorts the flux in the system, increasing the importance of the reflector. An application of the pulsed neutron method to the measurement of critical distributed boron concentrations of various absorber elements is described. Less time is required than for the usual super-critical techniques, and the experimental analysis is simplified. It is interesting to note that the results are not influenced by the spectral sensitivity of the control element. A modified pulsed neutron method has been tried out. This procedure was used to determine by measurements at sub-critical the critical water level of uranium-heavy water lattices with a high precision. (author) [French] Differents modes operatoires pour definir la valeur du rapport pourcentage effectif de neutrons retardes / temps de vie, {alpha}{sub c}, sont exposes. La methode classique par neutrons pulses definit l'anti-reactivite d'un element absorbant a partir de la relation: {rho} / {beta}{sub eff} {alpha} / {alpha}{sub c} -1 Les manipulations effectuees montrent qu'on ne peut considerer dans ce cas {alpha}{sub c} constant pour de tres grandes anti-reactivites. L'absorbant introduit dans la pile deforme le flux et augmente l'importance du reflecteur. Une application de la methode des neutrons pulses pour mesurer le titre critique en mg de B/l de divers absorbants est signalee. Les operations sont effectuees en regime sous-critique avec un certain gain de temps et une grande facilite de depouillement. Il est interessant de noter que les resultats ne sont pas affectes par la

Application of the pulsed neutron technique on the reactors ALIZE - AQUILON (1963); Application de la methode des neutrons pulses sur les piles ALIZE et AQUILON (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

Different methods of measuring the ratio effective delayed fraction / prompt neutron lifetime, {alpha}{sub c}, are described. According to the classic pulsed neutron technique the negative reactivity due to a localized absorber is given by {rho} / {beta}{sub eff} = {alpha} / {alpha}{sub c} -1 Experiments are reported which show that in this case {alpha}{sub c} can not be considered constant for large reactivities. The absorber element distorts the flux in the system, increasing the importance of the reflector. An application of the pulsed neutron method to the measurement of critical distributed boron concentrations of various absorber elements is described. Less time is required than for the usual super-critical techniques, and the experimental analysis is simplified. It is interesting to note that the results are not influenced by the spectral sensitivity of the control element. A modified pulsed neutron method has been tried out. This procedure was used to determine by measurements at sub-critical the critical water level of uranium-heavy water lattices with a high precision. (author) [French] Differents modes operatoires pour definir la valeur du rapport pourcentage effectif de neutrons retardes / temps de vie, {alpha}{sub c}, sont exposes. La methode classique par neutrons pulses definit l'anti-reactivite d'un element absorbant a partir de la relation: {rho} / {beta}{sub eff} {alpha} / {alpha}{sub c} -1 Les manipulations effectuees montrent qu'on ne peut considerer dans ce cas {alpha}{sub c} constant pour de tres grandes anti-reactivites. L'absorbant introduit dans la pile deforme le flux et augmente l'importance du reflecteur. Une application de la methode des neutrons pulses pour mesurer le titre critique en mg de B/l de divers absorbants est signalee. Les operations sont effectuees en regime sous-critique avec un certain gain de temps et une grande facilite de depouillement. Il est interessant de noter que les resultats ne sont pas

Current trends in methods for neutron diffusion calculations

International Nuclear Information System (INIS)

Adams, C.H.

1977-01-01

Current work and trends in the application of neutron diffusion theory to reactor design and analysis are reviewed. Specific topics covered include finite-difference methods, synthesis methods, nodal calculations, finite-elements and perturbation theory

Nodal spectrum method for solving neutron diffusion equation

International Nuclear Information System (INIS)

Sanchez, D.; Garcia, C. R.; Barros, R. C. de; Milian, D.E.

1999-01-01

Presented here is a new numerical nodal method for solving static multidimensional neutron diffusion equation in rectangular geometry. Our method is based on a spectral analysis of the nodal diffusion equations. These equations are obtained by integrating the diffusion equation in X, Y directions and then considering flat approximations for the current. These flat approximations are the only approximations that are considered in this method, as a result the numerical solutions are completely free from truncation errors. We show numerical results to illustrate the methods accuracy for coarse mesh calculations

Diffusion theory model for optimization calculations of cold neutron sources

International Nuclear Information System (INIS)

Azmy, Y.Y.

1987-01-01

Cold neutron sources are becoming increasingly important and common experimental facilities made available at many research reactors around the world due to the high utility of cold neutrons in scattering experiments. The authors describe a simple two-group diffusion model of an infinite slab LD 2 cold source. The simplicity of the model permits to obtain an analytical solution from which one can deduce the reason for the optimum thickness based solely on diffusion-type phenomena. Also, a second more sophisticated model is described and the results compared to a deterministic transport calculation. The good (particularly qualitative) agreement between the results suggests that diffusion theory methods can be used in parametric and optimization studies to avoid the generally more expensive transport calculations

EFFET DU GETTERING SUR LA LONGUEUR DE DIFFUSION DES PLAQUETTES DE SILICIUM SEMICRISTALLIN "SOPLIN"

Directory of Open Access Journals (Sweden)

A MOUHOUB

2003-06-01

Full Text Available L’effet des diffusions du phosphore à haute température (T = 900°C sur la longueur de diffusion (Ln des plaquettes de silicium semicristallin SOPLIN (SOlidification by PLanar INterface est déterminé par la mesure de la réponse spectrale de la photopile réalisée sur le matériau traité. L’influence de ces recuits sur les paramètres photovoltaïques est enregistrée dans les caractéristiques courant - tension sous éclairement simulé et la réponse spectrale.

The epithermal critical experiments; Experiences critiques avec des neutrons epitliermiques; Nadteplovye kriticheskie ehksperimenty; Experimentos criticos con neutrones epitermicos

Energy Technology Data Exchange (ETDEWEB)

Morewitz, H A; Carpenter, S O [Atomics International, Canoga Park, CA (United States)

1962-03-15

The epithermal critical experiments. The present phase of the advanced-epithermal-thorium-reactor programme consists of integral-reactor-physic s experiments designed to provide neutron-cross-section information in the 10-MeV to 1-keV range. A series of nine, multi-region, slow-fast, pseudospherica l critical assemblies of the honey- comb, split-table type are being studied. So far, three assemblies have' been run. The outer driver-decouple r region drives an interior U{sup 233}-Th fuelled spherical test region whose neutron-flux spectrum is successively degraded by increasing the graphite moderator to fuel ratio. A square-wave oscillator experiment defines the central reactivity worths of forty small samples of different materials to 10{sup -8} {Delta}k for each assembly. Additionally, intercalibrated artificial neutron sources are oscillated to determine the various central neutron importance functions. The spectra are obtained by fission-counter measurements with calibrated foils of different thresholds and by a Li{sup 6}-solid-state- counter sandwich spectrometer. A digital computer routine will be used to compile all measurements into a self-consistent library of spectrum averaged cross-sections. (author) [French] La phase actuelle du programme de reacteur au thorium a neutrons epithermiques comprend des experiences integrales de physique des reacteurs pour obtenir des renseignements sur les sections efficaces neutroniques pour la gamme d'energie comprise entre 1 keV et 10 MeV. Les auteurs etudient une serie de neuf ensembles critiques pseudospheriques, a plusieurs regions, a couplage neutrons lents et neutrons rapides du type a alveoles et a coeur divise. A ce jour, trois de ces ensembles ont ete mis en service. La region exterieure, mettant en service ou hors service, commande une zone d'essai interieure de forme spherique ou le combustible est constitue de {sup 233}U-Th, dont le spectre du flux de neutrons est degrade progressivement par augmentation du

Diffusion Parameters of BeO by the Pulsed Neutron Method

International Nuclear Information System (INIS)

Joshi, B.V.; Nargundkar, V.R.; Subbarao, K.

1965-01-01

The use of the pulsed neutron method for the precise determination of the diffusion parameters of moderators is described. The diffusion parameters of BeO have been obtained by this method. The neutron bursts were produced from a cascade accelerator by pulsing the ion source and using the Be (d, n) reaction. The detector was an enriched boron trifluoride proportional counter. It is shown that by a proper choice of the counter position arid length, and the source position, most of the space harmonics can be eliminated. Any constant background can be accounted for in the calculation of the decay constant. Very large bucklings were not used to avoid time harmonics. Any remaining harmonic content was rendered ineffective by the use of adequate time delay. The decay constant of the fundamental mode of the thermal neutron population was determined for several bucklings. Conditions to be satisfied for an accurate determination of the diffusion cooling constant C are discussed. The following values are obtained for BeO: λ 0 = absorption constant = 156.02 ± 4.37 s -1 D = diffusion coefficient = (1.3334 ± 0.0128) x 10 5 cm 2 /s C = diffusion cooling constant = (-4.8758 ± 0.5846) x 10 5 cm 4 /s. The effect of neglecting the contribution of the B 6 term on the determination of the diffusion parameters was estimated and is shown to be considerable. The reason for the longstanding discrepancy between the values of C obtained for the same moderator by different workers is attributed to this. (author) [fr

Economic study of an installation for uranium isotope separation by gaseous diffusion; Etude economique d'une installation de separation des isotopes de l'uranium par diffusion gazeuse

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

This report describes the major problems which arise in the choice of characteristics required in a gaseous diffusion installation for the separation of uranium isotopes. This choice depends largely on economic evaluations, and also on considerations of simplicity. The choice of working pressures and of the characteristics of the membrane are described, as are the possible alternatives regarding the structure of the stages and the problems of control. (author) [French] Ce rapport decrit les problemes majeurs qui se posent dans le choix des caracteristiques d'une installation de diffusion gazeuse destinee a la separation des isotopes de l'uranium. Ce choix depend en grande partie d'evaluations economiques et repose egalement sur des considerations de simplicite. On decrit ainsi le choix des pressions d'operation, celui des caracteristiques de la barriere, les alternatives possibles concernant la structure des etages et les problemes de regulation. (auteur)

The Pulsed Neutron Technique Applied to Fast Non-Multiplying Assemblies; Application de la Methode des Neutrons Pulses aux Assemblages Non Multiplicateurs a Neutrons Rapides; Primenenie metoda impul'snykh nejtronov pri izuchenii povedeniya bystrykh nejtronov v nerazmnozhayushchikh sborkakh; Aplicacion de la Tecnica de los Neutrones Pulsados a Conjuntos Rapidos de Materiales No Multiplicadores

Energy Technology Data Exchange (ETDEWEB)

Beghian, L. E.; Wilensky, S. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1965-10-15

Graaff a impulsions de l'ordre de la nanoseconde pour etudier le comportement de neutrons rapides dans des assemblages metalliques non multiplicateurs. Ils ont mesure les sections efficaces non elastiques du fer pour les neutrons rapides. La methode employee est semblable a celle qu'on utilise pour mesurer les sections efficaces d'absorption dans des assemblages a neutrons thermiques, avec cette difference que les temps de decroissance pour les neutrons rapides sont de l'ordre de la nanoseconde et non de la microseconde. Des bouffees de neutrons monoenergetiques d'une nanoseconde sont introduites dans des assemblages en fer de dimensions diverses. On remarque que dans ces assemblages le flux de neutrons decroit exponentiellement avec une constante de decroissance caracteristique. La constante de decroissance est composee d'une somme de termes qui representent la perte de neutrons due aux fuites et a la degradation en energie. La degradation en energie represente une perte de neutrons, car un detecteur de neutrons dont les donnees sont entachees d'une erreur systematique est utilise. On peut determiner le terme correspondant a la diffusion elastique et inelastique en mesurant la constante de decroissance consideree comme une fonction des dimensions de l'assemblage. Suit un developpement theorique pour le calcul de la fraction de ce terme qui est due a la diffusion elastique et, par consequent, pour la determination de la section efficace non elastique. Le procede theorique de calcul de cette fraction a fait l'objet de verifications experimentales. Les auteurs ont mesure la section efficace non elastique du fer par cette methode pour des energies de neutrons primaires comprises entre 0,8 et 1,5 MeV. La methode de la source puisee decrite ci-dessus a ete utilisee pour mesurer des constantes de decroissance pour des plaques de plomb. Les conditions de l'experience sont approximativement conformes aux hypotheses generalement admises lorsqu'on resout l'equation de transport de

Monte-Carlo method for studying the slowing down of neutrons in a thin plate of hydrogenated matter; Methode de Monte-Carlo pour l'etude du ralentissement des neutrons dans une plaque mince de matiere hydrogenee

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

The studies of interaction of slow neutrons with atomic nuclei by means of the time of flight methods are made with a pulsed neutron source with a broad energy spectrum. The measurement accuracy needs a high intensity and an output time as short as possible and well defined. If the neutrons source is a target bombarded by the beam of a pulsed accelerator, it is usually required to slow down the neutrons to obtain a sufficient intensity at low energies. The purpose of the Monte-Carlo method which is described in this paper is to study the slowing down properties, mainly the intensity and the output time distribution of the slowed-down neutrons. The choice of the method and parameters studied is explained as well as the principles, some calculations and the program organization. A few results given as examples were obtained in the line of this program, the limits of which are principally due to simplifying physical hypotheses. (author) [French] l'etude de l'interaction des neutrons lents avec les noyaux atomiques par la methode du temps de vol s'effectue avec une source pulsee de neutrons dont le spectre en energie est assez etendu. La precision des mesures demande que la source soit intense et que la duree d'emission des neutrons soit breve et bien definie. Si la source est une cible bombardee par le faisceau de particules d'un accelerateur pulse, il est generalement indispensable de ralentir les neutrons pour avoir une intensite suffisante a basse energie. Nous presentons ici une methode de Monte-Carlo pour l'etude detaillee de ce ralentissement, notamment l'intensite et la distribution des temps de sortie des neutrons ralentis. Cette presentation comprend: la justification du choix de la methode de Monte-Carlo, les principes generaux, les differentes etapes du calcul et du programme ecrit pour le calculateur electronique IBM 7090. Nous indiquons aussi les restrictions qui sont apportees au domaine d'application de ce programme et qui proviennent surtout des

Results in pion proton scattering near the higher resonances (1961); Resultats pour la diffusion des mesons pi par les protons dans le domaine des hautes resonances (1962)

Energy Technology Data Exchange (ETDEWEB)

Falk-Vairant, P; Valladas, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

We present briefly the available Information on the total cross sections for pion proton scattering in the energy region from 400 MeV to 1.5 GeV. We also have collected all results on total cross sections for particular channels like elastic scattering, inelastic scattering and charge exchange. Using new results on the total cross section for neutral events, we have plotted separately the cross section for elastic and for inelastic scattering in the T = 1/2 state. (authors) [French] On presente brievement les donnees connues concernant la section efficace totale pour la diffusion des mesons pi par les protons dans le domaine d'energie de 400 MeV a 1,5 GeV. On a egalement rassemble tous les resultats concernant les sections efficaces totales pour des canaux particuliers: diffusion elastique, diffusion inelastique et echange de charge. En partant des nouveaux resultats sur la section efficace pour la diffusion elastique et inelastique dans l'etat T = 1/2. (auteurs)

Neutron thermalization and spectra; Thermalisation et spectres de neutrons

Energy Technology Data Exchange (ETDEWEB)

Cadilhac, M; Soule, J L; Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

interpretation of experiments (spectral indices critical experiments on plutonium, diffusion parameters). (authors) [French] L'etude des spectres de neutrons regnant dans les reacteurs thermiques a pris une importance croissante en raison du role qu'y joue le plutonium. Quelle que soit par ailleurs la loi d'absorption, on constate que la loi de diffusion du moderateur n'agit sur le spectre que par certaines proprietes globales. Il est donc a priori possible de mettre au point une representation simplifiee de cette action, qui conduise a une comprehension claire des phenomenes tout en reduisant le volume des calculs numeriques necessaires. Le modele synthetique adopte par les auteurs presente l'avantage de ramener la determination des-spectres en milieu homogene a la resolution d'une equation differentielle du second ordre, comme le modele de Wigner-Wilkins (hydrogene gazeux monoatomique) et le modele gaz lourd generalise de J. Horowitz, qui en sont d'ailleurs l'un et l'autre des cas particuliers. Mais il est d'autre part assez general pour permettre de traiter correctement les situations rencontrees en pratique et en particulier le cas important ou la presence de plutonium fait apparaitre des resonances d'absorption a basse energie. Les liaisons chimiques ou cristallines du moderateur interviennent en effet, dans le modele propose, par l'intermediaire de deux fonctions de l'energie. Ces fonctions ont ete ajustees, pour les moderateurs usuels (Graphite. Eau lourde. Eau legere), a partir de lois de diffusion theoriques connues. En milieu heterogene, il importe avant tout de bien connaitre le spectre moyen dans le combustible d'une cellule ce que permet une generalisation de la methode d'Amouyal-Benoist-Horowitz. Le modele presente se prete particulierement bien a ces calculs et permet de traiter egalement les effets de rethermalisation (lorsque par exemple le refroidisseur est a une temperature differente de celle du moderateur). On donne enfin des exemples d'utilisation pratique

OGM Diffusion des OGM et impact pour les agriculteurs : quels sont les liens ?

Directory of Open Access Journals (Sweden)

Lemarie Stéphane

2001-05-01

Full Text Available Les OGM actuellement sur le marché présentent des innovations sur des fonctions de protection des plantes. Selon les problèmes de protection des plantes, les gains des agriculteurs liés à l’adoption des OGM sont très variables. Cette hétérogénéité pose d’importants problèmes méthodologiques pour pouvoir estimer correctement la distribution des gains pour l’ensemble des agriculteurs. Une synthèse de plusieurs études américaines est réalisée ici et met bien en évidence la diversité des méthodes utilisées, chacune d’elles présentant ses avantages et ses limites. Globalement, si la diffusion des OGM est un fait indéniable, les études ne permettent pas de tirer des conclusions aussi nettes quant à l’impact des OGM pour tous les agriculteurs qui adoptent ces produits, et cela en particulier pour le soja. Cette absence de conclusion s’explique en grande partie par la difficulté à mesurer certains facteurs qui influencent le comportement d’adoption, comme par exemple la simplification de la conduite des cultures. En contrepartie, il faut noter que les études présentées ici ne prennent pas en compte un certain nombre d’effets indirects qui peuvent faire diminuer les gains des agriculteurs.

Neutron Investigation of Magnon Spectrum in Haematite; Etude du Spectre de Magnons dans l'Hematite, au Moyen des Neutrons; Nejtronnoe issledovanie spektra magnona v gematite; Estudio, por Metodos Neutronicos, del Espectro de Magnones en la Hematita

Energy Technology Data Exchange (ETDEWEB)

Dimitrijevic, Z.; Rzany, H.; Todorovic, J.; Wanic, A. [Institute for Nuclear Physics Cracow (Poland)

1965-04-15

faites a Vinea, dans le reacteur RA, au moyen d'un spectrometre a cristal. Des neutrons monochromatiques ({lambda} = 1.314A) ont ete diffuses par un grand monocristal d'hematite ({alpha}-Fe{sub 2}O{sub 3}). Les auteurs ont etudie la distribution angulaire des neutrons diffuses inelastiquement. Pour plusieurs angles de decalage {Delta}{theta} du cristal, ils ont mesure la largeur {Gamma} du faisceau de neutrons diffuses (appele cone de diffusion). Ils ont attribue ce cone a la surface de diffusion des magnons qui entoure le point du reseau reciproque (1,1,1). Pour plusieurs vitesses de magnons, ils ont mesure {Gamma}, en fonction de l'angle de decalage, et l*ont comparee aux points experimentaux. Ils ont constate que dans la direction [111], la vitesse avait une valeur egale a 25,5{+-}1,0 km/s. Ils ont decouvert une anisotropie structurelle de la relation de dispersion des magnons. Ils ont note que la vitesse etait plus elevee pour les directions de propagation paralleles a l'axe [111], ce qui concorde qualitativement avec les mesures anterieures (faites par Riste et d'autres) qui ont donne v = 38 km/s. Il n'a pas ete possible de mettre en evidence l'existence d'une discontinuite Eg dans la bande acoustique de l'energie des magnons. Selon les evaluations, Eg doit avoir une valeur inferieure a 1 meV. En utilisant le formalisme de Wallace, les auteurs ont calcule les relations de dispersion des magnons dans l'hematite. A cet effet, ils ont admis l'existence de deux integrales d'echange J{sub 1} et J{sub 2} qui ne disparaissent pas. J{sub 1} et J{sub 2} expriment le couplage a superechange d'energie entre les spins d'ions Fe voisins lies par l'intermediaire d'un ion oxygene. Les angles Fe-O-Fe des liaisons considerees sont respectivement de 132 et 126 Degree-Sign . Les resultats theoriques et les donnees experimentales actuelles concordent de facon satisfaisante lorsque SJ{sub 1} = SJ{sub 2} = 5,1 meV. Outre la branche acoustique des magnons, les calculs ont revele l

DNS: Diffuse scattering neutron time-of-flight spectrometer

Directory of Open Access Journals (Sweden)

Yixi Su

2015-08-01

Full Text Available DNS is a versatile diffuse scattering instrument with polarisation analysis operated by the Jülich Centre for Neutron Science (JCNS, Forschungszentrum Jülich GmbH, outstation at the Heinz Maier-Leibnitz Zentrum (MLZ. Compact design, a large double-focusing PG monochromator and a highly efficient supermirror-based polarizer provide a polarized neutron flux of about 107 n cm-2 s-1. DNS is used for the studies of highly frustrated spin systems, strongly correlated electrons, emergent functional materials and soft condensed matter.

Measurement of the diffusion length of thermal neutrons in the beryllium oxide

International Nuclear Information System (INIS)

Koechlin, J.C.; Martelly, J.; Duggal, V.P.

1955-01-01

The diffusion length of thermal neutrons in the beryllium oxide has been obtained while studying the spatial distribution of the neutrons in a massive parallelepiped of this matter placed before the thermal column of the reactor core of Saclay. The mean density of the beryllium oxide (BeO) is 2,95 gr/cm 3 , the mean density of the massif is 2,92 gr/cm 3 . The value of the diffusion length, deducted of the done measures, is: L = 32,7 ± 0,5 cm (likely gap). Some remarks are formulated about the influence of the spectral distribution of the neutrons flux used. (authors) [fr

RBE of Monoenergetic Fast Neutrons: Cytogenetic Effects in Maize; EBR des Neutrons Rapides Monoenergeniques: Effets Cytogenetiques sur le Mais; Obeh monoehnergeticheskikh bystrykh nejtronov: tsitogeneticheskie izmeneniya u kukuruzy; EBR de los Neutrones Rapidos Monoenergeticos: Efectos Citogeneticos en el Maiz

Energy Technology Data Exchange (ETDEWEB)

Smith, H. H.; Bateman, J. L.; Quastler, H.; Rossi, H. H. [Biology and Medical Departments, Brookhaven National Laboratory, Upton, NY (United States)

1964-05-15

The maize used in these experiments has the advantage for RBE studies of yielding a basically first-order dose-response curve with low as well as with high LET radiations. An exposure apparatus was used which produced essentially equal dose rates in five rings of seeds placed so as to intercept neutrons of 0.43, 0.65, 1.00, 1.50 and 1.80 MeV. The mutant sectors produced in leaves are believed to be due mostly to simple chromosome breakage and deletion. Experiments were performed at dosages that gave responses which were linear, below saturation levels, and overlapping in range for the monoenergetic fast neutrons and 250 kVp X-rays. RBE values, calculated from relative slopes of linear-regression lines for neutrons and X-rays ranged from 42 to 135 with an overall average of about 70. Fast neutrons of 0.43 MeV energy were the most efficient, of those used, in producing g{sub 2} mutant sectors. (author) [French] Pour les etudes sur l'EBR, le maft utilise au cours des experiences offre l'avantage de donner une courbe dose-reponse qui est essentiellement de premier ordre, que le TLE du rayonnement soit faible ou eleve. Les auteurs ont utilise un appareil d'irradiation assurant des debits de dose pratiquement egaux dans cinq couronnes de semences disposees de maniere a intercepter les neutrons de 0,43, 0,65, 1,00, 1,50 et 1,80 MeV. On pense que les secteurs mutants produits dans les feuilles sont dus essentiellement a une rupture et une disparition de chromosomes simples. Les auteurs ont fait des experiences a des doses qui ont donne des reponses lineaires, inferieures au niveau de saturation et se chevauchant dans le cas des neutrons rapides et des rayons X de 250 kV-crete. Les valeurs de l'EBR calculees d'apres la comparaison des pentes des courbes de regression lineaire pour les neutrons et les rayons X varient de 42 a 135, avec une valeur moyenne d'environ 70. Parmi les neutrons utilises, les neutrons rapides de 0,43 MeV ont ete les plus efficaces pour produire des

Measurement of the diffusion length of thermal neutrons inside graphite

International Nuclear Information System (INIS)

Ertaud, A.; Beauge, R.; Fauquez, H.; De Laboulay, H.; Mercier, C.; Vautrey, L.

1948-11-01

The diffusion length of thermal neutrons inside a given industrial graphite is determined by measuring the neutron density inside a parallelepipedal piling up of graphite bricks (2.10 x 2.10 x 2.442 m). A 3.8 curies (Ra α → Be) source is placed inside the parallelepipedal block of graphite and thin manganese detectors are used. Corrections are added to the unweighted measurements to take into account the effects of the damping of supra-thermal neutrons in the measurement area. These corrections are experimentally deduced from the differential measurements made with a cadmium screen interposed between the source and the first plane of measurement. An error analysis completes the report. The diffusion length obtained is: L = 45.7 cm ± 0.3. The average density of the graphite used is 1.76 and the average apparent density of the piling up is 1.71. (J.S.)

Studies on the rice seed (Oryza sativa) sensitivity to gama radiation, neutrons di-enthyl sulphate (DES) and sodium azide (SA)

International Nuclear Information System (INIS)

Guimaraes, E.P.

1978-04-01

Dry rice seeds of 'Dourado Precoce' variety were treated with gama-rays, thermal and fast neutrons, di-ethyl sulphate (DES) and sodium azide (SA). Biologial effects on survival, seedling height, plant fertility and chlorophy 11 mutation frequency were studied in the M 1 generation. The data obtained were analysed by 'independent test'. Survival was not affected by increased gama ray dosage, but by increased DES and SA dosages, as well as by SA treatments with increased solution acidity for seeds pretreated in distilled water. Seedling height was decreased with the increase of gama-ray dosage, DES concentration. SA acidity and concentration. Plant fertility was reduced drastically with higher gama-ray and neutron dosages and with high SA acidity and concentrations. The maximum chlorophy 11 mutation frequency was obtained in the SA treatment. Seeds pretreated with distilled water showed larger physiological damages in the SA treatments and higher mutation frequencies. Regarding chlorophyll mutation frequency, thermal neutrons were more efficient than gama-rays. SA was the most efficient mutagen used . DES showed lower physiological damages in the M 1 plants, but did not induce mutations in the used conditions [pt

Thermal diffuse scattering in angular-dispersive neutron diffraction

International Nuclear Information System (INIS)

Popa, N.C.; Willis, B.T.M.

1998-01-01

The theoretical treatment of one-phonon thermal diffuse scattering (TDS) in single-crystal neutron diffraction at fixed incident wavelength is reanalysed in the light of the analysis given by Popa and Willis [Acta Cryst. (1994), (1997)] for the time-of-flight method. Isotropic propagation of sound with different velocities for the longitudinal and transverse modes is assumed. As in time-of-flight diffraction, there exists, for certain scanning variables, a forbidden range in the one-phonon TDS of slower-than-sound neutrons, and this permits the determination of the sound velocity in the crystal. A fast algorithm is given for the TDS correction of neutron diffraction data collected at a fixed wavelength: this algorithm is similar to that reported earlier for the time-of-flight case. (orig.)

Neutron-Phonon Interaction Studies in Copper, Zinc and Magnesium Single Crystals; Etude des interactions neutron-phonon dans des monocristaux de cuivre, de zinc et de magnesium; Izuchenie vzaimodejstviya nejtronov i fononov v monokristallakh medi, ninki i magniya; Estudio de las interacciones neutron-fonon en monocristales de cobre, cinc y magnesio

Energy Technology Data Exchange (ETDEWEB)

Maliszewski, E.; Sosnowski, J.; Blinowski, K.; Kozubowski, J.; Padlo, L.; Sledziewska, D. [Institute of Nuclear Research, Warsaw (Poland)

1963-01-15

trouve nettement au-dessus de la courbe de Jacobsen et les points experimentaux signales dans le present memoire confirment les resultats de Cribier er de Jacrot. Les relations de dispersion des phonons dans des monocristaux de zinc et de magnesium ont ete etudiees a l'aide de la methode des neutrons froids et d'un spectrometre neutronique a cristal triaxe, Les auteurs ont trace les surfaces de diffusion dans le plan [1010], trouve les branches AT et AL ainsi que les relations de dispersion des phonons dans les directions [001] et [010] . Us ont compare ces resultats a ceux qu'a obtenus Johnson avec des rayons X . Dans la direction [001] , les resultats s'adaptent bien a la courbe de Johnson en ce qui concerne la branche AL. Dans la direction [010] , pour la branche AT, les auteurs observent un grand ecart entre leurs resultats et ceux de Johnson. 11s en donnent une explication. Ils decrivent des mesures semblables faites a l'heure actuelle dans les memes directions, dans des monocristaux de magnesium. (author) [Spanish] Los autores han estudiado las relaciones de dispersion de fonones en monocristales de cobre con ayuda de un espectrometro neutronico triaxial. En la direccion 100 han observado la rama transversal, no mencionada en las memorias de Cribier y Jacrot. Esta rama concuerda satisfactoriamente con los datos recientes relativos a la velocidad del sonido, pero difiere en parte de los resultados roentgenograncos obtenidos por Jacobsen. En cuanto a la rama longitudinal en la direccion 100, la curva de dispersion obtenida por Cribier y Jacrot se encuentra netamente por encima de la curva de Jacobsen y los puntos experimentales a que se refiere la presente memoria connrman los resultados de Cribier y Jacrot. Los autores han estudiado las relaciones de dispersion de fonones en monocristales de cinc y magnesio por el procedimiento de los neutrones frios, utilizando ademas un espectrometro neutronico triaxial. Han determinado las superficies de dispersion en el plano

Diffuse neutron scattering study of Cu2−xSe

DEFF Research Database (Denmark)

Cava, R. J.; Andersen, Niels Hessel; Clausen, Kurt Nørgaard

1986-01-01

We have measured the diffuse neutron scattering in the hkk plane for Cu2Se and Cu1.8Se at 180°C and 51°C, respectively, in the cubic antifluorite type phase. The diffuse scattering shows significant structure, indicative of correlated short range mobile ion ordering. The short range order is foun...

Some Measurements of Thermal Neutron Spectra; Mesure de Spectres de Neutrons Thermiques; Nekotorye izmereniya spektrov teplovykh nejtronov; Algunas Mediciones de Espectros de Neutrones Termicos

Energy Technology Data Exchange (ETDEWEB)

Poole, M. J.; Schofield, P.; Sinclair, R. N. [United Kingdom Atomic Energy Authority, Research Group, Atomic Energy Research Establishment, Harwell (United Kingdom)

1964-04-15

River scattering-law data, in an attempt to predict the shape of the energy spectrum at positions in both blocks. A second experiment introduces a spatial discontinuity instead of thermal. A disc of cadmium sufficiently thick to be black to all neutrons below 0.4 eV is fixed centrally in a tank of water and spectra have been measured near to the disc as a function both of position and of direction of travel of the neutrons (relative to the plane of the cadmium). Finally may be mentioned spectrum measurements in a graphite-uranium lartice which provide a check on spectrum calculations in reactor geometry. Many spectrum experiments can be carried out in either critical or subcritical geometry, and in some discussion is given of the relative merits of those two methods of approach. (author) [French] Les auteurs ont entrepris a Harwell l'execution de travaux complementaires pour determiner la loi de diffusion dans le ralentisseur et pour verifier la validite de cette loi en procedant a des mesures du spectre. La plupart des experiences sur la loi de diffusion ont ete faites a Chalk River (Canada); le traitement des donnees se fait au Laboratoire national d'Argonne (Etats-Unis), et l'analyse ainsi que l'extrapolation necessaire des resultats des mesures se font a Harwell. Les mesures du spectre se divisent naturellement en deux groupes. En utilisant la spectroscopie du temps de vol, les auteurs ont fait un grand nombre de mesures de spectres de neutrons thermiques dans des ralentisseurs empoisonnes homogenes. Ces travaux completent ceux qui ont deja ete effectues par les auteurs et par Beyster et al.; ils servent a verifier les valeurs des sections efficaces du transfert d'energie o(E -> E) deduites de la loi de diffusion en usage. Toutefois, cette experience n'est nullement sensible aux variations angulaires de la diffusion ni a la partie de la section efficace de diffusion qui ne comporte pas de changement de l'energie du neutron diffuse; or, ces deux facteurs ont leur

L'analyse par activation de neutrons de réacteur

Science.gov (United States)

Meyer, G.

2003-02-01

Quand les neutrons traversent la matière, certains sont transmis sans interaction, les autres interagissent avec le milieu traversé par diffusion et par absorption. Ce phénomène d'absorption est utilisé pour se protéger des neutrons, mais aussi pour les détecter; il peut également être utilisé pour identifier les noyaux “absorbants" et ainsi analyser le milieu traversé. En effet par différentes réactions nucléaires (n,γ), (n,p), (n,α), (n,fission), on obtient des noyaux résiduels qui sont souvent radioactifs; on dit que l'échantillon est “activé". Si l'on connaît le rendement d'activation et donc le pourcentage de noyaux ainsi “transmutés", les mesures de radioactivité induite vont permettre de déterminer la composition de l'échantillon irradié. Cette méthode dite d'analyse par activation neutronique est pratiquée depuis la découverte du neutron. Elle a permis grâce à sa sélectivité et à sa sensibilité d'avoir accès au domaine des traces et des ultra-traces dans des champs d'application très divers comme la métallurgie, l'archéologie, la biologie, la géochimie etc...

Automation on computers of the partial area method for the analysis of 's' neutron induced resonances. I. Without interference terms. (1962); Automatisation sur ordinateur de la methode des aires partielle dans l'analyse des resonances induites par les neutrons 's'. I. sans terme d'interference. (1962)

Energy Technology Data Exchange (ETDEWEB)

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

1962-07-01

This report deals with numerical analysis of transmission resonances induced by 's' wave neutrons in time of flight experiments. The analysis method used is the partial area one. The only case treated here, in this first part, is the one when the interference term can be neglected. Programs and subroutines are thoroughly described, which determine from experimental raw data the resonant transmission, the partial areas, and the resonance parameters. (authors) [French] Le present rapport a pour objet l'analyse numerique sur ordinateur IBM 7090, des resonances dues aux neutrons 's' dans les experiences de transmission par temps de vol, la methode d'analyse utilisee etant la methode des aires partielles. Dans cette premiere partie, seul a ete envisage le cas ou le terme d'interference peut etre neglige. On y trouvera une description detaillee des programmes et sous programmes elabores pour determiner les transmissions resonantes avec leur trace, a partir des donnees experimentales brutes, les aires partielles afferentes ainsi que les parametres caracteristiques des resonances. (auteurs)

Diffuse neutron scattering from anion-excess strontium chloride

DEFF Research Database (Denmark)

Goff, J.P.; Clausen, K.N.; Fåk, B.

1992-01-01

The defect structure and diffusional processes have been studied in the anion-excess fluorite (Sr, Y)Cl2.03 by diffuse neutron scattering techniques. Static cuboctahedral clusters found at ambient temperature break up at temperatures below 1050 K, where the anion disorder is highly dynamic. The a...

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

Energy Technology Data Exchange (ETDEWEB)

Naudet, R.

1963-05-15

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

Measurement of diffusion length of thermal neutrons in concrete

International Nuclear Information System (INIS)

Moser, M.

2007-04-01

The diffusion length of neutrons with a medium energy < 0.025 eV in concrete were determined using 4π-β detector and gamma detectors. Then it was possible to determine how deep can neutrons penetrate diverse concrete construction parts in a reactor in operation, with this method the dismantling process of a reactor can be planned in terms of what parts can be removed without danger and what parts can be assumed still are activated. (nevyjel)

The use of diffusion theory to compute invasion effects for the pulsed neutron thermal decay time log

International Nuclear Information System (INIS)

Tittle, C.W.

1992-01-01

Diffusion theory has been successfully used to model the effect of fluid invasion into the formation for neutron porosity logs and for the gamma-gamma density log. The purpose of this paper is to present results of computations using a five-group time-dependent diffusion code on invasion effects for the pulsed neutron thermal decay time log. Previous invasion studies by the author involved the use of a three-dimensional three-group steady-state diffusion theory to model the dual-detector thermal neutron porosity log and the gamma-gamma density log. The five-group time-dependent code MGNDE (Multi-Group Neutron Diffusion Equation) used in this work was written by Ferguson. It has been successfully used to compute the intrinsic formation life-time correction for pulsed neutron thermal decay time logs. This application involves the effect of fluid invasion into the formation

The n,{gamma} discrimination in recoil-proton proportional counters. Application to the measurement of fast neutron spectra; Discrimination n,{gamma} dans les compteurs proportionnels a protons de recul. Application a la mesure des spectres de neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

A description is given of a spectrometry chain working in the energy range of a few keV to 1 MeV, and designed for measurement of fast neutron spectra. It consists of detectors, recoil proton proportional counters built especially for this work by R. COMTE (DEG/SER) and which make it possible to cover the energy range and also associated electronic equipment. A brief description is first given of the physical processes involved: (n,p) collisions in the gas, influence of {gamma} radiation; the method of discrimination is then presented. It is based on the difference in the rise-times of the pulses. In the experiments described here the use of a bi-parametric system made it possible to employ the most simple discrimination device, based on the fact that the high frequency gamma pulse components are, at a given energy, weaker than those of the neutron pulses. Results are given of measurements carried out on the Van der Graaff (mono-energetic neutrons for testing the linearity of the chain and the resolving power of the counters), and of those made in a sub-critical system Hug at Cadarache. (author) [French] On decrit une chaine de spectrometrie travaillant dans le domaine d'energie de quelques keV a 1 MeV destinee a la mesure des spectres de neutrons rapides. Elle comprend les detecteurs, compteurs proportionnels a protons de recul, realises specialement pour cette etude par M. R. COMTE (DEG/SER), permettant de couvrir la gamme d'energie et l'electronique associee. Apres un rappel des processus physiques mis en jeu: chocs (n,p) dans les gaz, influence des rayonnements {gamma}, on expose la methode de discrimination utilisee. Celle-ci est basee sur la difference des temps de montee des impulsions. Au cours des experiences rapportees ici, la mise en oeuvre d'un ensemble bi-parametrique a permis d'utiliser le dispositif de discrimination le plus simple, base sur la remarque que les composantes a haute frequence des impulsions {gamma} sont, a energie egale, plus faibles

The n,{gamma} discrimination in recoil-proton proportional counters. Application to the measurement of fast neutron spectra; Discrimination n,{gamma} dans les compteurs proportionnels a protons de recul. Application a la mesure des spectres de neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

A description is given of a spectrometry chain working in the energy range of a few keV to 1 MeV, and designed for measurement of fast neutron spectra. It consists of detectors, recoil proton proportional counters built especially for this work by R. COMTE (DEG/SER) and which make it possible to cover the energy range and also associated electronic equipment. A brief description is first given of the physical processes involved: (n,p) collisions in the gas, influence of {gamma} radiation; the method of discrimination is then presented. It is based on the difference in the rise-times of the pulses. In the experiments described here the use of a bi-parametric system made it possible to employ the most simple discrimination device, based on the fact that the high frequency gamma pulse components are, at a given energy, weaker than those of the neutron pulses. Results are given of measurements carried out on the Van der Graaff (mono-energetic neutrons for testing the linearity of the chain and the resolving power of the counters), and of those made in a sub-critical system Hug at Cadarache. (author) [French] On decrit une chaine de spectrometrie travaillant dans le domaine d'energie de quelques keV a 1 MeV destinee a la mesure des spectres de neutrons rapides. Elle comprend les detecteurs, compteurs proportionnels a protons de recul, realises specialement pour cette etude par M. R. COMTE (DEG/SER), permettant de couvrir la gamme d'energie et l'electronique associee. Apres un rappel des processus physiques mis en jeu: chocs (n,p) dans les gaz, influence des rayonnements {gamma}, on expose la methode de discrimination utilisee. Celle-ci est basee sur la difference des temps de montee des impulsions. Au cours des experiences rapportees ici, la mise en oeuvre d'un ensemble bi-parametrique a permis d'utiliser le dispositif de discrimination le plus simple, base sur la remarque que les composantes a haute frequence des impulsions {gamma} sont, a

Thermal neutron diffusion parameters dependent on the flux energy distribution in finite hydrogenous media

International Nuclear Information System (INIS)

Drozdowicz, K.

1999-01-01

Macroscopic parameters for a description of the thermal neutron transport in finite volumes are considered. A very good correspondence between the theoretical and experimental parameters of hydrogenous media is attained. Thermal neutrons in the medium possess an energy distribution, which is dependent on the size (characterized by the geometric buckling) and on the neutron transport properties of the medium. In a hydrogenous material the thermal neutron transport is dominated by the scattering cross section which is strongly dependent on energy. A monoenergetic treatment of the thermal neutron group (admissible for other materials) leads in this case to a discrepancy between theoretical and experimental results. In the present paper the theoretical definitions of the pulsed thermal neutron parameters (the absorption rate, the diffusion coefficient, and the diffusion cooling coefficient) are based on Nelkin's analysis of the decay of a neutron pulse. Problems of the experimental determination of these parameters for a hydrogenous medium are discussed. A theoretical calculation of the pulsed parameters requires knowledge of the scattering kernel. For thermal neutrons it is individual for each hydrogenous material because neutron scattering on hydrogen nuclei bound in a molecule is affected by the molecular dynamics (characterized with internal energy modes which are comparable to the incident neutron energy). Granada's synthetic model for slow-neutron scattering is used. The complete up-dated formalism of calculation of the energy transfer scattering kernel after this model is presented in the paper. An influence of some minor variants within the model on the calculated differential and integral neutron parameters is shown. The theoretical energy-dependent scattering cross section (of Plexiglas) is compared to experimental results. A particular attention is paid to the calculation of the diffusion cooling coefficient. A solution of an equation, which determines the

Study of niobium isotopes having excess neutrons and a short half-life; Etude des isotopes du niobium excedentaires en neutrons et de courte periode

Energy Technology Data Exchange (ETDEWEB)

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

1968-02-01

By irradiating Mo with 14 MeV neutrons isomers have been found for {sup 98}Nb (2.8 s half-life) {sup 99}Nb (9 s) and {sup 100}Nb (2.4 s). No isomer of this type seems to exist for {sup 96}Nb. Rapid separation methods are developed for isolating {sup 98}Zr from fission products, and for separating Zr and Nb. The half-life of {sup 98}Zr is measured (31 s) and the formation of {sup 98}Nb (2.8 s) from {sup 98}Zr (31 s) is shown by milking. Rough {beta} and {gamma} measurements of these nuclei are described. The {gamma} spectrum of {sup 98}Nb (51 mn) is studied with a high-resolution Ge/Li - detector. (authors) [French] Des irradiations des isotopes de molybdene avec des neutrons de 14 MeV ont mis en evidence l'existence des isomeres de {sup 98}Nb (periode 2.8 s) {sup 99}Nb (9 s) et {sup 100}Nb (2.4 s). Pour le {sup 96}Nb un isomere de ce type ne semble pas exister. Des methodes rapides de separation sont mises au point pour isoler le zirconium 98 des produits de fission, et pour separer ensuite le niobium du zirconium. La periode du {sup 98}Zr est de 3l s, et on demontre la formation du {sup 98}Nb (2.8 s) a partir du Zr (31 s). Ces corps sont etudies sommairement en spectroscopie {beta} et {gamma}. Le spectre gamma de {sup 98}Nb (periode 51 mn) est etudie avec un detecteur Ge/Li de haute resolution. (auteurs)

Parallel diffusion length on thermal neutrons in rod type lattices

International Nuclear Information System (INIS)

Ahmed, T.; Siddiqui, S.A.M.M.; Khan, A.M.

1981-11-01

Calculation of diffusion lengths of thermal neutrons in lead-water and aluminum water lattices in direction parallel to the rods are performed using one group diffusion equation together with Shevelev transport correction. The formalism is then applied to two practical cases, the Kawasaki (Hitachi) and the Douglas point (Candu) reactor lattices. Our results are in good agreement with the observed values. (author)

GPU-accelerated 3D neutron diffusion code based on finite difference method

Energy Technology Data Exchange (ETDEWEB)

Xu, Q.; Yu, G.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ. (China)

2012-07-01

Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

GPU-accelerated 3D neutron diffusion code based on finite difference method

International Nuclear Information System (INIS)

Xu, Q.; Yu, G.; Wang, K.

2012-01-01

Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

Interpretation of the quasi-elastic neutron scattering on PAA by rotational diffusion models

International Nuclear Information System (INIS)

Bata, L.; Vizi, J.; Kugler, S.

1974-10-01

First the most important data determined by other methods for para azoxy anisolon (PAA) are collected. This molecule makes a rotational oscillational motion around the mean molecular direction. The details of this motion can be determined by inelastic neutron scattering. Quasielastic neutron scattering measurements were carried out without orienting magnetic field on a time-of-flight facility with neutron beam of 4.26 meV. For the interpretation of the results two models, the spherical rotation diffusion model and the circular random walk model are investigated. The comparison shows that the circular random walk model (with N=8 sites, d=4A diameter and K=10 10 s -1 rate constant) fits very well with the quasi-elastic neutron scattering, while the spherical rotational diffusion model seems to be incorrect. (Sz.N.Z.)

Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

Energy Technology Data Exchange (ETDEWEB)

Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1970-07-01

Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

Survey of Pulsed Neutron Source Methods for Multiplying Media; Methodes des Neutrons Pulses Pour l'Etude des Milieux Multiplicateurs; Obzor metodov s ispol'zovaniem istochnikov impul'snykh nejtronov dlya razmnozhayushchej sredy; Estudio Panoramico de los Metodos de Empleo de Fuentes de Neutrons Pulsados en Medios Multiplicadores

Energy Technology Data Exchange (ETDEWEB)

Garelis, E. [General Electric Company, Vallecitos Atomic Laboratory, Pleasanton, CA (United States)

1965-10-15

. Thus, this makes it possible to apply the methods developed for the repetitive to that of the pseudo-random technique. (author) [French] Depuis quelques annees, il existe deux opinions sur la maniere la plus efficace de mesurer la reactivite d'arret au moyen de generateurs de neutrons puises, savoir: a) les mesures classiques au moyen d'une source puisee a repetition; b) des methodes qui sont fondees sur la reponse a des impulsions pseudo-aleatoires et utilisent une correlation entre l'entree et la sortie. Dans les deux cas, les renseignements pertinents que l'on obtient sont identiques, en d'autres termes, idealement les deux methodes servent a determiner la fonction de reponse. L'auteur retrace l'evolution des methodes de mesure de la reactivite fondees sur les sources puisees dans les systemes a neutrons thermiques, depuis les premieres tentatives de Sjoestrand jusqu'a la recente methode de (k{beta}/ Script-Small-L ). Dans la methode habituelle, la fonction de Green relative a l'assemblage sous-critique - la reponse du reacteur a une source de neutrons a fonction delta - constitue la propriete recherchee. La decroissance exponentielle, exp(-{alpha}t), de la fonction de Green donne une constante de decroissance des neutrons instantanes independante du lieu. L'auteur examine les methodes permettant de deduire la reactivite de la mesure de a , par exemple, la mesure de a a l'etat critique retarde et la recente methode de (k{beta}/ Script-Small-L ). Il etudie notamment les traitements du mode fondamental a la lumiere de la theorie des methodes des neutrons puises, telles qu'elle a ete mise au point pour le modele (k{beta}/ Script-Small-L )- L'auteur examine les incidences de la theorie des neutrons puises en vue d'obtenir des constantes de decroissance precises et les donnees voulues pour l'analyse des systemes puises. Il passe en revue les travaux experimentaux qui font ressortir a la fois les avantages et les limitations de la technique de (k{beta}/ Script

X-ray and neutron diffuse scattering in LiNbO3 from 38 to 1200 K

International Nuclear Information System (INIS)

Zotov, N.; Mayer, H.M.; Guethoff, F.; Hohlwein, D.

1995-01-01

A semi-quantitative description of X-ray and neutron diffuse scattering from congruent lithium niobate, LiNbO 3 , is given. The diffuse scattering is concentrated in three sets of diffuse planes perpendicular to the pseudo-cubic symmetry-related [221], [241] and [ anti 4 anti 21] directions and can be attributed to one-dimensional displacive and chemical disorder along these directions. The variation of the X-ray and neutron diffuse intensities with the scattering vector, as well as the comparison between X-ray and neutron data, indicate that more than one type of atom is involved. Temperature variations are followed from 38 to 1200 K. Different disorder models are discussed. The increase of the integrated intensities of the diffuse lines along the [0 1k 2l] * and [0 anti 1k 4l] * directions (i.e. sections of the diffuse planes) up to 800 K followed by a slight decrease at higher temperatures may be interpreted either by static disorder related to temperature-dependent variation of disorder/defect clusters or by dynamic disorder. Inelastic neutron scattering experiments do not show any anomaly of the transversal acoustic (TA) modes. (orig.)

Effects of Monoenergetic Neutron Radiation on Human Cells in Tissue Culture; Effets des Neutrons Monoenergetiques sur des Cellules de Tissus Humains en Culture; Dejstvie oblucheniya monoehnergeticheskimi nejtronami na kletki kul'tury tkanej cheloveka; Efectos de la Radiacion Neutronica Monoenergetica sobre las Celulas Humanas en Cultivos de Tejidos

Energy Technology Data Exchange (ETDEWEB)

Broerse, J. J.; Barendsen, G. W. [Radiobiological Institute of the Organization for Health Research TNO, Rijswijk (Z.H.) (Netherlands)

1964-03-15

Dose effect relations concerning inhibition of clone formation by kidney cells of human origin in culture after irradiation with monoenergetic neutrons of about 3-MeV and about 15-MeV energy, are reported. For 3-MeV neutrons the dose rate was equal to 0.2 rad/min, while for 15-MeV neutrons dose rates of 2.7 rad/min and 12 rad/min were obtained. The cells were irradiated in culture dishes with a thin Melinex bottom and in small flexible nylon and silicon rubber tubes. Neutron fluxes were determined by measuring the activation of sulphur pellets mounted respectively between the dishes and the tubes. Calculation of the dose on the basis of the neutron flux presented a number of problems which are discussed. Because no radiation equilibrium exists around the cells, the resulting inaccuracies in the dose determinations are investigated. Dose-survival curves for both types of neutrons are presented. For 3-MeV neutrons an RBE was found ranging from 6.5 at low doses to about 3.1 at high doses, while for 15-MeV neutrons the RBE varied from 1.9 to 1.6. By comparison of these RBE-values with the RBE-LET relation obtained in earlier experiments with deuterons and {alpha}-particles, it is possible to derive an effective LET for the neutrons used. In order to elucidate certain problems related to the death of experimental animals after X- or neutron- irradiation, a technique has been developed whereby the cells, contained in small tubes, are used to measure the effective dose (i.e . dose corrected for the RBE) at sites of interest in phantoms or animals. (author) [French] Les auteurs indiquent les relations dose-effet pour l 'inhibition de la formation de lignees par les cellules renales d' origine humaine apres irradiation par des neutrons monoenergetiques de 3 MeV et de 15 MeV. Pour les neutrons de 3 MeV, le debit de dose etait egal a 0,2 rad/min, alors que pour les neutrons de 15 MeV on a obtenu des debits de dose de 2,7 rad/min et de 12 rad/min. L'irradiation des cellules

Incoherent neutron scattering functions for random jump diffusion in bounded and infinite media

International Nuclear Information System (INIS)

Hall, P.L.; Ross, D.K.

1981-01-01

The incoherent neutron scattering function for unbounded jump diffusion is calculated from random walk theory assuming a gaussian distribution of jump lengths. The method is then applied to calculate the scattering function for spatially bounded random jumps in one dimension. The dependence on momentum transfer of the quasi-elastic energy broadenings predicted by this model and a previous model for bounded one-dimensional continuous diffusion are calculated and compared with the predictions of models for diffusion in unbounded media. The one-dimensional solutions can readily be generalized to three dimensions to provide a description of quasi-elastic scattering of neutrons by molecules undergoing localized random motions. (author)

Study of heterogeneous multiplying and non-multiplying media by the neutron pulsed source technique; Etude des milieux heterogenes multiplicateurs et non-multiplicateurs par la technique de la source pulsee de neutrons

Energy Technology Data Exchange (ETDEWEB)

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

1968-06-01

The pulsed neutron technique consists essentially in sending in the medium to be studied a short neutron pulse and in determining the asymptotic decay constant of the generated population. The variation of the decay constant as a function of the size of the medium allows the medium characteristics to be defined. This technique has been largely developed these last years and has been applied as well to moderator as to multiplying media, in most cases homogeneous ones. We considered of interest of apply this technique to lattices, to see if useful informations could be collected for lattice calculations. We present here a general theoretical study of the problem, and results and interpretation of a series of experiments made on graphite lattices. There is a good agreement for non-multiplying media. In the case of multiplying media, it is shown that the age value used until now in graphite lattices calculations is over-estimated by about 10 per cent. [French] La technique de la pulsation neutronique consiste essentiellement a envoyer dans le milieu a etudier une courte bouffee de neutrons et a determiner la constante de decroissance asymptotique de la population engendree. La variation de cette constante de decroissance en fonction des dimensions du milieu permet de determiner ses caracteristiques. Cette technique a connu ces dernieres annees un grand essor et a ete appliquee a des moderateurs et des milieux multiplicateurs. Il s'agissait dans la plupart des cas de milieux homogenes. Il nous a semble interessant de l'utiliser dans le cas des reseaux, afin de voir si ces experiences peuvent fournir des renseignements utiles aux calculs. Nous presentons ici une etude theorique generale du probleme, ainsi que les resultats et l'interpretation d'une serie d'experiences faites sur des reseaux a graphite. L'accord est bon dans le cas des reseaux non-multiplicateurs. Dans le cas des reseaux multiplicateurs, on montre que la valeur de l'age utilisee jusqu'ici dans les calculs

Problems Associated with the Monochromatisation of Slow Neutrons; Quelques aspects de la monochromatisation des neutrons thermiques; Nekotorye voprosy monokhromatizatsii medlennykh nejtronov; Algunos problemas de la monocromatizacion de neutrones lentos

Energy Technology Data Exchange (ETDEWEB)

Vertebnyj, V P; Kolotyj, V V; Majstrenko, A N

1963-01-15

The paper sets out the result of calculations to determine the shape of the spectral line of twin-rotor, pulsing, slow-neutron monochromators, as a function of the shifts in the phases between the rotors. Consideration is also given to die possibility of using certain light elements as slow-neutron filters. (author) [French] Les auteurs exposent les resultats de leurs calculs relatifs a la forme du spectre des monochromateurs a impulsions et a deux rotors pour neutrons thermiques en fonction de la distance et du dephasage entre les rotors. Ils examinent egalement la possibilite d*utiliser certains elements legers comme filtres a neutrons lents. (author) [Spanish] Los autores exponen ios resultadas de sus calculos relativos a la forma de la linea espectral de los monocromadores pulsantes de dos rotores para neutrones lentos en funcion de la distancia y del desias amiento de los rotores. Examinan asimismo la posibilidad de utilizar algunos elementos ligeros como filtros de neutrones lentos. (author)

A New Apparatus for Inelastic, Quasi-Elastic and Elastic Cold Neutron Measurements; Un nouveau appareil pour les mesures de diffusion inelastique , quasi-elastique et elastique des neutrons lents; Novyj pribor dlya izmereniya neuprugogo, kvaziuprugogo i uprugogo rasseyaniya kholodnykh nejtrohov; Nuevo aparato para mediciones inelastic as, cuasi elasticas y elasticas de neutrones frios

Energy Technology Data Exchange (ETDEWEB)

Otnes, K; Palevsky, H [Brookhaven National Laboratory, Upton, NY (United States)

1963-01-15

despersion de la longueur d'onde (largeur totale a mi-hauteur) seront respectivement de 16 {mu}s et 0,16 A pour des neutrons incidents ayant une longueur d'onde de 4 A; l'intensite de la bouffee sur l'echantillon (4 x 1,6 cm) sera de 2 x 10{sup 6} n/s pouf mesures de diffusion quasi-elastique et elastique, la configuration a trois rotors conviendra parfaitement. La duree de la bouffee et la dispersion de longueur d'onde correspondante peuvent atteindre des valeurs aussi faibles que 8 {mu}s et 0, 04 A, ce qui donne une intensite de 10{sup 4} n/s sur un echantillon de 4 x 0, 8 cm. La longueur d'onde et la resolution en temps peuvent etre ajustees entre les; deux limites susmentionnees, de maniere a obtenir l'intensite de flux maximum pour une experience determinee. (author) [Spanish] Se esta construyendo un nuevo selector mecanico destinado al reactor de flujo intenso de Brookhaven. El aparato es del tipo de tres elementos rotores en fase. Los rotores, de 80 cm de diametro, giran a una velo- cidad maxima de 15000 rev/min y se han disenado de modo que emitan tres rafagas de neutrones monocromaticos por revolucion. Dos de los rotores giran alrededor de un eje horizontal, mientras que el tercero lo hace verticalmente. El sistma puede funcionar con uno, dos or tres elementos selectores, segun el tipo de medicion que se quiera efectuar. Para las mediciones inelasticas en que los neutrones ganan energia, lo mas indicado es utilizar un sistema de dos rotores. En este sistema, la duracion de las rafagas sera de 16 {mu}s y el ensanchamiento de longitudes de onda (amplitud plena a la mitad del valor maximo) de 0,16 A para neutrones incidentes de 4 A; la intensidd en la muestra (4 x 1,6 cm) sera de 2 x 10{sup 6} n/s. para las mediciones cuasi elasticas y elasticas resultara mas apropiado el sistema de tres rotores. La duracion de las rafagas y el ensanchamiento de longitudes de onda pueden llegar a un minimo de 8 {mu}s y 0,04 A, respectivamente, lo que representa una intensidad de 10{sup

The use of multi-energy-group neutron diffusion theory to numerically evaluate the relative utility of three dial-detector neutron porosity well logging tools

International Nuclear Information System (INIS)

Zalan, T.A.

1988-01-01

Multi-energy-group neutron diffusion theory is used to numerically evaluate the utility of two different dual-detector neutron porosity logging devices, a 14 MeV (accelerator) neutron source - epithermal neutron detector device and a 4 MeV neutron source - capture gamma-ray detector device, relative to the traditional 4 MeV neutron source - thermal neutron detector device. Fast and epithermal neutron diffusion parameters are calculated using Monte Carlo - derived neutron flux distributions. Thermal parameters are calculated from tabulated cross sections. An existing analytical method to describe the transport of gamma-rays through common earth materials is modified in order to accommodate the modeling of the 4 MeV neutron - capture gamma-ray device. The 14 MeV neutron - epithermal neutron device is found to be less sensitive to porosity than the 4 MeV neutron - capture gamma-ray device, which in turn is found to be less sensitive to porosity than the traditional 4 MeV neutron - thermal neutron device. Salinity effects are found to be comparable for the 4 MeV neutron - capture gamma-ray and 4 MeV neutron - thermal neutron devices. The 4 MeV neutron capture gamma-ray measurement is found to be deepest investigating

Ohmic ion temperature and thermal diffusivity profiles from the JET neutron emission profile monitor

Energy Technology Data Exchange (ETDEWEB)

Esposito, B. (ENEA, Frascati (Italy). Centro Ricerche Energia); Marcus, F.B.; Conroy, S.; Jarvis, O.N.; Loughlin, M.J.; Sadler, G.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell (United Kingdom))

1993-10-01

The JET neutron emission profile monitor was used to study ohmically heated deuterium discharges. The radial profile of the neutron emissivity is deduced from the line-integral data. The profiles of ion temperature, T[sub i], and ion thermal diffusivity, [chi][sub i], are derived under steady-state conditions. The ion thermal diffusivity is higher than, and its scaling with plasma current opposite to, that predicted by neoclassical theory. (author).

Ohmic ion temperature and thermal diffusivity profiles from the JET neutron emission profile monitor

International Nuclear Information System (INIS)

Esposito, B.

1993-01-01

The JET neutron emission profile monitor was used to study ohmically heated deuterium discharges. The radial profile of the neutron emissivity is deduced from the line-integral data. The profiles of ion temperature, T i , and ion thermal diffusivity, χ i , are derived under steady-state conditions. The ion thermal diffusivity is higher than, and its scaling with plasma current opposite to, that predicted by neoclassical theory. (author)

Estimating anisotropic diffusion of neutrons near the boundary of a pebble bed random system

Energy Technology Data Exchange (ETDEWEB)

Vasques, R. [Department of Mathematics, Center for Computational Engineering Science, RWTH Aachen University, Schinkel Strasse 2, D-52062 Aachen (Germany)

2013-07-01

Due to the arrangement of the pebbles in a Pebble Bed Reactor (PBR) core, if a neutron is located close to a boundary wall, its path length probability distribution function in directions of flight parallel to the wall is significantly different than in other directions. Hence, anisotropic diffusion of neutrons near the boundaries arises. We describe an analysis of neutron transport in a simplified 3-D pebble bed random system, in which we investigate the anisotropic diffusion of neutrons born near one of the system's boundary walls. While this simplified system does not model the actual physical process that takes place near the boundaries of a PBR core, the present work paves the road to a formulation that may enable more accurate diffusion simulations of such problems to be performed in the future. Monte Carlo codes have been developed for (i) deriving realizations of the 3-D random system, and (ii) performing 3-D neutron transport inside the heterogeneous model; numerical results are presented for three different choices of parameters. These numerical results are used to assess the accuracy of estimates for the mean-squared displacement of neutrons obtained with the diffusion approximations of the Atomic Mix Model and of the recently introduced [1] Non-Classical Theory with angular-dependent path length distribution. The Non-Classical Theory makes use of a Generalized Linear Boltzmann Equation in which the locations of the scattering centers in the system are correlated and the distance to collision is not exponentially distributed. We show that the results predicted using the Non-Classical Theory successfully model the anisotropic behavior of the neutrons in the random system, and more closely agree with experiment than the results predicted by the Atomic Mix Model. (authors)

Estimating anisotropic diffusion of neutrons near the boundary of a pebble bed random system

International Nuclear Information System (INIS)

Vasques, R.

2013-01-01

Due to the arrangement of the pebbles in a Pebble Bed Reactor (PBR) core, if a neutron is located close to a boundary wall, its path length probability distribution function in directions of flight parallel to the wall is significantly different than in other directions. Hence, anisotropic diffusion of neutrons near the boundaries arises. We describe an analysis of neutron transport in a simplified 3-D pebble bed random system, in which we investigate the anisotropic diffusion of neutrons born near one of the system's boundary walls. While this simplified system does not model the actual physical process that takes place near the boundaries of a PBR core, the present work paves the road to a formulation that may enable more accurate diffusion simulations of such problems to be performed in the future. Monte Carlo codes have been developed for (i) deriving realizations of the 3-D random system, and (ii) performing 3-D neutron transport inside the heterogeneous model; numerical results are presented for three different choices of parameters. These numerical results are used to assess the accuracy of estimates for the mean-squared displacement of neutrons obtained with the diffusion approximations of the Atomic Mix Model and of the recently introduced [1] Non-Classical Theory with angular-dependent path length distribution. The Non-Classical Theory makes use of a Generalized Linear Boltzmann Equation in which the locations of the scattering centers in the system are correlated and the distance to collision is not exponentially distributed. We show that the results predicted using the Non-Classical Theory successfully model the anisotropic behavior of the neutrons in the random system, and more closely agree with experiment than the results predicted by the Atomic Mix Model. (authors)

Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

Energy Technology Data Exchange (ETDEWEB)

Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

- Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

Neutrons in a highly diffusive medium a new propulsion tool for deep space exploration?

CERN Document Server

Rubbia, Carlo

1998-01-01

The recently completed TARC Experiment at the CERN-PS has shown how it is possible to confine neutrons by diffusion in a limited volume of a highly transparent medium for very long times (tens of milliseconds), with correspondingly very long diffusive paths (> 60 m neutron path ÒwoundÓ within a ~ 60 cm effective radius). Assume an empty cavity is introduced inside the previous volume of diffusing medium. The inner walls of the cavity are covered with a thin layer of highly fissionable material, which acts as a neutron multiplying source. This configuration, called Òn-HohlraumÓ, is reminiscent of a classic black-body radiator, with the exception that now neutrons rather than photons are propagated. The flux can be sufficiently enhanced as to permit to reach criticality with a ~ 1 mm thick Americium deposit, corresponding to a mere 1100 atomic layers. Such a layer is so thin that the Fission Fragments (FF) exit freely into the cavity. The energy carried by FF can be recovered directly, thus making use of th...

Study on neutron diffusion and time dependence heat ina fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Vilhena, M.T. de.

1988-01-01

The purpose of this work is to model the neutron diffusion and heat transfer for a Fluidized Bed Nuclear Reactor and its solution by Laplace Transform Technique with numerical inversion using Fourier Series. Also Gaussian quadrature and residues techniques were applied for numerical inversion. The neutron transport, diffusion, and point Kinetic equation for this nuclear reactor concept are developed. A matricial and Taylor Series methods are proposed for the solution of the point Kinetic equation which is a time scale problem of Stiff type

Asymptotic neutron scattering laws for anomalously diffusing quantum particles

Energy Technology Data Exchange (ETDEWEB)

Kneller, Gerald R. [Centre de Biophysique Moléculaire, CNRS, Rue Charles Sadron, 45071 Orléans (France); Université d’Orléans, Chateau de la Source-Ave. du Parc Floral, 45067 Orléans (France); Synchrotron-SOLEIL, L’Orme de Merisiers, 91192 Gif-sur-Yvette (France)

2016-07-28

The paper deals with a model-free approach to the analysis of quasielastic neutron scattering intensities from anomalously diffusing quantum particles. All quantities are inferred from the asymptotic form of their time-dependent mean square displacements which grow ∝t{sup α}, with 0 ≤ α < 2. Confined diffusion (α = 0) is here explicitly included. We discuss in particular the intermediate scattering function for long times and the Fourier spectrum of the velocity autocorrelation function for small frequencies. Quantum effects enter in both cases through the general symmetry properties of quantum time correlation functions. It is shown that the fractional diffusion constant can be expressed by a Green-Kubo type relation involving the real part of the velocity autocorrelation function. The theory is exact in the diffusive regime and at moderate momentum transfers.

Apports de la simulation multi-agents à l’étude des processus de diffusion

Directory of Open Access Journals (Sweden)

Eric Daudé

2004-02-01

Full Text Available Cet article présente les principaux apports d’une approche individu-centrée pour modéliser et simuler les processus de diffusion dans l’espace géographique. L’intérêt d’une telle approche est discuté au regard des limites d’une modélisation d'échelle macro-géographique traditionnellement employée pour décrire de tels phénomènes. L’équation logistique est présentée comme un exemple représentatif de cette catégorie de modèle. Les modèles de simulation à base d’agents représentent une alternative possible à une approche macro-géographique des phénomènes de diffusion. Associés aux théories de l’auto-organisation, on postule que les phénomènes observés à un certain niveau sont le résultat des nombreuses interactions qui se produisent à un ou plusieurs niveaux inférieurs. En tant qu’outil de simulation prédisposé à l’analyse de telles dynamiques, les systèmes multi-agents sont brièvement présentés. Cet article se termine par une application, la modélisation et la simulation de la diffusion d’une innovation agricole (le modèle d'Hägerstrand revisité.

The pulsed neutron method applied to the determination of the nuclear constants of graphite (1961); La methode des neutrons pulses appliquee a la determination des constantes nucleaires du graphite (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

A method for determining the nuclear constants {sigma}{sub a} and {lambda}{sub t} of a moderator is described. The disappearance of a burst of neutrons introduced into a finite medium is studied as a function of time. This decrease in the thermal neutron density is the product of two exponentials; one representing the absorption, the other the leakage. By varying one or other of these factors, the constants of the factor left unvaried can be determined, and from this the nuclear constant values are deduced. (author) [French] On decrit une methode permettant de determiner les constantes nucleaires {sigma}{sub a} et {lambda}{sub t} d'un moderateur. On etudie la decroissance dans le temps d'une bouffee de neutrons introduite dans un milieu fini. Cette decroissance de la densite en neutrons thermiques est le produit de deux exponentielles; l'une represente l'absorption, l'autre les fuites. Par variation de l'un ou l'autre de ces facteurs, on determine les constantes du facteur laisse invariant dont on deduit les valeurs des constantes nucleaires. (auteur)

Fe and N diffusion in nitrogen-rich FeN measured using neutron ...

Indian Academy of Sciences (India)

E-mail: mgupta@csr.ernet.in. Abstract. Grazing incidence neutron reflectometry provides an opportunity to measure the depth profile of a thin film sample with a resolution <1 nm, in a non-destructive way. In this way the diffusion across the interfaces can also be measured. In addition, neutrons have contrast among the ...

Contribution to the study of thermal diffusivity of solids; Contribution a l'etude de la diffusivite thermique des solides

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

Angstroem method has been reviewed for its application to measurements of thermal diffusivity and conductivity on short specimens. An apparatus and a technique have been developed for rapid and precise measurements of a large variety of materials, which might also contain heat sources. This technique allows measurements at both high and low temperatures. Stainless steel, nickel and uranium monocarbide specimens were tested and the results of the thermal diffusivity measurements between 50 deg. C and 700 deg. C are presented. (author) [French] L'application de la methode d'Angstroem pour la mesure de la diffusivite et de la conductivite thermique sur des echantillons courts est examinee. Un appareillage est decrit, qui permet non seulement des mesures sur une grande variete de materiaux, mais qui est aussi concu pour des mesures rapides, precises et ou des sources thermiques peuvent etre introduites au sein de l'echantillon. La methode s'adapte egalement aux mesures a basses et hautes temperatures. Des resultats de mesure sur un echantillon en acier inoxydable, en nickel et en carbure d'uranium pour des temperatures comprises entre 50 et 700 deg. C sont reportes. (auteur)

EFFET DES TRAITEMENTS THERMIQUES SUR LA REACTION ENTRE DES COUCHES MINCES DE TITANE ET DES SUBSTRATS EN ACIER

Directory of Open Access Journals (Sweden)

D Slimani

2015-06-01

Full Text Available Des couches minces du titane pur ont été déposées avec la méthode de pulvérisation cathodique sur des substrats en acier, type FF80 K-1 contenants ~1% mass. en carbone. La réaction entre les deux parties du système substrat-couche mince est activée avec des traitements thermiques sous vide dans l’intervalle de températures de 400 à900°Cpendant 30 minutes. Les Spectres de diffraction de rayons x confirment l’inter- diffusion des éléments  chimiques du système résultants la formation et la croissance des nouvelles phases en particulier le carbure binaire TiC ayant des caractéristiques thermomécaniques importantes. L’analyse morphologique des échantillons traités  avec le microscope électronique à balayage (MEB montre l’augmentation du flux de diffusion atomique avec la température de recuit, notamment la diffusion du manganèse et du fer vers la surface libre des échantillons aux températures élevées provoquant la dégradation des propriétés mécaniques des revêtements contrairement au premiers stades d’interaction où on a obtenu des bonnes valeurs de la microdureté.

Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

International Nuclear Information System (INIS)

Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

2009-10-01

In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

A multidimensional multigroup diffusion model for the determination of the frequency-dependent field of view of a neutron detector

International Nuclear Information System (INIS)

van der Hagen, T.H.J.J.; Hoogenboom, J.E.; van Dam, H.

1992-01-01

This paper reports on the sensitivity of a neutron detector to parametric fluctuations in the core of a reactor which depends on the position and the frequency of the perturbation. The basic neutron diffusion model for the calculation of this so-called field of view (FOV) of the detector is extended with respect to the dimensionality of the problem and the number of energy groups involved. The physical meaning of the FOV concept is illustrated by means of some simple examples, which can be handled analytically. The possibility of calculating the FOV by a conventional neutron diffusion code is demonstrated. In that case, the calculation in n neutron energy groups leads to 2n modified neutron diffusion equations

Characteristics of membranes which are suitable for isotopic separation by gaseous diffusion; Caracteristiques des barrieres utilisables pour la separation isotopique par diffusion gazeuse

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

In this article we shall briefly describe how results obtained from the experimental analysis of the structure of membranes can be used on the one hand for determining this structure and on the other hand as a result for example of the detection of too many large pores or micro fissures and the existence of leaks at the joints, for eliminating membranes which can not be used industrially. Here we shall only consider the utilisation of information given by measurement of the specific permeability G{sub 0}, the pore radius for the molecular (diffusion) regime r{sub mol.}, the pore radius for the viscous diffusion regime the pore radius for the intermediate stage of diffusion, r{sub wicke}, the pore radius r{sub sep.} which is used for determining the actual separation, the distribution of the pore radius defined by mercury symmetry, the B.E.T. and central X ray diffusion. First of all we shall give the definition which is adopted in practice for these empirical values, which can be adapted for a rapid analysis of the quality of membranes by comparing them with the same measurements carried out on standard porous samples. In a second section we show in a few cases which have occurred in routine measurement how by comparison it is possible to eliminate membranes which do not have a suitable structure. (author) [French] Dans cet expose, nous allons montrer sommairement comment les resultats donnes par les methodes experimentales de controle de la structure des barrieres peuvent, d'une part, conduire a une certaine representation de cette structure et, d'autre part, permettre d'eliminer les barrieres non utilisables industriellement en detectant par exemple les trop nombreux gros pores ou microfissures et l'existence de fuites aux joints. Nous nous limiterons ici a l'exploitation des renseignements donnes par la permeabilite specifique G{sub 0}, le rayon de pore r{sub mol.} effectif en regime moleculaire, le rayon de pore r{sub visq.} effectif en regime visqueux, le

Simulation of a parallel processor on a serial processor: The neutron diffusion equation

International Nuclear Information System (INIS)

Honeck, H.C.

1981-01-01

Parallel processors could provide the nuclear industry with very high computing power at a very moderate cost. Will we be able to make effective use of this power. This paper explores the use of a very simple parallel processor for solving the neutron diffusion equation to predict power distributions in a nuclear reactor. We first describe a simple parallel processor and estimate its theoretical performance based on the current hardware technology. Next, we show how the parallel processor could be used to solve the neutron diffusion equation. We then present the results of some simulations of a parallel processor run on a serial processor and measure some of the expected inefficiencies. Finally we extrapolate the results to estimate how actual design codes would perform. We find that the standard numerical methods for solving the neutron diffusion equation are still applicable when used on a parallel processor. However, some simple modifications to these methods will be necessary if we are to achieve the full power of these new computers. (orig.) [de

Elastic neutron diffuse scattering in Zr(Ca, Y)O2-x

International Nuclear Information System (INIS)

Barberis, P.; Beuneu, B.; Novion, C.H. de.

1990-01-01

Elastic neutron diffuse scattering has been measured in cubic Zr(Ca, Y)O 2-x at room temperature. The very high diffuse scattering (up to 70 Laue) is explained mostly by the oxygen displacements along directions, and by Ca displacements along . The weak short-range order contribution strongly suggests that oxygen vacancies tend to place as second rather than at first neighbours of a Ca stabilizing ion

Neutron diffuse scattering in magnetite due to molecular polarons

International Nuclear Information System (INIS)

Yamada, Y.; Wakabayashi, N.; Nicklow, R.M.

1980-01-01

A detailed neutron diffuse scattering study has been carried out in order to verify a model which describes the property of valence fluctuations in magnetite above T/sub V/. This model assumes the existence of a complex which is composed of two excess electrons and a local displacement mode of oxygens within the fcc primitive cell. The complex is called a molecular polaron. It is assumed that at sufficiently high temperatures there is a random distribution of molecular polarons, which are fluctuating independently by making hopping motions through the crystal or by dissociating into smaller polarons. The lifetime of each molecular polaron is assumed to be long enough to induce an instantaneous strain field around it. Based on this model, the neutron diffuse scattering cross section due to randomly distributed dressed molecular polarons has been calculated. A precise measurement of the quasielastic scattering of neutrons has been carried out at 150 K. The observed results definitely show the characteristics which are predicted by the model calculation and, thus, give evidence for the existence of the proposed molecular polarons. From this standpoint, the Verwey transition of magnetite may be viewed as the cooperative ordering process of dressed molecular polarons. Possible extensions of the model to describe the ordering and the dynamical behavior of the molecular polarons are discussed

A method to measure the diffusion coefficient by neutron wave propagation for limited samples

International Nuclear Information System (INIS)

Woznicka, U.

1986-03-01

A study has been made of the use of the neutron wave and pulse propagation method for measurement of thermal neutron diffusion parameters. Earlier works an homogenous and heterogeneous media are reviewed. A new method is sketched for the determination of the diffusion coefficient for samples of limited size. The principle is to place a relatively thin slab of the material between two blocks of a medium with known properties. The advantages and disadvantages of the method are discussed. (author)

A diffuse neutron scattering study of clustering in copper-nickel alloys

International Nuclear Information System (INIS)

Vrijen, J.

1977-01-01

The amount of clustering in Cu-Ni alloys in thermal equilibrium at several temperatures between 400degC and 700degC and ranging in composition between 20 and 80 atomic percent Ni has been determined by means of diffuse neutron scattering. A rough calculation of the excess elastic energy due to alloying Cu with Ni shows that the contribution of size effects to the configurational energy is asymmetric in the composition with its maximum located between 60 and 70 atomic percent Ni. This asymmetry is caused by different elastic constants for Cu and Ni and it might explain part of the asymmetry of clustering in Cu-Ni and its temperature dependence. With the help of the measured cluster parameters, the magnetic diffuse neutron scattering cross-sections of several differently clustered compositions in Cu-Ni could be interpreted, both well inside the ferromagnetic phase and in the transition region between ferromagnetism and superparamagnetism. Giants moments have been observed. Non-equilibrium distributions and their changes during relaxing towards equilibrium have been investigated by measuring the time-evolution of the diffuse scattering. The relaxation of the null matrix (composition without Bragg reflections for neutron scattering) has been measured at five temperatures between 320degC and 450degC. The results of these relaxations were compared with a few available kinetic models

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

Measure of the efficiency of a long counter of Hanson's type and use of this counter for the survey of the slow neutrons coming from the reactor of Chatillon; Mesure de l'efficacite d'un long compteur du type Hanson et utilisation de ce compteur a l'etude des neutrons lents sortant de la pile de Chatillon

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

A detection device of fast neutrons of efficiency almost independent of the energy of the neutrons has been achieved. It efficiency has been measured in absolute value for groups of neutrons of different energies. This device allowed to get some indications on the energy composition of the neutrons leaving from the reactor of Chatillon. (author) [French] Un dispositif de detection de neutrons rapides d'efficacite pratiquement independante de l'energie des neutrons a ete realise. Son efficacite a ete mesuree en valeur absolue pour des groupes de neutrons de diverses energies. Ce dispositif a permis obtenir quelques indications sur la composition energetique des neutrons sortant de la pile de Chatillon. (auteur)

Kinetics of heterogeneous systems; La cinetique des milieux heterogenes

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

In this report, a general transport theory treatment is presented for the kinetics study as applied to finite heterogeneous systems. The theory is applicable to experiments near the critical point as well as to pulsed neutron experiments on multiplying or non-multiplying lattices. The general method is also applied to exponential experiments on infinite non-diverging lattices. The particularity of the present study is the explicit introduction of heterogeneity in the formulation and the search for the dependence of the parameters on the buckling of the finite medium. As a result of this, the finite medium parameters are in the first place expressed in terms of the corresponding infinite medium ones through the buckling and the anisotropic migration areas, and in the second place all the parameters are expressed as integrals only over an unit cell instead of over the whole pile. A preliminary less detailed study is first made in order to distinguish clearly between what are called 'dynamic parameters' and 'static parameters', and to define the meanings given in this report to these two terms. In the appendices are given approximate one-group treatments for the study of the dynamic fine structure, the time constant in infinite lattices, and the anisotropic diffusion coefficients in non-multiplying lattices. (author) [French] On presente dans ce rapport une methode generale, utilisant la theorie du transport pour l'etude de la cinetique des milieux finis heterogenes. La theorie est applicable aussi bien aux experiences pres de la criticite qu'aux experiences par sources pulsees de neutrons sur des reseaux multiplicateurs ou non-multiplicateurs. La methode generale est aussi appliquee aux experiences exponentielles sur des reseaux infinis non-divergents. La particularite de l'etude est l'introduction explicite de l'heterogeneite dans la formulation et la recherche de la dependance des parametres par rapport au laplacien du reseau fini. Il en resulte d'une part que les

Non probabilistic solution of uncertain neutron diffusion equation for imprecisely defined homogeneous bare reactor

International Nuclear Information System (INIS)

Chakraverty, S.; Nayak, S.

2013-01-01

Highlights: • Uncertain neutron diffusion equation of bare square homogeneous reactor is studied. • Proposed interval arithmetic is extended for fuzzy numbers. • The developed fuzzy arithmetic is used to handle uncertain parameters. • Governing differential equation is modelled by modified fuzzy finite element method. • Fuzzy critical eigenvalues and effective multiplication factors are investigated. - Abstract: The scattering of neutron collision inside a reactor depends upon geometry of the reactor, diffusion coefficient and absorption coefficient etc. In general these parameters are not crisp and hence we get uncertain neutron diffusion equation. In this paper we have investigated the above equation for a bare square homogeneous reactor. Here the uncertain governing differential equation is modelled by a modified fuzzy finite element method. Using modified fuzzy finite element method, obtained eigenvalues and effective multiplication factors are studied. Corresponding results are compared with the classical finite element method in special cases and various uncertain results have been discussed

Etude de la diagraphie neutron du granite de Beauvoir. Effet neutron des altérations et de la matrice du granite. Calibration granite. Porosité totale à l'eau et porosité neutron Analysis of the Beauvoir Granite Neutron Log. Neutron Effect of Alterations and of the Granite Matrix. Granite Calibration. Total Water Porosity and Neutron Porosity

Directory of Open Access Journals (Sweden)

Galle C.

2006-11-01

Full Text Available Cet article rend compte des travaux effectués sur la porosité du granite de Beauvoir (Sondage GPF 1 d'Echassières, Massif Central français. L'objectif de notre étude est de pouvoir obtenir des valeurs représentatives de la saturation en eau (porosité totale à l'eau n du granite de Beauvoir à partir des mesures de porosité neutron PorositéN (diagraphie neutron BRGM sans avoir recours aux mesures sur carottes. Notre démarche est expérimentale et nous avons tenté d'approfondir certains problèmes liés à l'utilisation de la diagraphie neutron dans une roche granitique. Deux facteurs principaux conditionnent la réponse neutron : la concentration en hydrogène de la formation (eau libre et eau de constitution de certains minéraux et la présence d'éléments absorbeurs à forte section de capture comme le gadolinium, le cadmium, le bore, . . . et dans le cas du granite de Beauvoir, le lithium. A partir des mesures de porosité totale à l'eau n sur carottes, des essais de pertes au feu sur poudre qui nous permettent de déterminer la porosité neutron liée à l'eau de constitution PorositéN(OH- et des analyses chimiques avec lesquelles nous évaluons la porosité neutron thermique PorositéN(ox (Programme SNUPAR, Schlumberger liée à la capture neutronique, nous reconstituons la porosité neutron totale PorositéNR du granite de Beauvoir. Pour 7 échantillons caractéristiques du granite de Beauvoir, nous réalisons grâce à ces résultats une nouvelle calibration du taux de comptage neutron initial corrigé du gradient thermique et de l'effet de trou. Grâce à cette opération, il est possible de déterminer, pour les échantillons traités, la porosité neutron du granite avec une calibration granite (PorositéNg et non calcaire (PorositéNc. La connaissance de l'effet neutron de la matrice nous permet enfin d'évaluer la teneur en eau du granite (porosité totale à l'eau et de comparer celle-ci avec la porosité mesurée sur

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

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

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

A scatter model for fast neutron beams using convolution of diffusion kernels

International Nuclear Information System (INIS)

Moyers, M.F.; Horton, J.L.; Boyer, A.L.

1988-01-01

A new model is proposed to calculate dose distributions in materials irradiated with fast neutron beams. Scattered neutrons are transported away from the point of production within the irradiated material in the forward, lateral and backward directions, while recoil protons are transported in the forward and lateral directions. The calculation of dose distributions, such as for radiotherapy planning, is accomplished by convolving a primary attenuation distribution with a diffusion kernel. The primary attenuation distribution may be quickly calculated for any given set of beam and material conditions as it describes only the magnitude and distribution of first interaction sites. The calculation of energy diffusion kernels is very time consuming but must be calculated only once for a given energy. Energy diffusion distributions shown in this paper have been calculated using a Monte Carlo type of program. To decrease beam calculation time, convolutions are performed using a Fast Fourier Transform technique. (author)

Neutron elastic scattering cross-sections measurement on carbon and fluorine in epithermal energy range using PEREN platform; Mesure des sections efficaces de diffusion elastique des neutrons sur le carbone et le fluor dans le domaine epithermique sur la plate-forme PEREN

Energy Technology Data Exchange (ETDEWEB)

Thiolliere, N

2005-10-15

Molten Salt Reactor (MSR) based on Th/U cycle is one of the new generation concepts for nuclear energy production. A typical MSR is a graphite-moderated core with liquid fuel ({sup 7}LiF +ThF{sub 4} + UF{sub 4}). Many numerical studies based on Monte-Carlo codes are currently carried out but the validity of these numerical result relies on the precise knowledge of neutron cross sections used such as elastic scattering on carbon ({sigma}{sub C}), fluorine ({sigma}{sub F}) and lithium 7 ({sigma}{sub Li}). The goal of this work is to obtain {sigma}{sub C} and {sigma}{sub F} between 1 eV and 100 keV. Such measurements have been performed at the Laboratoire de Physique Subatomique et de Cosmologie (LPSC) de Grenoble on the experimental platform PEREN using slowing-down time spectrometers (C and CF{sub 2}) associated to a pulsed neutron generator (GENEPI). Capture rates are obtained for reference materials (Au, Ag, Mo and In) using YAP scintillator coupled to a photo-multiplier. Very precise simulations (MCNP code) of the experimental setup have been performed and comparison with experiments has led to the determination of {sigma}{sub C} and {sigma}{sub F} with accuracies of 1% and 2% respectively. These results show a small discrepancy to evaluated nuclear data file (ENDF). Measures of total cross-sections {sigma}{sub C} and {sigma}{sub F} at higher energy (200 - 600 keV) were also carried out at Centre des Etudes Nucleaires de Bordeaux using a transmission method. Mono-energetic neutrons were produced by protons accelerated by a Van de Graaff accelerator on a LiF target and transmitted neutrons are counted in a proportional hydrogen gaseous detector. Discrepancies of 5% and 9% for {sigma}{sub C} and {sigma}{sub F} respectively with ENDF have been shown. (author)

An extension of diffusion theory for thermal neutrons near boundaries

International Nuclear Information System (INIS)

Alvarez Rivas, J. L.

1963-01-01

The distribution of thermal neutron flux has been measured inside and outside copper rods of several diameters, immersed in water. It has been found that these distributions can be calculated by means of elemental diffusion theory if the value of the coefficient of diffusion is changed. this parameter is truly a diffusion coefficient, which now also depends on the diameter of the rod. Through a model an expression of this coefficient is introduced which takes account of the measurements of the author and of those reported in PIGC P/928 (1995), ANL-5872 (1959), DEGR 319 (D) (1961). This model could be extended also to plane geometry. (Author) 19 refs

The general formulation and practical calculation of the diffusion coefficient in a lattice containing cavities; Formulation generale et calcul pratique du coefficient de diffusion dans un reseau comportant des cavites

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

The calculation of diffusion coefficients in a lattice necessitates the knowledge of a correct method of weighting the free paths of the different constituents. An unambiguous definition of this weighting method is given here, based on the calculation of leakages from a zone of a reactor. The formulation obtained, which is both simple and general, reduces the calculation of diffusion coefficients to that of collision probabilities in the different media; it reveals in the expression for the radial coefficient the series of the terms of angular correlation (cross terms) recently shown by several authors. This formulation is then used to calculate the practical case of a classical type of lattice composed of a moderator and a fuel element surrounded by an empty space. Analytical and numerical comparison of the expressions obtained with those inferred from the theory of BEHRENS shows up the importance of several new terms some of which are linked with the transparency of the fuel element. Cross terms up to the second order are evaluated. A practical formulary is given at the end of the paper. (author) [French] Le calcul des coefficients de diffusion dans un reseau suppose la connaissance d'un mode de ponderation correct des libres parcours des differents constituants. On definit ici sans ambiguite ce mode de ponderation a partir du calcul des fuites hors d'une zone de reacteur. La formulation obtenue, simple et generale, ramene le calcul des coefficients de diffusion a celui des probabilites de collision dans les differents milieux; elle fait apparaitre dans l'expression du coefficient radial la serie des termes de correlation angulaire (termes rectangles), mis en evidence recemment par plusieurs auteurs. Cette formulation est ensuite appliquee au calcul pratique d'un reseau classique, compose d'un moderateur et d'un element combustible entoure d'une cavite; la comparaison analytique et numerique des expressions obtenues avec celles deduites de la theorie de BEHRENS

Contribution to the study of thermal diffusivity of solids; Contribution a l'etude de la diffusivite thermique des solides

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

Angstroem method has been reviewed for its application to measurements of thermal diffusivity and conductivity on short specimens. An apparatus and a technique have been developed for rapid and precise measurements of a large variety of materials, which might also contain heat sources. This technique allows measurements at both high and low temperatures. Stainless steel, nickel and uranium monocarbide specimens were tested and the results of the thermal diffusivity measurements between 50 deg. C and 700 deg. C are presented. (author) [French] L'application de la methode d'Angstroem pour la mesure de la diffusivite et de la conductivite thermique sur des echantillons courts est examinee. Un appareillage est decrit, qui permet non seulement des mesures sur une grande variete de materiaux, mais qui est aussi concu pour des mesures rapides, precises et ou des sources thermiques peuvent etre introduites au sein de l'echantillon. La methode s'adapte egalement aux mesures a basses et hautes temperatures. Des resultats de mesure sur un echantillon en acier inoxydable, en nickel et en carbure d'uranium pour des temperatures comprises entre 50 et 700 deg. C sont reportes. (auteur)

Homogenization of neutronic diffusion models

International Nuclear Information System (INIS)

Capdebosq, Y.

1999-09-01

In order to study and simulate nuclear reactor cores, one needs to access the neutron distribution in the core. In practice, the description of this density of neutrons is given by a system of diffusion equations, coupled by non differential exchange terms. The strong heterogeneity of the medium constitutes a major obstacle to the numerical computation of this models at reasonable cost. Homogenization appears as compulsory. Heuristic methods have been developed since the origin by nuclear physicists, under a periodicity assumption on the coefficients. They consist in doing a fine computation one a single periodicity cell, to solve the system on the whole domain with homogeneous coefficients, and to reconstruct the neutron density by multiplying the solutions of the two computations. The objectives of this work are to provide mathematically rigorous basis to this factorization method, to obtain the exact formulas of the homogenized coefficients, and to start on geometries where two periodical medium are placed side by side. The first result of this thesis concerns eigenvalue problem models which are used to characterize the state of criticality of the reactor, under a symmetry assumption on the coefficients. The convergence of the homogenization process is proved, and formulas of the homogenized coefficients are given. We then show that without symmetry assumptions, a drift phenomenon appears. It is characterized by the mean of a real Bloch wave method, which gives the homogenized limit in the general case. These results for the critical problem are then adapted to the evolution model. Finally, the homogenization of the critical problem in the case of two side by side periodic medium is studied on a one dimensional on equation model. (authors)

Absolute measurement of {beta} activities and application to the determination of neutronic densities; Mesure absolue d'activites {beta} et application a la determination des densites neutronique

Energy Technology Data Exchange (ETDEWEB)

Cohen, R [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1951-01-15

M. Berthelot, to my entrance to the ''Commissariat a l 'Energie Atomique'', proposed me to study the absolute measurement of neutron densities. Very quickly the problem of the absolute activity of {beta} sources became the central object of this work. In a first part, we will develop the methods of absolute determination for {beta} activities. The use of a 4{pi} counter permits to get the absolute activity of all beta radioactive source, susceptible to be put as thin leaf and of period superior than some minutes. The method is independent of the spectra of the measured radioelement. we will describe in the second part some applications which use neutron densities measurement, neutron sources intensities and ratio of cross sections of capture of thermal neutrons. (M.B.) [French] M. Berthelot, a mon entree au ''Commissariat a l 'Energie Atomique'', m'a propose d'etudier la mesure absolue des densites neutroniques. Tres rapidement le probleme de l'activite absolue des sources beta est devenu l'objet central de ce travail. Dans une premiere partie, on abordera les methodes de determination absolue des activites beta. L'utilisation d'un compteur 4{pi} permet d 'obtenir l'activite absolue de toute source radioactive beta, susceptible d'etre mise sous forme de feuille mince et de periode superieure a quelques minutes. La methode est independante du spectre du radioelement mesure. On decrira dans la seconde partie quelques applications a des mesures de densites neutroniques, d'intensites de sources de neutrons et de rapport de sections efficaces de capture de neutrons thermiques. (M.B.)

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

Energy Technology Data Exchange (ETDEWEB)

Jacquet, P.

2011-05-23

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

An analytical solution for the two-group kinetic neutron diffusion equation in cylindrical geometry

International Nuclear Information System (INIS)

Fernandes, Julio Cesar L.; Vilhena, Marco Tullio; Bodmann, Bardo Ernst

2011-01-01

Recently the two-group Kinetic Neutron Diffusion Equation with six groups of delay neutron precursor in a rectangle was solved by the Laplace Transform Technique. In this work, we report on an analytical solution for this sort of problem but in cylindrical geometry, assuming a homogeneous and infinite height cylinder. The solution is obtained applying the Hankel Transform to the Kinetic Diffusion equation and solving the transformed problem by the same procedure used in the rectangle. We also present numerical simulations and comparisons against results available in literature. (author)

Study of magnetic thin films by polarized neutron reflectivity. Off-specular diffusion on periodical structures; Etude de couches minces magnetiques par reflectivite de neutrons polarises. Diffusion non speculaire sur des structures periodiques

Energy Technology Data Exchange (ETDEWEB)

Ott, F

1998-11-26

Theoretical (Zeeman energy effects) and experimental (beam polarisation problems) progress have been made in the understanding of polarized neutron reflectivity with polarisation analysis. It has been shown that modelization and numerical simulations makes it possible to avoid to have to systematically measure a full set of reflectivity curves for each field and temperature condition. It has been possible to determine a magnetic profile as a function of the field in a magnetic bilayer system by using only a few points in the reciprocal space. This technique allows to considerable reduce the experiment time. In single nickel layer systems, we have shown that it is possible to induce magnetic rotation inhomogeneities when these systems are subjects to deformation strains. The effect are related to magneto-elastic constants gradients. In trilayer systems, with a ME constant modulation, we have been able to induce large magnetic rotation gradients. A new magneto-optic technique to measure the magnetization direction without rotating the magnetic field has been developed. The field of neutron reflectivity has been extended to off-specular studies. It has been possible to account quantitatively of the off-specular diffusion on 2-D model systems (prepared by optical lithography). This new technique should make it possible in the future to determine magnetic structures with a in-depth as well as lateral resolution. (author)

SNAP-3D: a three-dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1975-10-01

A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)

Matrix-type multiple reciprocity boundary element method for solving three-dimensional two-group neutron diffusion equations

International Nuclear Information System (INIS)

Itagaki, Masafumi; Sahashi, Naoki.

1997-01-01

The multiple reciprocity boundary element method has been applied to three-dimensional two-group neutron diffusion problems. A matrix-type boundary integral equation has been derived to solve the first and the second group neutron diffusion equations simultaneously. The matrix-type fundamental solutions used here satisfy the equation which has a point source term and is adjoint to the neutron diffusion equations. A multiple reciprocity method has been employed to transform the matrix-type domain integral related to the fission source into an equivalent boundary one. The higher order fundamental solutions required for this formulation are composed of a series of two types of analytic functions. The eigenvalue itself is also calculated using only boundary integrals. Three-dimensional test calculations indicate that the present method provides stable and accurate solutions for criticality problems. (author)

Homotopy analysis method for neutron diffusion calculations

International Nuclear Information System (INIS)

Cavdar, S.

2009-01-01

The Homotopy Analysis Method (HAM), proposed in 1992 by Shi Jun Liao and has been developed since then, is based on a fundamental concept in differential geometry and topology, the homotopy. It has proved useful for problems involving algebraic, linear/non-linear, ordinary/partial differential and differential-integral equations being an analytic, recursive method that provides a series sum solution. It has the advantage of offering a certain freedom for the choice of its arguments such as the initial guess, the auxiliary linear operator and the convergence control parameter, and it allows us to effectively control the rate and region of convergence of the series solution. HAM is applied for the fixed source neutron diffusion equation in this work, which is a part of our research motivated by the question of whether methods for solving the neutron diffusion equation that yield straightforward expressions but able to provide a solution of reasonable accuracy exist such that we could avoid analytic methods that are widely used but either fail to solve the problem or provide solutions through many intricate expressions that are likely to contain mistakes or numerical methods that require powerful computational resources and advanced programming skills due to their very nature or intricate mathematical fundamentals. Fourier basis are employed for expressing the initial guess due to the structure of the problem and its boundary conditions. We present the results in comparison with other widely used methods of Adomian Decomposition and Variable Separation.

An analytical solution of the one-dimensional neutron diffusion kinetic equation in cartesian geometry

International Nuclear Information System (INIS)

Ceolin, Celina; Vilhena, Marco T.; Petersen, Claudio Z.

2009-01-01

In this work we report an analytical solution for the monoenergetic neutron diffusion kinetic equation in cartesian geometry. Bearing in mind that the equation for the delayed neutron precursor concentration is a first order linear differential equation in the time variable, to make possible the application of the GITT approach to the kinetic equation, we introduce a fictitious diffusion term multiplied by a positive small value ε. By this procedure, we are able to solve this set of equations. Indeed, applying the GITT technique to the modified diffusion kinetic equation, we come out with a matrix differential equation which has a well known analytical solution when ε goes to zero. We report numerical simulations as well study of numerical convergence of the results attained. (author)

Moderation of Neutrons Emitted by a Pulsed Source and Neutron Spectrometry Based on Slowing-Down Time; Ralentissement des Neutrons Emis par une Source Pulsee et Leur Spectrometrie en Fonction du Temps de Ralentissement; Zamedlenie nejtronov, ispuskaemykh impul'snym istochnikom, i spektrometriya nejtronov po vremeni zamedleniya; Moderacion de Neutrones Emitidos por una Pitente Pulsada y Espectrometria Neutronica Basada en el Tiempo de Frenado

Energy Technology Data Exchange (ETDEWEB)

Bergman, A. A.; Isakov, A. I.; Kazarnovskij, M. V.; Popov, Ju. P.; Shapiro, F. L. [Fizicheskij Institut Im. P.N. Lebedeva AN SSSR, Moskva, SSSR (Russian Federation)

1965-08-15

-sections. Measurements carried out on a large number of substances, including separated isotopes, have led to the discovery of hitherto unknown resonance levels with small neutron widths and have made it possible to determine for many nuclei the magnitude of the force function for the capture of neutrons with orbital angular momentum 1 = 1. (author) [French] A l'Institut de physique P. N. Lebedev de l'Academie des sciences de l'URSS, on a procede, au cours de ces six dernieres annees, a une serie de travaux visant a etudier le ralentissement non stationnaire des neutrons dans les milieux lourds, a elaborer une methode de spectro- metrie des neutrons en fonction du temps de ralentissement et a appliquer cette methode a l'etude de l'influence que l'energie exerce sur les sections efficaces des reactions nuceaires provoquees par des neutrons d'une energie de 30 keV ou moins. Les auteurs donnent un apercu de ces recherches et en analysent les resultats. Apres avoir examine brievement la theorie du ralentissement non stationnaire et de la thermalisation des neutrons, ils indiquent les donnees fournies par une etude experimentale du ralentissement des neutrons dans le graphite, le fer et le plomb et par l'etude de leur thermalisation dans le plomb. En utilisant une source de neutrons puisee et des detecteurs a resonance, ils ont mesure la distribution des temps de ralentissement pour une serie de valeurs determinees de l'energie finale des neutrons. Ils comparent les resultats obtenus a la theorie qui tient compte du mouvement thermique des atomes du ralentisseur, lequel provoque, dans le cas du plomb, une dispersion mesurable des temps de ralentissement pour des energies inferieures a 10 eV. Dans le domaine d'energies inferieures au seuil cadmium, on a mesure l'influence du temps de ralentissement sur la vitesse moyenne des neutrons dans le plomb; une comparaison a ete faite avec la theorie a plusieurs groupes. Les auteurs decrivent la methode employee pour determiner, selon le temps de

MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

International Nuclear Information System (INIS)

Camiciola, P.; Cundari, D.; Montagnini, B.

1992-01-01

1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

Consumer skills contribute to maintaining and diffusing heritage food products Les compétences des consommateurs contribuent au maintien et à la diffusion d’aliments patrimoniaux

Directory of Open Access Journals (Sweden)

Didier Chabrol

2011-05-01

Full Text Available Our food heritage is not only limited to food products and production techniques, it also includes the consumption skills and technologies (CST used to obtain what the consumer requires from a food product. In a specific food culture, a given food product is linked to specific CSTs. In this paper, we examine this aspect of food heritage through the examples of raw ham in Italy, France and Spain, rice varieties and chocolate. We then discuss two types of conjunction / disjunction between food products and food CSTs: controlled diffusion and diffusion through education. We conclude that the sustainability of localized, “origin” or “terroir” products depends on the existence of skilled consumers. The transmission and diffusion of these skills and technologies are thus factors in the permanence and diffusion of the product itself.Le patrimoine alimentaire n’est pas constitué exclusivement de produits et de techniques de production, mais aussi de compétences et techniques de consommation (CTC utilisées par les consommateurs pour obtenir d’un produit alimentaire l’utilité ou le service qu’ils en attendent. Dans une culture alimentaire donnée, un produit alimentaire spécifique est lié à des CTC spécifiques. Cette dimension du patrimoine alimentaire est examinée à travers les exemples du jambon cru en Italie, en Espagne et en France, de différents riz, et du chocolat. Nous discutons ensuite deux types de cohérences / incohérences entre produits et CTC : contrôle par la source, diffusion par l’éducation. Nous concluons que la durabilité de produits locaux, d’origine ou de terroir dépend de l’existence de consommateurs compétents. La transmission et la diffusion des compétences et techniques de consommation sont donc des facteurs de la pérennité et de la diffusion du produit lui-même.

Neutron spectroscopy of fast hydrogen diffusion in BCC transition metals

International Nuclear Information System (INIS)

Richter, D.; Lottner, V.

1979-01-01

Quasielastic neutron scattering reveals microscopic details of both the time and space development of the H-diffusion process on an atomic scale. After outlining the method on the example of PdH/sub x/, new results on the jump geometry in bcc metals are surveyed. In particular, the anomalous diffusion behavior of H in Nb, Ta, and V at elevated temperature is emphasized, where correlated jump processes are important. The influence of impurities on the H-diffusion process is demonstrated by experiments performed on NbH/sub x/ doped with nitrogen impurities, which act as trapping centers for the diffusing hydrogen. The results are discussed in terms of a two-state random walk model which includes multiple trapping and detrapping processes. The concentration and temperature dependence of the capture and escape rates of traps are obtained

Parallel solutions of the two-group neutron diffusion equations

International Nuclear Information System (INIS)

Zee, K.S.; Turinsky, P.J.

1987-01-01

Recent efforts to adapt various numerical solution algorithms to parallel computer architectures have addressed the possibility of substantially reducing the running time of few-group neutron diffusion calculations. The authors have developed an efficient iterative parallel algorithm and an associated computer code for the rapid solution of the finite difference method representation of the two-group neutron diffusion equations on the CRAY X/MP-48 supercomputer having multi-CPUs and vector pipelines. For realistic simulation of light water reactor cores, the code employees a macroscopic depletion model with trace capability for selected fission product transients and critical boron. In addition to this, moderator and fuel temperature feedback models are also incorporated into the code. The validity of the physics models used in the code were benchmarked against qualified codes and proved accurate. This work is an extension of previous work in that various feedback effects are accounted for in the system; the entire code is structured to accommodate extensive vectorization; and an additional parallelism by multitasking is achieved not only for the solution of the matrix equations associated with the inner iterations but also for the other segments of the code, e.g., outer iterations

Localization of alpha emitters by damage production in a thin film. Application to the study of alpha emitter diffusion in irradiated samples; Localisation des emetteurs alpha par creation de dommages dans un film mince. Application a l'etude de la diffusion des emetteurs alpha dans des echantillons irradies

Energy Technology Data Exchange (ETDEWEB)

Houdaille, B; Perrot, M [Commissariat a l' Energie Atomique, 91 - Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The process of recording {alpha} particles on cellulose nitrate films, called alpha-graphy, is applied to the study of the diffusion of {alpha}-emitting elements in irradiated alloys. The existence of diffusion is shown by attacking the film with concentrated caustic soda after exposition. The insensitivity of the recorder to {beta} {gamma} radiation emitted by the sample after passing in the reactor makes it possible to operate with long exposure times and to detect small diffusions. The concentration-penetration curves are drawn up after carrying out a densitometric analysis of the alpha-graphies. - As the cellulose nitrate is affected only by {alpha} particles of energies of between 0.5 and 4 MeV, it was first necessary to determine the yield of the recorder for {alpha} particles emitted by a thick source, i.e. whose energy varies between 0 and E{sub 0}, E{sub 0} being the energy of the alpha emitter. - The concentration C of the {alpha}-emitter, as a function of the optical density D of the alpha-graphy, and of the exposure time t is given by a simple relationship: C = D/at where a is an experimental constant determined by calibration. It depends on the nature of the cellulose nitrate, of the {alpha}-emitting element and of the alloy studied. (authors) [French] Le procede d'enregistrement des particules alpha sur film de nitrate de cellulose, ou alphagraphie, est applique a l'etude de la diffusion d'elements emetteurs alpha dans des alliages irradies. La diffusion est mise en evidence par une attaque du film de nitrate, apres exposition, dans de la soude concentree. L'insensibilite de l'enregistreur au rayonnement {beta} {gamma}, emis par l'echantillon apres son sejour en pile, permet d'operer sur de longs temps de pose et de detecter des diffusions faibles. Les courbes concentration - penetration sont etablies par exploitation densitometrique des alphagraphies. - Comme le nitrate de cellulose n'est impressionne que par des particules alpha dont l'energie est

Solution of two-dimensional neutron diffusion equation for triangular region by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke; Ishibashi, Hideo

1978-01-01

A two-dimensional neutron diffusion equation for a triangular region is shown to be solved by the finite Fourier transformation. An application of the Fourier transformation to the diffusion equation for triangular region yields equations whose unknowns are the expansion coefficients of the neutron flux and current in Fourier series or Legendre polynomials expansions only at the region boundary. Some numerical calculations have revealed that the present technique gives accurate results. It is shown also that the solution using the expansion in Legendre polynomials converges with relatively few terms even if the solution in Fourier series exhibits the Gibbs' phenomenon. (auth.)

Nested element method in multidimensional neutron diffusion calculations

International Nuclear Information System (INIS)

Altiparmakov, D.V.

1983-01-01

A new numerical method is developed that is particularly efficient in solving the multidimensional neutron diffusion equation in geometrically complex systems. The needs for a generally applicable and fast running computer code have stimulated the inroad of a nonclassical (R-function) numerical method into the nuclear field. By using the R-functions, the geometrical components of the diffusion problem are a priori analytically implemented into the approximate solution. The class of functions, to which the approximate solution belongs, is chosen as close to the exact solution class as practically acceptable from the time consumption point of view. That implies a drastic reduction of the number of degrees of freedom, compared to the other methods. Furthermore, the reduced number of degrees of freedom enables calculation of large multidimensional problems on small computers

On the solution of a few problems of multiple scattering by Monte Carlo method; Sur la solution de quelques problemes de diffusions multiples par la methode de Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

Bluet, J C [Commissariat a l' Energie Atomique, Cadarache (France)

1966-02-01

isotrope. 3.- La particule ne perd pas d'energie au cours des collisions successives. Ces hypotheses sont justifiees par les materiaux et les gammes d'energie etudies. 1er probleme: une cible de sodium de geometrie donnee est soumise a un flux de neutrons. Certains noyaux sont actives. L'activation est ensuite mesuree par un cristal au iodure de sodium. La methode de Monte-Carlo permet de calculer la distribution des noyaux actives dans l'echantillon en tenant compte des diffusions et l'efficacite du compteur. Un disque en verre au Li{sup 6} est soumis a un flux de neutrons perpendiculaire a ses faces planes. Si l'epaisseur du verre n'est pas negligeable devant le libre parcours de diffusion {lambda}, il y aura des collisions et le parcours moyen e' d'un neutron dans le verre sera different de l'epaisseur. e' est calcule par la methode de Monte-Carlo. On en tire une correction relative d'efficacite egale a (e' - e)/ e. 3eme probleme: un detecteur de neutrons est place au fond d'un cylindre entoure d'eau. Un flux de neutrons provient d'une source placee sur l'axe, en avant de la protection; on calcule le rapport du nombre de neutrons allant directement de la source au detecteur au nombre de neutrons faisant un trajet non direct avec des collisions dans l'eau. Pour les trois problemes on donne les resultats trouves pour quelques cas concrets tels qu'ils se sont poses experimentalement, et des courbes plus generales ou on utilise des variables reduites comme le rapport libre parcours/dimension geometrique ou son inverse. On donne egalement les listes d'instructions des programmes, avec les organigrammes complets et les tableaux de correspondances entre symboles. (auteur)

Domain decomposition methods for the neutron diffusion problem

International Nuclear Information System (INIS)

Guerin, P.; Baudron, A. M.; Lautard, J. J.

2010-01-01

The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, simplified transport (SPN) or diffusion approximations are often used. The MINOS solver developed at CEA Saclay uses a mixed dual finite element method for the resolution of these problems. and has shown his efficiency. In order to take into account the heterogeneities of the geometry, a very fine mesh is generally required, and leads to expensive calculations for industrial applications. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose here two domain decomposition methods based on the MINOS solver. The first approach is a component mode synthesis method on overlapping sub-domains: several Eigenmodes solutions of a local problem on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is an iterative method based on a non-overlapping domain decomposition with Robin interface conditions. At each iteration, we solve the problem on each sub-domain with the interface conditions given by the solutions on the adjacent sub-domains estimated at the previous iteration. Numerical results on parallel computers are presented for the diffusion model on realistic 2D and 3D cores. (authors)

The influence of pressure on diffusion leading to intermetallic compounds; Influence de la pression sur les diffusions donnant naissance a des composes intermetalliques

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Beyeler, M; Kirianenko, A; Pernot, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

-ray diffraction study, mainly in the U-Cu system. Experiments in progress seem to show that deviation from stoichiometry cannot be observed in alloys prepared by the classical techniques of fusion. (author) [French] Il a ete montre par A.D. LE CLAIRE, J.L. ZAMBROW et L. CASTLEMAN, que l'application d'une pression uniaxiale parallele a la direction de diffusion peut modifier de facon appreciable la cinetique de croissance des phases intermediaires qui peuvent se former dans cette direction. L'interpretation de ce phenomene prete a discussion, et on essaie donc de l'expliquer par une analyse approfondie des donnees experimentales. Il ressort des etudes microscopiques de la cinetique de croissance des zones, particulierement pour les systemes uranium-cuivre et uranium-nickel, que cette croissance est affectee de maniere similaire par une pression uniaxiale ou par une pression hydrostatique. Par contre, la vitesse de croissance de ces zones augmente en fonction de la pression appliquee pour les systemes uranium-cuivre, uranium-nickel et uranium-aluminium (cet effet etant particulierement marque dans le cas d'uranium-aluminium). Afin de determiner avec precision les limites de la gamme de stabilite des composes intermetalliques, les courbes des caracteristiques de concentration et de penetration ont ete etablies au moyen du micro-analyseur electronique de CASTAING. Les resultats montrent que lorsqu'une diffusion a lieu sans pression exterieure (les couples U-Cu et U-Ni) ou avec une pression a 300 kg/cm{sup 2} (le couple U-Al), la concentration varie de facon notable dans les composes obtenus, ceux-ci etant theoriquement stoechiometriques. On retrouve, par exemple, en traversant la zone de diffusion d'un metal a l'autre, les variations continues suivantes: UCu{sub 4,70} a UCu{sub 5.25} dans le couple U-Cu; UNi{sub 4,75} a UNi{sub 5,25} dans le couple U-Ni; UAl{sub 2,2} a UAl{sub 3,3} dans le couple U-Al. Si l'on applique une pression uni-axiale ou hydrostatique de plus de 500 kg

Absolute measurement of {beta} activities and application to the determination of neutronic densities; Mesure absolue d'activites {beta} et application a la determination des densites neutronique

Energy Technology Data Exchange (ETDEWEB)

Cohen, R. [Commissariat a l' Energie Atomique, Lab. du Fort de Chatillon, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1951-01-15

M. Berthelot, to my entrance to the ''Commissariat a l 'Energie Atomique'', proposed me to study the absolute measurement of neutron densities. Very quickly the problem of the absolute activity of {beta} sources became the central object of this work. In a first part, we will develop the methods of absolute determination for {beta} activities. The use of a 4{pi} counter permits to get the absolute activity of all beta radioactive source, susceptible to be put as thin leaf and of period superior than some minutes. The method is independent of the spectra of the measured radioelement. we will describe in the second part some applications which use neutron densities measurement, neutron sources intensities and ratio of cross sections of capture of thermal neutrons. (M.B.) [French] M. Berthelot, a mon entree au ''Commissariat a l 'Energie Atomique'', m'a propose d'etudier la mesure absolue des densites neutroniques. Tres rapidement le probleme de l'activite absolue des sources beta est devenu l'objet central de ce travail. Dans une premiere partie, on abordera les methodes de determination absolue des activites beta. L'utilisation d'un compteur 4{pi} permet d 'obtenir l'activite absolue de toute source radioactive beta, susceptible d'etre mise sous forme de feuille mince et de periode superieure a quelques minutes. La methode est independante du spectre du radioelement mesure. On decrira dans la seconde partie quelques applications a des mesures de densites neutroniques, d'intensites de sources de neutrons et de rapport de sections efficaces de capture de neutrons thermiques. (M.B.)

Évolution des contraintes résiduelles dans la couche de diffusion d’un acier modèle Fe-Cr-C nitruré

DEFF Research Database (Denmark)

Jegou, Sébastien; Barrallier, Laurent; Somers, Marcel A. J.

2011-01-01

Limiter la fatigue et la corrosion des pièces est possible grâce à une nitruration. Des contraintes résiduelles en découlent. Le rôle de la diffusion du carbone sur le développement de ces contraintes a été étudié sur un acier modèle Fe-3%m.Cr-0.35%m.C.......Limiter la fatigue et la corrosion des pièces est possible grâce à une nitruration. Des contraintes résiduelles en découlent. Le rôle de la diffusion du carbone sur le développement de ces contraintes a été étudié sur un acier modèle Fe-3%m.Cr-0.35%m.C....

Carcinogenesis by Fast Neutrons Relative to X-Rays in Mice; Carcinogenese chez les Souris sous l'Effet des Neutrons Rapides et des Rayons X; Sravnitel'naya chastota vozniknoveniya rakovykh opukholej u myshej pri obluchenii bystrymi nejtronami i rentgenovskimi luchami; Accion Carcinogenica Comparada de los Neutrones Rapidos y de los Rayos X en el Raton

Energy Technology Data Exchange (ETDEWEB)

Cole, L. J.; Nowell, P. C. [Division of Biological and Medical Sciences, U.S. Naval Radiological Defense Laboratory, San Francisco, CA (United States); Department of Pathology, School of Medicine, University of Pennsylvania, Philadelphia, PA (United States)

1964-05-15

Young adult LAF{sub 1} mice received a single sublethal exposure to fission-spectrum neutrons, and were treated as follows: Group 1 - 195 to 199 rad, followed by a single subcutaneous injection of CCI{sub 4} one month later; Group II - CCl{sub 4} injected one day prior to neutron irradiation (280 - 329 rad); Group III - CCI{sub 4} only; no irradiation; Group IV - unilateral nephrectomy carried out one day before neutron irradiation (320 - 328 rad); Group V - neutron irradiation only (200 - 320 rad> In the Group I-mice, sacrificed 12-16 months post-irradiation, 75% exhibited hepatomas, while 92% of the mice in Group II showed hepatomas. By contrast, liver tumours appeared in 14% of mice receiving neutron irradiation only (Group V), and in 2.4% of mice exposed to a single dose (500 rad) of 250 kVp X-rays. The incidence of kidney neoplasms approached 100% in the uninephrectomized irradiated mice (Group IV), whereas such lesions were infrequent in control mice. The induction of these neoplasms is discussed in terms of initiation by the neutron radiation, and promotion by the specific proliferative stimuli exployed. The data indicate a potency factor of approximately 2 for fission neutrons relative to 250 kVp X-rays for initiation of kidney neoplasms in mice. With respect to the occurrence of lymphoma in this strain, the neutron-irradiated mice showed a reduced incidence (11 %) as compared with the X-irradiated animals (29%). (author) [French] On a expose a une dose sub letale unique de neutrons de fission des jeunes sujets adultes de souche LAF, que l'on a traites de la facon suivante: Groupe 1 - injection souscutanee de CCl{sub 4} un mois apres l 'exposition(195 a 199rad); Groupe II - injection de CCl{sub 4} un jour avant l'exposition aux neutrons (280 a 329 rad); Groupe III - CCl{sub 4} seulement, pas d'exposition; Groupe IV - nephrectomie unilaterale un jour avant l'exposition aux neutrons (320 a 328 rad); Groupe V - exposition aux mettrons seulement (200 a 320 rad

Influence of diffusion on extraction kinetics in porous bodies. The case of uranium oxides; Influence de la diffusion sur la cinetique d'extraction dans un corps poreux. Cas des oxydes de l'uranium

Energy Technology Data Exchange (ETDEWEB)

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

1966-12-01

The study of the leaching of heaped uranium ore can be considered theoretically as the problem of the diffusion of liquids in porous bodies and in particular as that of its influence on the chemical reaction rates of conventional uranium oxides. Below a certain value of the pore diameter, it is diffusion which is responsible for mass transfer. The porous structure can be characterized by various physical constants which modify the free diffusion equation and, as long as the pores have a diameter greater than a few microns, it can be shown that the pore walls have a negligible effect on the diffusion. The diffusion coefficients for the nitrate, the sulfate, the chloride, the acetate and the perchlorate of uranium have been determined. In the case of the reaction of uranium trioxide with acids in a porous body, the reaction kinetics are governed by the arrival of the reagent by diffusion. The attack of uranium dioxide by an acid ferric iron solution has been studied under the same conditions and it has been found that the diffusion modifies the influence of the ferrous and ferric iron concentrations on the reaction kinetics. The same is true for the oxide U{sub 3}O{sub 8}. All the results concerning these reactions studied in the absence of the influence of diffusion should be modified to take this factor into account when it intervenes in an extraction process. (authors) [French] L'etude de la lixiviation en tas d'un minerai d'uranium peut se ramener theoriquement au probleme de la diffusion des liquides dans les corps poreux et en particulier a celui de son influence sur les vitesses de reaction chimique des oxydes classiques de l'uranium. En dessous d'une certaine limite de diametre des pores la diffusion est responsable du transfert de masse. La structure poreuse peut se caracteriser par differentes constantes physiques qui modifient l'equation de la diffusion libre et tant que les pores ont un diametre superieur a quelques microns, on a montre que l

DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation

International Nuclear Information System (INIS)

Anderson, E.C.; Putnam, G.E.

1975-01-01

1 - Description of problem or function: DWARF allows one-dimensional simulation of reactor burnup and xenon oscillation problems in slab, cylindrical, or spherical geometry using a few-group diffusion theory model. 2 - Method of solution: The few-group, neutron diffusion theory equations are reduced to a system of finite-difference equations that are solved for each group by the Gauss method at each time point. Fission neutron source iteration can be accelerated with Chebyshev extrapolation. A thermal feedback iterative loop is used to obtain consistent solutions for the distributions of reactor power, neutron flux, and fuel and coolant properties with the neutron group constants functions of the latter. Solutions for the new nuclide concentrations of a time-point are made with the flux assumed constant in the time interval. 3 - Restrictions on the complexity of the problem - Maxima of: 4 groups; 40 regions; 50 macroscopic materials (Only 10 are functions of the feedback variables); 50 nuclides per region; 250 mesh points

Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

International Nuclear Information System (INIS)

Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

2017-01-01

Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

Fast neutron breeder reactor Rapsodie - situation of physics, hydraulic, thermal and dynamics studies and studies of stability early in 1963; Pile rapide rapsodie - point des etudes neutroniques, hydrauliques, thermiques et dynamiques et des etudes de stabilite au debut de l'annee 1963

Energy Technology Data Exchange (ETDEWEB)

NONE

1964-07-01

) [French] Au debut de 1963, il a ete necessaire de faire un choix entre les deux combustibles etudies pour Rapsodie: l'alliage UPuMo avec double gainage Nb et acier inoxydable et l'oxyde mixte UO{sub 2}-PuO{sub 2}. Ce rapport donne les resultats des etudes menees avec les deux types de combustible. On rappelle d'abord les differents modeles qui ont ete etudies et on donne une description detaillee de coeurs alliage et oxyde tels qu'ils sont envisages au debut de 1963. On rapporte ensuite les principales caracteristiques neutroniques des deux coeurs: flux et spectre des neutrons, valeur des barres de securite - compensation, economie des neutrons, puissance specifique, fraction effective des neutrons retardes, temps de vie des neutrons prompts, coefficients de reactivite. On decrit les etudes et essais hydrauliques qui ont ete faits a propos de ces deux coeurs. On discute les criteres qui ont ete a l'origine des calculs de debits. On donne les resultats des essais de pertes de charge, de soulevement des assemblages, de vibration et d'ecoulement fin entre les aiguilles. On discute les constantes utilisees pour les calculs thermiques et on donne les temperatures maximales du sodium et des combustibles alliage et oxyde, les majorations dues aux points chauds et la limitation du taux de combustion de l'aiguille oxyde ayant pour origine la pression des gaz de fission. On rapporte les hypotheses qui ont ete utilisees pour les calculs dynamiques et l'on decrit les differents incidents qui ont ete etudies. On donne les resultats des calculs, pour chaque incident et pour chaque combustible et l'on montre que l'on peut eviter la fusion du combustible ou l'ebullition du sodium, meme dans le cas des hypotheses les plus pessimistes, en agissant sur les caracteristiques de la pile (valeur de la barre de pilotage ou puissance de la pile avec un seul circuit de refroidissement). La stabilite de la pile a ete evaluee avec les hypotheses utilisees pour les calculs dynamiques, a l

Numerical method for solving the three-dimensional time-dependent neutron diffusion equation

International Nuclear Information System (INIS)

Khaled, S.M.; Szatmary, Z.

2005-01-01

A numerical time-implicit method has been developed for solving the coupled three-dimensional time-dependent multi-group neutron diffusion and delayed neutron precursor equations. The numerical stability of the implicit computation scheme and the convergence of the iterative associated processes have been evaluated. The computational scheme requires the solution of large linear systems at each time step. For this purpose, the point over-relaxation Gauss-Seidel method was chosen. A new scheme was introduced instead of the usual source iteration scheme. (author)

Measurement of resonance parameters of cross-sections affecting fast-neutron propagation in various media; Mesure des parametres de resonance de sections efficaces lies a la propagation des neutrons dans differents milieux; Izmerenie parametrov rezonansnoj struktury sechenij, vliyayushchikh na rasprostranenie bystrykh nejtronov v sredakh; Medicion de los parametros de resonancia de las secciones eficaces que afectan a la propagacion de los neutrones rapidos en distintos medios

Energy Technology Data Exchange (ETDEWEB)

Nikolaev, M N; Filippov, V V; Bondarenko, I I [Academy of Sciences, Moscow, Union of Soviet Socialist Republics (Russian Federation)

1962-03-15

In compiling a system of multi-group constants for fast- and intermediate-reactor calculations, it is frequently necessary to account for the self-shielding effect associated with the resonance structure of the cross-sections. The energy region in which the resonance structure has a considerable influence on neutron propagation in the material extends up to several decades of kilo-electron- volts for heavy nuclei and to several mega-electron-volts for intermediate nuclei. To compute resonance effects in calculating group parameters, it is essential to know not only the average interaction cross-sections between neutrons and materials, but also values such as (1/{Sigma}{sub t}), (1/{Sigma}{sub t}{sup 2}), etc. - the brackets signify averaging over the energy group. Present information about resonance parameters for fast neutrons is not sufficient to calculate these values with the necessary precision. In this connection it is interesting to take direct measurements of the values (1/{Sigma}{sub t}), (1/{Sigma}{sub t}{sup 2}) and other analogous characteristics. The paper describes the results obtained from measurements of a number of parameters such as ({Sigma}{sub t}), ({Sigma}{sub t}{sup 2}), (1/{Sigma}{sub t}), (1/{Sigma}{sub t}{sup 2}), (1/{Sigma}{sub t}{sup 3}), etc. which characterize the cross-section structure of a number of intermediate nuclei within the range from 300 keV to 3 MeV. These values were arrived at by an analysis of transmission curves obtained by good geometry for transmissions up to {approx}10{sup -2}. The data show that resonance effects exercise a strong influence on the diffusion characteristics of the material. (author) [French] Lorsqu'on etablit un systeme de constantes multi-groupes pour les calculs de reacteurs a neutrons rapides et intermediaires, il est souvent necessaire de tenir compte de l'effet d'auto-protectio n lie a la structure de resonance des sections efficaces. Le domaine d'energie dans lequel la structure de resonance a

Development of 3D multi-group neutron diffusion code for hexagonal geometry

International Nuclear Information System (INIS)

Sun Wei; Wang Kan; Ni Dongyang; Li Qing

2013-01-01

Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)

Determination of thermal neutrons diffusion length in graphite; Determinacion de la Longitud de Difusion de los Neutrones Termicos en Grafito

Energy Technology Data Exchange (ETDEWEB)

Garcia Fite, J

1959-07-01

The diffusion length of thermal neutrons in graphite using the less possible quantity of material has been determined. The proceeding used was the measurement in a graphite pile which has a punctual source of rapid neutrons inside surrounded by a reflector medium (paraffin or water). The measurement was done in the following conditions: a) introducing an aluminium plate between both materials. b) Introducing a cadmium plate between both materials. (Author) 91 refs.

A comparison of Nodal methods in neutron diffusion calculations

Energy Technology Data Exchange (ETDEWEB)

Tavron, Barak [Israel Electric Company, Haifa (Israel) Nuclear Engineering Dept. Research and Development Div.

1996-12-01

The nuclear engineering department at IEC uses in the reactor analysis three neutron diffusion codes based on nodal methods. The codes, GNOMERl, ADMARC2 and NOXER3 solve the neutron diffusion equation to obtain flux and power distributions in the core. The resulting flux distributions are used for the furl cycle analysis and for fuel reload optimization. This work presents a comparison of the various nodal methods employed in the above codes. Nodal methods (also called Coarse-mesh methods) have been designed to solve problems that contain relatively coarse areas of homogeneous composition. In the nodal method parts of the equation that present the state in the homogeneous area are solved analytically while, according to various assumptions and continuity requirements, a general solution is sought out. Thus efficiency of the method for this kind of problems, is very high compared with the finite element and finite difference methods. On the other hand, using this method one can get only approximate information about the node vicinity (or coarse-mesh area, usually a feel assembly of a 20 cm size). These characteristics of the nodal method make it suitable for feel cycle analysis and reload optimization. This analysis requires many subsequent calculations of the flux and power distributions for the feel assemblies while there is no need for detailed distribution within the assembly. For obtaining detailed distribution within the assembly methods of power reconstruction may be applied. However homogenization of feel assembly properties, required for the nodal method, may cause difficulties when applied to fuel assemblies with many absorber rods, due to exciting strong neutron properties heterogeneity within the assembly. (author).

Pathogenesis of Intrapulmonary Haemorrhage in Dogs Exposed to Pulsed Fission-Spectrum Neutrons; Pathogenese de l'Hemorragie Intrapulmonaire chez des Chiens Exposes a des Neutrons de Fission Pulses; Patogenez vnutrilegochnykh krovoizliyanij u sobak pri obluchenii impul'snymi nejtronami spektra deleniya; Patogenesis de la Hemorragia Intrapulmonar en Perros Expuestos a Neutrones de Fision Pulsados

Energy Technology Data Exchange (ETDEWEB)

Jones, R. K.; Engel, R. E.; Godden, W. R. [Kirtland AFB, New Mexico (United States)

1964-05-15

marrow megakaryocytes was found to be inversely proportional to the severity of intrapulmonary haemorrhage. (author) [French] Des comparaisons precedemment effectuees dans le laboratoire des auteurs entre les modifications tissulaires observees chez des chiens exposes a des rayons X de 250 kV-crete et a des neutrons de fission puises ont montre que, dans les deux cas, les reactions etaient en general semblables. Cependant, les poumons des animaux exposes aux neutrons de fission ont ete beaucoup plus souvent le siege d'hemorragies peribronchiales et peri vasculaires. On a essaye de determiner la pathogenese de cette lesion en exposant cinq bigles a une dose de 400 rad (DMA) de neutronsrapides emanant d'un reacteur du type GODIVA et en sacrifiant successivement les animaux a intervalles reguliers. Cinq jours apres l 'exposition, on a pu discerner un ensemble d'alterations pathologiques constituees par de petites hemorragies pulmonaires sporadiques a la peripherie. Leur nombre a augmente progressivement vers le neuvieme jour. Les modifications les plus frappantes sont survenues au bout de 13 j sous forme de grandes hemorragies perivasculaires autour des ramuscules des arteres pulmonaires, depuis le hile jusqu' a la suriace de la plevre. Les premieres modifications histologiques observees dans les poumons sont survenues le cinquieme jour, sour forme de foyers nettement demarques de congestion alveolaire,- de gonflement des cellules capillaires endotheliales et d'extravasion alveolaire legere des globules rouges. Ces alterations se sont generalisees au bout de neuf jours et on.a pu observer en outre des foyers d'extravasion peribronchiale du sang autour des deux bronches primaires. On a constate, chez les animaux autopsies le treizieme jour apres l ' exposition, une grosse hemorragie-peri vasculaire autour des grandes et des petites arteres pulmonaires. Le sang se trouvait dans le tissu conjonctif externe et dans les vaisseaux lymphatiques periarteriels. Les vaisseaux

Les mécanismes des réductions budgétaires et leurs conséquences sur la diffusion de la langue française : Analyse de la situation en 2013

Directory of Open Access Journals (Sweden)

Frédéric Mazières

2013-12-01

Full Text Available Considérer la diplomatie linguistique sans se référer aux crédits qui l’alimentent peut entraîner des erreurs d’appréciation. On évoque souvent les budgets linguistiques sans vraiment connaître leurs modalités de fonctionnement. La qualité et l’efficacité de la diffusion de la langue française dépendent des paramètres didactiques, soit, mais aussi des paramètres budgétaires. Nous retrouvons un des axes « méta-didactiques » de la diffusion linguistique. Cet article est, au moins, l’occasion d’une explication des modalités budgétaires de la diffusion linguistique.

Study of the effect of neutron and electron irradiations on the low temperature thermal conductivity of germanium and silicon; Etude de l'effet des irradiations neutronique et electronique sur la conductibilite thermique aux basses temperatures du germanium et du silicium

Energy Technology Data Exchange (ETDEWEB)

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

1967-06-15

The main results obtained from this work are the following: 1 Neutron irradiation (at 300 deg. K) produces lattice defects in germanium and silicon, and a corresponding very large lowering of the thermal conductivity is observed in the low temperature region (4-300 ). The results obtained have been explained with the help of the following hypotheses: for silicon a scattering of phonons by the stress fields produced by the defects; for germanium, a supplementary scattering of the electron phonon type. 2 Annealing treatments carried out on these materials above 373 deg. K restored the thermal conductivity over the whole temperature range of the measurements (4-300 deg. K); in the case of both germanium and silicon there were two steps in the annealing process. 3 A study of the thermal conductivity of germanium (initially P or N) after an electronic irradiation showed that the scattering of phonons could depend on the state of charge of the defects thus produced. (author) [French] Les principaux resultats obtenus au cours de ce travail sont les suivants : 1 Les irradiations neutroniques (a 300 deg. K) introduisent des defauts de reseau dans le germanium et le silicium et l'on observe correlativement pour ces materiaux, une tres importante diminution de conductibilite thermique dans le domaine des basses temperatures (4-300 deg. K). Les resultats obtenus ont pu etre interpretes en admettant principalement: pour le silicium, une diffusion des phonons par les champs de contrainte dus aux defauts; pour le germanium, une diffusion additionnelle du type electron-phonon. 2 Des recuits effectues sur ces materiaux au-dessus de 373 deg. K ont montre une restauration de la conductibilite thermique dans tout l'intervalle de temperature de mesure (4-300 deg. K) et comportant pour le germanium et le silicium, deux etapes de recuit 3 L'etude de la conductibilite thermique de germanium (initialement N ou P) apres une irradiation electronique, a montre que la diffusion des phonons

Generalization of the Fourier Convergence Analysis in the Neutron Diffusion Eigenvalue Problem

International Nuclear Information System (INIS)

Lee, Hyun Chul; Noh, Jae Man; Joo, Hyung Kook

2005-01-01

Fourier error analysis has been a standard technique for the stability and convergence analysis of linear and nonlinear iterative methods. Lee et al proposed new 2- D/1-D coupling methods and demonstrated several advantages of the new methods by performing a Fourier convergence analysis of the methods as well as two existing methods for a fixed source problem. We demonstrated the Fourier convergence analysis of one of the 2-D/1-D coupling methods applied to a neutron diffusion eigenvalue problem. However, the technique cannot be used directly to analyze the convergence of the other 2-D/1-D coupling methods since some algorithm-specific features were used in our previous study. In this paper we generalized the Fourier convergence analysis technique proposed and analyzed the convergence of the 2-D/1-D coupling methods applied to a neutron diffusion Eigenvalue problem using the generalized technique

Finite difference solution of the time dependent neutron group diffusion equations

International Nuclear Information System (INIS)

Hendricks, J.S.; Henry, A.F.

1975-08-01

In this thesis two unrelated topics of reactor physics are examined: the prompt jump approximation and alternating direction checkerboard methods. In the prompt jump approximation it is assumed that the prompt and delayed neutrons in a nuclear reactor may be described mathematically as being instantaneously in equilibrium with each other. This approximation is applied to the spatially dependent neutron diffusion theory reactor kinetics model. Alternating direction checkerboard methods are a family of finite difference alternating direction methods which may be used to solve the multigroup, multidimension, time-dependent neutron diffusion equations. The reactor mesh grid is not swept line by line or point by point as in implicit or explicit alternating direction methods; instead, the reactor mesh grid may be thought of as a checkerboard in which all the ''red squares'' and '' black squares'' are treated successively. Two members of this family of methods, the ADC and NSADC methods, are at least as good as other alternating direction methods. It has been found that the accuracy of implicit and explicit alternating direction methods can be greatly improved by the application of an exponential transformation. This transformation is incompatible with checkerboard methods. Therefore, a new formulation of the exponential transformation has been developed which is compatible with checkerboard methods and at least as good as the former transformation for other alternating direction methods

Neutrons detection by scintillation; Detection de neutrons par scintillations

Energy Technology Data Exchange (ETDEWEB)

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

1955-01-15

The absence of charge of neutrons and their elevated penetration power make difficult their detection. Techniques vary otherwise with the energy of the particle. The author proposes the realization of a scintillation detector with a big volume of liquid scintillator and containing boron for the detection of slowing-down neutrons in the domain of intermediate energies from 1 to 10{sup 5} eV about. (M.B.) [French] L'absence de charge du neutron et son pouvoir de penetration eleve rendent difficile sa detection. Les techniques par ailleurs varient avec l'energie de cette particule. L'auteur propose la realisation d'un detecteur a scintillations comprenant un grand volume de scintillateur liquide et contenant du bore pour la detection des neutrons en ralentissement dans le domaine des energies intermediaires de 1 a 10{sup 5} eV environ. (M.B.)

Chapter 9: Experimental measurements of the diffusion area of neutrons in graphite

International Nuclear Information System (INIS)

Brown, G.; McCulloch, D.B.

1963-01-01

This report describes measurements of the diffusion area of neutrons in a solid graphite exponential stack, and in a stack containing cylindrical air channels of 4.5 in. diameter, arranged on a square lattice of 8 in. pitch. The resulting diffusion area ratios are compared with the theoretical predictions of a number of authors. The diffusion area ratios deduced from a pair of experiments in which the orientation of the air channels with respect to the source-plane is changed are found to be in agreement with those deduced from experiments in which the stack size is changed but a constant air channel orientation maintained. (author)

Thermal neutron scattering from a hydrogen-metal system in terms of a general multi-sublattice jump diffusion model

International Nuclear Information System (INIS)

Kutner, R.; Sosnowska, I.

1977-01-01

A Multi-Sublattice Jump Diffusion Model (MSJD) for hydrogen diffusion through interstitial-site lattices is presented. The MSJD approach may, in principle, be considered as an extension of the Rowe et al (J. Phys. Chem. Solids; 32:41 (1971)) model. Jump diffusion to any neighbours with different jump times which may be asymmetric in space is discussed. On the basis of the model a new method of calculating the diffusion tensor is advanced. The quasielastic, double differential cross section for thermal neutron scattering is obtained in terms of the MSJD model. The model can be used for systems in which interstitial jump diffusion of impurity particles occurs. In Part II the theoretical results are compared with those for quasielastic neutron scattering from the αNbHsub(x) system. (author)

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1975-10-01

The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

Inter-atomic force constants of BaF{sub 2} by diffuse neutron scattering measurement

Energy Technology Data Exchange (ETDEWEB)

Sakuma, Takashi, E-mail: sakuma@mx.ibaraki.ac.jp; Makhsun,; Sakai, Ryutaro [Institute of Applied Beam Science, Ibaraki University, Mito 310-8512 (Japan); Xianglian [College of Physics and Electronics Information, Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Takahashi, Haruyuki [Institute of Applied Beam Science, Ibaraki University, Hitachi 316-8511 (Japan); Basar, Khairul [Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Igawa, Naoki [Quantum Beam Science Directorate, Japan Atomic Energy Agency, Tokai 319-1195 (Japan); Danilkin, Sergey A. [Bragg Institute, Australian Nuclear Science and Technology Organisation, Kirrawee DC NSW 2232 (Australia)

2015-04-16

Diffuse neutron scattering measurement on BaF{sub 2} crystals was performed at 10 K and 295 K. Oscillatory form in the diffuse scattering intensity of BaF{sub 2} was observed at 295 K. The correlation effects among thermal displacements of F-F atoms were obtained from the analysis of oscillatory diffuse scattering intensity. The force constants among neighboring atoms in BaF{sub 2} were determined and compared to those in ionic crystals and semiconductors.

Domain decomposition method for solving the neutron diffusion equation

International Nuclear Information System (INIS)

Coulomb, F.

1989-03-01

The aim of this work is to study methods for solving the neutron diffusion equation; we are interested in methods based on a classical finite element discretization and well suited for use on parallel computers. Domain decomposition methods seem to answer this preoccupation. This study deals with a decomposition of the domain. A theoretical study is carried out for Lagrange finite elements and some examples are given; in the case of mixed dual finite elements, the study is based on examples [fr

Neutrons detection by scintillation; Detection de neutrons par scintillations

Energy Technology Data Exchange (ETDEWEB)

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

1955-01-15

The absence of charge of neutrons and their elevated penetration power make difficult their detection. Techniques vary otherwise with the energy of the particle. The author proposes the realization of a scintillation detector with a big volume of liquid scintillator and containing boron for the detection of slowing-down neutrons in the domain of intermediate energies from 1 to 10{sup 5} eV about. (M.B.) [French] L'absence de charge du neutron et son pouvoir de penetration eleve rendent difficile sa detection. Les techniques par ailleurs varient avec l'energie de cette particule. L'auteur propose la realisation d'un detecteur a scintillations comprenant un grand volume de scintillateur liquide et contenant du bore pour la detection des neutrons en ralentissement dans le domaine des energies intermediaires de 1 a 10{sup 5} eV environ. (M.B.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Concentration transients in a gaseous diffusion plant (1961); Cinetique des concentrations dans une usine de separation isotopique (1961)

Energy Technology Data Exchange (ETDEWEB)

Jacques, R; Bilous, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

Concentration transients are examined in the case of a gaseous diffusion plant for uranium isotope separation. An application is made for a plant built with two rectifying cascades of different sizes and a stripping cascade. Transients are calculated for a change in the feed concentration, the transport and also for shutdown of a group of separating stages in one of the cascades. (authors) [French] On examine l'evolution des concentrations dans une usine de separation isotopique de l'uranium basee sur le procede de diffusion gazeuse et formee de cascades carrees. Une application est faite pour une installation formee de deux cascades enrichissantes de tailles differentes et d'une cascade appauvrissante. On calcule en particulier les regimes transitoires apres variation de la concentration d'alimentation, du transport et apres mise hors circuit d'un groupe d'etages dans l'une des cascades. (auteurs)

One-velocity neutron diffusion calculations based on a two-group reactor model

Energy Technology Data Exchange (ETDEWEB)

Bingulac, S; Radanovic, L; Lazarevic, B; Matausek, M; Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1965-07-01

Many processes in reactor physics are described by the energy dependent neutron diffusion equations which for many practical purposes can often be reduced to one-dimensional two-group equations. Though such two-group models are satisfactory from the standpoint of accuracy, they require rather extensive computations which are usually iterative and involve the use of digital computers. In many applications, however, and particularly in dynamic analyses, where the studies are performed on analogue computers, it is preferable to avoid iterative calculations. The usual practice in such situations is to resort to one group models, which allow the solution to be expressed analytically. However, the loss in accuracy is rather great particularly when several media of different properties are involved. This paper describes a procedure by which the solution of the two-group neutron diffusion. equations can be expressed analytically in the form which, from the computational standpoint, is as simple as the one-group model, but retains the accuracy of the two-group treatment. In describing the procedure, the case of a multi-region nuclear reactor of cylindrical geometry is treated, but the method applied and the results obtained are of more general application. Another approach in approximate solution of diffusion equations, suggested by Galanin is applicable only in special ideal cases.

Domain decomposition methods for the mixed dual formulation of the critical neutron diffusion problem

International Nuclear Information System (INIS)

Guerin, P.

2007-12-01

The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, diffusion approximation is often used. For this problem, the MINOS solver based on a mixed dual finite element method has shown his efficiency. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose in this dissertation two domain decomposition methods for the resolution of the mixed dual form of the eigenvalue neutron diffusion problem. The first approach is a component mode synthesis method on overlapping sub-domains. Several Eigenmodes solutions of a local problem solved by MINOS on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is a modified iterative Schwarz algorithm based on non-overlapping domain decomposition with Robin interface conditions. At each iteration, the problem is solved on each sub domain by MINOS with the interface conditions deduced from the solutions on the adjacent sub-domains at the previous iteration. The iterations allow the simultaneous convergence of the domain decomposition and the eigenvalue problem. We demonstrate the accuracy and the efficiency in parallel of these two methods with numerical results for the diffusion model on realistic 2- and 3-dimensional cores. (author)

New applications of radio-isotopes in France; Quelques nouvelles applications des radioelements en France

Energy Technology Data Exchange (ETDEWEB)

Leveque, P; Hours, R; Martinelli, P; May, S; Sandier, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Brillant, J [Societe Soletanche (France)

1958-07-01

By measuring the transmission of a flat beam of thermal neutrons, the moisture content of a parallelepiped shaped soil sample can be measured to {+-} 4 per cent and the moisture gradient along the longitudinal axis determined. The method permits the determination of chemically bound water and the measurement of diffusion coefficients of water into low hydrogenated materials. By measuring the intensity of fluorescence excited by 13 radiation it is possible to determine the thickness of metal coatings of less than 20 p. for metals of atomic number less than 40. This method has been applied to chromium, manganese, iron, cobalt, nickel, copper and zinc. By using a suitable metal filter it is possible to measure coating thicknesses of metals differing by only one atomic number from the supporting material. By employing labeled cement it is possible to determine the extent or movement of cement grout used for soil stabilization and waterproofing. The kinetic of ion exchange of different ultramarines in aqueous solutions were studied by tracing the movement of labeled ions in the solution or in the exchanger. Values of the diffusion coefficients and activation energies were determined from the exchange studies. (author)Fren. [French] La mesure de la transmission d'un faisceau plat de neutrons thermiques permet d'evaluer a {+-} 4 pour cent l'humidite d'un echantillon parallelepipedique de sol et d'en effectuer une exploration longitudinale systematique. La methode peut s'appliquer a la determination de l'eau de constitution ainsi qu'a la mesure des coefficients de diffusion de l'eau a travers des substances faiblement hydrogenees. En utilisant la fluorescence X excitee au moyen de rayons {beta} il est possible de mesurer des epaisseurs (< 20) de depots metalliques de faible numero atomique (Z < 40). Ces essais ont porte sur: Cr, Mn, Fe, Co, Ni, Cu et Zn. L'emploi d'ecrans metalliques convenables permet de mesurer l'epaisseur d'un metal dont le numero atomique differe de 1

Asymptotic equivalence of neutron diffusion and transport in time-independent reactor systems

International Nuclear Information System (INIS)

Borysiewicz, M.; Mika, J.; Spiga, G.

1982-01-01

Presented in this paper is the asymptotic analysis of the time-independent neutron transport equation in the second-order variational formulation. The small parameter introduced into the equation is an estimate of the ratio of absorption and leakage to scattering in the system considered. When the ratio tends to zero, the weak solution to the transport problem tends to the weak solution of the diffusion problem, including properly defined boundary conditions. A formula for the diffusion coefficient different from that based on averaging the transport mean-free-path is derived

Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method

Energy Technology Data Exchange (ETDEWEB)

Zanette, Rodrigo; Petersen, Caudio Zen [Univ. Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcello [Univ. Federal de Pelotas (Brazil). Centro de Engenharias; Zabadal, Jorge Rodolfo [Univ. Federal do Rio Grande do Sul, Tramandai (Brazil)

2017-05-15

In this paper a solution for the one-dimensional steady state Multilayer Multigroup Neutron Diffusion Equation in cartesian geometry by Fictitious Borders Power Method and a perturbative analysis of this solution is presented. For each new iteration of the power method, the neutron flux is reconstructed by polynomial interpolation, so that it always remains in a standard form. However when the domain is long, an almost singular matrix arises in the interpolation process. To eliminate this singularity the domain segmented in R regions, called fictitious regions. The last step is to solve the neutron diffusion equation for each fictitious region in analytical form locally. The results are compared with results present in the literature. In order to analyze the sensitivity of the solution, a perturbation in the nuclear parameters is inserted to determine how a perturbation interferes in numerical results of the solution.

Patterns of Lethality and Absorbed Dose Distributions in Mice for Monoenergetic Neutrons; Letalite et Distribution de la Dose Absorbee chez la Souris pour des Neutrons Monoenergetiques; Letal'nost' i raspredelenie pogloshchennoj dozy pri obluchenii myshej monoehnergeticheskimi neitronami; Letalidad y Distribucion de las Dosis Absorbidas por el Raton para Neutrones Monoenergeticos

Energy Technology Data Exchange (ETDEWEB)

Frigerio, N. A.; Jordan, D. L. [Argonne National Laboratory, Argonne, IL (United States)

1964-03-15

gamma and epithermal contaminations, produce the early ''intestinal death'' almost exclusively. (author) [French] La presence de fortes resonances de C, N et O dans la region du spectre situee entre 100 et 1500 keV a permis d'etudier les interactions specifiques neutrons-nucleides d'apres-la letalite, les valeurs maxima de l'EBR, etc. Grace a la reaction {sup 7}Li(p, n){sup 7}Be, on a produit des neutrons monoenergetiques avec des protons de 62 {mu}A resolus a leur sortie d*une machine van de Graaf et dont l'energie allait de 1882 a 2738 keV. Des souris femelles vierges CF-1, placees dans des capsules en celluloide, ont ete exposees aux neutrons monoenergetiques, a des distances de 3,1 a 11,3 cm de la source, sous des angles de 0 a 1 radian. On les a exposees bilateralement, au cours d*un deplacement sur une orbite, circulaire ou elliptique, normale a l'axe du faisceau. On pouvait ainsi controler la distribution de la dose dans le corps de l'animal. On a procede a des mesures absolues du flux avec des compteurs a fission a {sup 235}U et par comptage absolu a l'aide de fils et de feuilles d'or actives a revetement de cadmium. On a mesure les doses absorbees a l'aide de chambres a ionisation et d'un dosimetre a FeSO{sub 4}-NH{sub 4}SCN de haute sensibilite, tout specialement etudie a cette fin. On a procede ensuite a des mesures relatives de la dose avec des compteurs de Hurst a gaz et protons de recul et avec des scintillateurs a {sup 10}B, {sup 6}Li et protons de recul. Les energies des neutrons ont ete mesurees avec des spectrometres speciaux a {sup 10}B, {sup 3}H, {sup 6}Li et semi-conducteurs. On a mesure l'action des rayons gamma avec des chamb rescompteuis a Ne/Ar. Ces mesures ont montre que les rayons gamma constituent moins de 0,8% et les neutrons thermo-epithermiques moins de 0 ,01% de la dose totale exprimee en rad. Les animaux ont ete exposes a des doses moyennes dans l'axe median de 180 a 1200 rad, les energies des neutrons s' echelonnant de 396 a 658 ke

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)

Inelastic Neutron Scattering from Doped Germanium and Silicon; Diffusion Inelastique des Neutrons dans du Germanium et du Silicium Contenant une 'Impurete'; Neuprugoe rasseyanie nejtronov na germanii i kremnii s prisadkoj; Dispersion Inelastica de Neutrones en Germanio y Silicio Deliberadamente Impurificados

Energy Technology Data Exchange (ETDEWEB)

Dolling, G. [Chalk River Nuclear Laboratories, Chalk River, ON (Canada)

1965-04-15

The normal modes of vibration of the pure semiconductors germanium and silicon have been extensively studied by means of coherent one-phonon scattering of slow neutrons from single-crystal specimens. The present paper describes similar experiments performed (i) on germanium heavily doped ({approx}0.1%) with (a) arsenic and (b) gallium, and also (ii) on silicon doped with phosphorus. In each case, control experiments were carried out on high- purity crystals. All measurements were performed with the triple-axis crystal spectrometer at Chalk River Nuclear Laboratories. The elastic constant C{sub 44} for germanium is known to be appreciably dependent on the dopant concentration, and so certain transverse acoustic (TA) modes of long wavelength were studied to see if such effects persisted into the dispersive region. Other TA modes whose frequencies could be measured with high precision were also studied in both materials to check as sensitively as possible for small effects which might be ascribed to the existence of excess electrons or holes. A particularly careful study was made of modes having the following wave vectors (aq/2{pi}, where a is the cubic unit cell side): (i) in germanium, (1,0,0), (ii) in silicon, (0.85, 0.85, 0) and (0.3, 0, 0). Such normal modes might be expected to show anomalous behaviour in the n-type crystals, since inter-valley scattering of electrons between adjacent conduction band minima would require their co-operation in order to conserve ''crystal momentum''. The results in all cases were negative, i. e. no differences in phonon frequencies or energy widths between pure and doped specimens were observed, within the experimental accuracy. In the most favourable cases, this (relative) accuracy is about 0.5% in frequency, rising to 2.0% for certain longitudinal optic modes in silicon. (author) [French] L'auteur a etudie, a l'aide d'experiences de diffusion coherente (a un phonon) de neutrons lents par des monocristaux, les modes normaux de

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-15

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

Rhodium SPND's Error Reduction using Extended Kalman Filter combined with Time Dependent Neutron Diffusion Equation

International Nuclear Information System (INIS)

Lee, Jeong Hun; Park, Tong Kyu; Jeon, Seong Su

2014-01-01

The Rhodium SPND is accurate in steady-state conditions but responds slowly to changes in neutron flux. The slow response time of Rhodium SPND precludes its direct use for control and protection purposes specially when nuclear power plant is used for load following. To shorten the response time of Rhodium SPND, there were some acceleration methods but they could not reflect neutron flux distribution in reactor core. On the other hands, some methods for core power distribution monitoring could not consider the slow response time of Rhodium SPND and noise effect. In this paper, time dependent neutron diffusion equation is directly used to estimate reactor power distribution and extended Kalman filter method is used to correct neutron flux with Rhodium SPND's and to shorten the response time of them. Extended Kalman filter is effective tool to reduce measurement error of Rhodium SPND's and even simple FDM to solve time dependent neutron diffusion equation can be an effective measure. This method reduces random errors of detectors and can follow reactor power level without cross-section change. It means monitoring system may not calculate cross-section at every time steps and computing time will be shorten. To minimize delay of Rhodium SPND's conversion function h should be evaluated in next study. Neutron and Rh-103 reaction has several decay chains and half-lives over 40 seconds causing delay of detection. Time dependent neutron diffusion equation will be combined with decay chains. Power level and distribution change corresponding movement of control rod will be tested with more complicated reference code as well as xenon effect. With these efforts, final result is expected to be used as a powerful monitoring tool of nuclear reactor core

Rhodium SPND's Error Reduction using Extended Kalman Filter combined with Time Dependent Neutron Diffusion Equation

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Hun; Park, Tong Kyu; Jeon, Seong Su [FNC Technology Co., Ltd., Yongin (Korea, Republic of)

2014-05-15

The Rhodium SPND is accurate in steady-state conditions but responds slowly to changes in neutron flux. The slow response time of Rhodium SPND precludes its direct use for control and protection purposes specially when nuclear power plant is used for load following. To shorten the response time of Rhodium SPND, there were some acceleration methods but they could not reflect neutron flux distribution in reactor core. On the other hands, some methods for core power distribution monitoring could not consider the slow response time of Rhodium SPND and noise effect. In this paper, time dependent neutron diffusion equation is directly used to estimate reactor power distribution and extended Kalman filter method is used to correct neutron flux with Rhodium SPND's and to shorten the response time of them. Extended Kalman filter is effective tool to reduce measurement error of Rhodium SPND's and even simple FDM to solve time dependent neutron diffusion equation can be an effective measure. This method reduces random errors of detectors and can follow reactor power level without cross-section change. It means monitoring system may not calculate cross-section at every time steps and computing time will be shorten. To minimize delay of Rhodium SPND's conversion function h should be evaluated in next study. Neutron and Rh-103 reaction has several decay chains and half-lives over 40 seconds causing delay of detection. Time dependent neutron diffusion equation will be combined with decay chains. Power level and distribution change corresponding movement of control rod will be tested with more complicated reference code as well as xenon effect. With these efforts, final result is expected to be used as a powerful monitoring tool of nuclear reactor core.

Data Evaluation Problems in the Pulsed Neutron Source Method; Problemes d'Evaluation des Donnees dans les Applications de la Methode de la Source Pulsee; Problemy otsenki dannykh pri primenenii metoda istochnika impul'snykh nejtronov; Problemas de Evaluacion de Datos en el Metodo de la Fuente de Neutrones Pulsados

Energy Technology Data Exchange (ETDEWEB)

Pal, L.; Bod, L.; Szatmary, Z. [Central Research Institute for Physics, Hungarian Academy of Sciences, Budapest (Hungary)

1965-08-15

the weighted least-squares method using the decay constants of minimum standard deviation evaluated from the maximum likelihood function. Since, for the materials investigated, the values of the transport mean free path were not known, a fast converging iteration procedure was elaborated for the determination of the transport mean free path. The computer evaluation made it possible to determine the coefficients of the B{sup 6} and higher order terms of the decay constant. The diffusion parameters of the hydrogenous materials investigated were calculated by the Radkowsky prescription from the energy dependence of the integral scattering cross-sections. The calculations have shown that the Radkowsky prescription is a useful empirical method for the evaluation of the transport mean free path of numerous hydrogenous materials. (author) [French] Pour obtenir les parametres de diffusion des neutrons thermiques, on emploie depuis plus de dix ans la methode des neutrons puises. Cette methode convient tout particulierement pour l'etude des ralentisseurs hydrogenes. Les auteurs donnent les resultats des mesures pour l'eau, le benzene, le toluene, le xylene, le cyclohexane, le n-hexane et le diphenyle et discutent les methodes permettant d'evaluer avec le minimum d'erreurs les parametres de diffusion interessants, a partir des donnees mesurees. Ils etudient en detail au point de vue theorique et experimental l'effet du temps de vol qui se fait sentir lors des recherches sur les ralentisseurs a des temperatures elevees pour des raisons techniques, et influe sur la valeur de la constante de decroissance; ils indiquent les conditions geometriques dans lesquelles on peut obtenir une bonne precision. Ilsdeterminent en outre l'effet du temps mort du detecteur de neutrons et du systeme d'amplification sur la valeur de la constante de decroissance: l'effet du temps mort peut etre neglige si [i(t) {tau}]{sup 2} << 1, ou i (t) dt est le nombre des neutrons detectes dans l'intervalle de

Performance characteristics of specified power reactors in multidimensional neutron diffusion problems

International Nuclear Information System (INIS)

Kim, M.G.

1980-01-01

The multigroup neutron diffusion equations with the constraint of specified power distributions are investigated by the application of the straight-line method which can be considered as the limiting case of zero mesh space in the finite difference method. The standard partial differential form of the diffusion equation is reduced to sets of ordinary differential equations and then converted into sets of integral equations by using Green's functions defined on the pseudo straight lines. Coupling of each straight line to the adjacent lines arises from the application of a three-point central difference formula. The interfaces encountered between two regions are taken into account by imposing the continuity conditions for the grown fluxes and net currents with Taylor expansions of internal fluxes at the interface positions. A few sample problems are selected to test the validity of the method. It is found that the proposed method of solution is similar to the finite Fourier sine transform. Numerical results for the solutions obtained by the method of straight lines are compared with the results of the exact analytical solutions for simple geometries. These comparisons indicate that the proposed method is compatible with the analytical method, and in some problems considered the straight-line solutions are much more efficient than the exact solutions. The method is also extended to the reactor kinetics problem by expressing the kinetics parameters in terms of the basis functions which are used to obtain the solutions of the steady-state neutron diffusion equations

Fast diffusion in the intermetallics Ni3Sb and Fe3Si: a neutron scattering study

International Nuclear Information System (INIS)

Randl, O.G.

1994-02-01

We present the results of neutron scattering experiments designed to elucidate the reason for the extraordinarily fast majority component diffusion in two intermetallic alloys of DO 3 structure, Fe 3 Si and Ni 3 Sb: We have performed diffraction measurements in order to determine the crystal structure and the state of order of both alloys as a function of composition and temperature. The results on Fe 3 Si essentially confirm the classical phase diagram: The alloys of a composition between 16 and 25 at % Si are DO 3 -ordered at room temperature and disorder at high temperatures. The high-temperature phase Ni 3 Sb also crystallizes in the DO 3 structure. Vacancies are created in one Ni sublattice at Sb contents beyond 25 at %. In a second step the diffusion mechanism in Ni 3 Sb has been studied by means of quasielastic neutron scattering. The results are reconcileable with a very simple NN jump model between the two different Ni sublattices. Finally, the lattice dynamics of Fe 3 Si and Ni 3 Sb has been studied by inelastic neutron scattering in dependence of temperature (both alloys) and alloy composition (Fe 3 Si only). The results on Fe 3 Si indicate clearly that phonon enhancement is not the main reason for fast diffusion in this alloy. In Ni 3 Sb no typical signs of phonon-enhanced diffusion have been found either. As a conclusion, fast diffusion in DO 3 intermetallics is explained by extraordinarily high vacancy concentrations (several atomic percent) in the majority component sublattices. (author)

A solution of the thermal neutron diffusion equation for a two-region cyclindrical system program for ODRA-1305 computer

International Nuclear Information System (INIS)

Drozdowicz, K.; Woznicka, U.

1982-01-01

The program in FORTRAN for the ODRA-1305 computer is described. The dependence of the decay constant of the thermal neutron flux upon the dimensions of the two-region concentric cylindrical system is the result of the program. The solution (with a constant neutron flux in the inner medium assumed) is generally obtained in the one-group diffusion approximation by the method of the perturbation calculation. However, the energy distribution of the thermal neutron flux and the diffusion cooling are taken into account. The program is written for the case when the outer medium is hydrogenous. The listing of the program and an example of calculation results are included. (author)

Contribution to the study of the diffusion {alpha}-proton for {alpha} particles of 3,1 and 5,3 MeV; Contribution a l'etude de la diffusion {alpha}-proton pour des particules {alpha} d'energie comprise entre 3,1 et 5,3 MeV

Energy Technology Data Exchange (ETDEWEB)

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

1953-07-01

The diffusion of the particles has by the light cores that present a weak gate of potential, must permit the survey of the nuclear strengths. Some authors, studying the distribution in energy of the protons given out by a hydrogenated target submitted to a bombardment has variable energy, signal that this distribution has a structure of groups. We tried to reproduce experiences of diffusion {alpha}-proton, in order to verify the existence of the groups of signaled protons in the previous works. However in spite of finer experimental conditions, we had recovered any group structures in the distribution of the protons. This work permits to conclude that there is not a resonance in the {alpha}-proton diffusion for included energies between 3,1 and 5,3 MeV. The absence of resonances confirms the existence of the fundamental level of {sup 5}Li above in the neighborhood of 1,8 MeV {sup 4}He + {sup 1}H. (M.B.) [French] La diffusion des particules a par les noyaux legers qui presentent une faible barriere de potentiel, doit permettre l'etude des forces nucleaires. certains auteurs, etudiant la distribution en energie des protons emis par une cible hydrogenee soumise a un bombardement a d'energie variable, signalent que cette distribution a une structure de groupes. Nous avons essaye de reproduire les experiences du type diffusion {alpha}-proton, afin de verifier l'existence des groupes de protons signales dans les travaux anterieurs. Cependant malgre des conditions experimentales plus fines, nous n'avons retrouve aucunce structure de groupe dans la distribution des protons. Ce travail permet de conclure qu'il n'y a pas de resonance dans la diffusion {alpha}-proton pour des energies comprises entre 3,1 et 5,3 MeV. L'absence de resonances confirme l'existence du niveau fondamental de {sup 5}Li au voisinage de 1,8 MeV au-dessus de {sup 4}He + {sup 1}H. (M.B.)

Diffusion and plasticity at high temperature

Science.gov (United States)

Philibert, J.

1991-06-01

High temperature plastic deformation requires atomic migration whatever the mechanism of deformation. The constitutive equations contain a diffusion coefficient and the deformation rate follows an Arrhenius law. This paper will only discuss the case of viscous creep in order to elucidate the nature of the diffusion processes and the expression of the diffusion coefficient involved in alloys or compounds. La déformation plastique à haute température met en jeu des migrations atomiques, quel que soit le mécanisme de déformation. Les lois de comportement contiennent donc un coefficient de diffusion et la vitesse de déformation obéit à une loi d'Arrhenius. Dans cet article, qui ne conceme qu'un seul type de déformation, lefluage visqueux, on s'efforce de préciser la nature des processus de diffusion et du coefficient de diffusion mis en jeu dans le cas des alliages et des composés.

Study of the diffusion of iron, of silver and of carbon in beryllium using radioactive tracers; Etude de la diffusion du fer, de l'argent et du carbone dans le beryllium au moyen des traceurs radioactifs

Energy Technology Data Exchange (ETDEWEB)

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

1964-06-01

A study has been made of the diffusion of radioactive iron and silver tracers in beryllium. The following values have been found. D{sub Fe} = 0.53 exp - [51800 / RT], D{sub Ag} = 62 exp - [46100 / RT]. The values for iron are in good agreement with those found previously for chemical diffusion. A comparison of the diffusion coefficients for iron, silver and nickel shows that the diffusion rate increases with increasing solute atom radius. The existence has been shown of diffusion anisotropy for silver; it has been studied quantitatively on monocrystals. It is found that: D{sub parallel} 0.41 exp - [39100 / RT], D{sub perpendicular} = 1.98 exp - [45700 / RT]. The anisotropy decreases as the temperature increases. The silver diffuses more rapidly along the C axis than in the base plane. This result cannot be explained in terms of the model proposed for the diffusion of foreign atoms in solution in zinc. A greater number of experiments is required before a model can be put forward. An attempt has also been made to determine the diffusion coefficients of carbon in beryllium by treatment in an atmosphere of acetylene marked with C-14. Diffusion coefficients have been obtained but these should not be considered to be very significant since a chemical reaction occurs at the surface of the samples. (author) [French] On a etudie la diffusion dans le beryllium de traceurs radioactifs du fer et de l'argent. On trouve: D{sub Fe} = 0.53 exp - [51800 / RT], D{sub Ag} = 62 exp - [46100 / RT].Les valeurs trouvees pour le fer sont en bon accord avec les valeurs obtenues precedemment pour la diffusion chimique. La comparaison des coefficients de diffusion du fer, de l'argent et du nickel montre que la diffusion est d'autant plus rapide que le rayon atomique du solute est plus grand. On a mis en evidence une anisotropie de diffusion de l'argent qui a ete etudiee quantitativement sur des monocristaux. On trouve: D{sub parallele} = 0.41 exp - [39100 / RT], D{sub perpendiculaire} = 1

Solution of the multigroup neutron diffusion Eigenvalue problem in slab geometry by modified power method

Energy Technology Data Exchange (ETDEWEB)

Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática

2017-07-01

We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)

Other applications of neutron beams in material sciences; Autres utilisations des faisceaux de neutrons en science des materiaux

Energy Technology Data Exchange (ETDEWEB)

Novion, C.H. de

1997-12-31

The various applications of neutron beams are reviewed. The different mechanisms involved in neutron interaction with matter are explained. We notice that generally neutron radiation effects are unfavorable but can be turned into efficient tools to add new structures or properties to materials, silicon doping is an example. The basis principles of neutron activation analysis and neutron radiography are described. (A.C.)

Methods of Particle Detection in Free Neutron Decay; Methode de detection des particules dans une desintegration de neutrons libres; Metod obnaruzheniya chastits pri raspade svobodnogo nejtrona; Metodo para la deteccion de particulas en la desintegracion de neutrones libres

Energy Technology Data Exchange (ETDEWEB)

Novey, T B [Argonne National Laboratory, Lemont, IL (United States)

1960-06-15

A number of experimental studies have recently been completed by the Argonne Group on the decay of polarized neutrons in order to elucidate the structure of the weak nuclear interaction. These studies have taken the form of the measurement of the. angular distributions of electrons and protons with respect to the spin direction of the decaying free neutrons. The basic components of the apparatus which will be discussed are: 1. The one-meter iron-cobalt mirror which selects a beam of highly polarized neutrons and the methods for determination of the polarization. 2. The electron detector comprising a 10 cm diameter, 6 mm thick mosaic of anthracene crystals, and its light pipe system. 3. The proton detector, a 14-stage electron multiplier system, the first stage with a 15x15 cm opening, tapering in four stages to a standard 10-stage multiplier structure, and its entrance baffles for angular resolution. 4. The electronic system which selects pulses from the detectors having the proper time sequence, relative time-delay and pulse-height to allow indentification of a neutron decay. (author) [French] Le groupe Argonne a termine recemment des etudes experimentales sur la desintegration des neutrons polarises, en vue de mettre en lumiere le processus de l'interaction nucleaire faible. Ces etudes consistaient a mesurer les distributions angulaires d'electrons et de protons par rapport a la direction du spin des neutrons libres en voie de desintegration. Les principaux elements de l'appareillage qui sera decrit sont: 1. Le miroir de ferro-cobalt d'un metre, qui isole un faisceau de neutrons hautement polarises, ainsi que les methodes permettant de determiner la polarisation. 2. Le detecteur electronique qui comprend une mosaique de 10 cm de diametre et de 6 mm d'epaisseur en cristaux d'anthracene, et son systeme selectif. 3. Le detecteur de protons, un systeme multiplicateur electronique a 14 etages, le premier ayant une ouverture de 15x15 cm se retrecissant en 4 etages pour

Diffuse Scattering from Alloys and Disordered Systems: Experimental Techniques and Potentialities; Diffusion dans les Alliages et les Systemes Desordonnes; Methodes Experimentales et Possibilites d'Application; Diffuznoe rasseyanie na splavakh i razuporyadochennykh sistemakh. ehksperimental'nye metody i potentsial'nye vozmozhnosti; Dispersion Difusa en Aleaciones y Sistemas Desordenados; Tecnicas Experimentales y Posibilidades

Energy Technology Data Exchange (ETDEWEB)

Low, G. G. [Atomic Energy Research Establishment, Harwell, Berks (United Kingdom)

1965-04-15

field. Included among the latter are measurements of the diffuse magnetic scattering from paramagnetic salts. Observations of such scattering are of great interest since they contain direct information on the amount of spin transferred as a result of covalency, from a magnetic ion on to the ligands which surround it. (author) [French] Le memoire a pour objet d'examiner, du point de vue des principes et de la pratique, les divers renseignements que l'on peut obtenir en etudiant la diffusion elastique des neutrons. L'auteur traite surtout des systemes comportant une disposition stochastique des defauts, au sujet desquels on peut supposer qu'aucun n'est affecte par la presence des autres, soit en raison de la dilution, soit en raison d'une independance intrinseque 1 l'egard de l'environnement. Dans ces conditions, l'intensite diffusee est proportionnelle au carre du module d'une transformee de Fourier sur la perturbation de l'amplitude de diffusion associee a un defaut. On peut ainsi obtenir des renseignements sur la distribution spatiale de cette amplitude de diffusion en procedant aux mesures appropriees de diffusion des neutrons et en operant une inversion de Fourier. En pratique, vu le grand nombre de phenomenes qui contribuent au bruit de fond de diffusion, on n'a pu jusqu'a present faire des mesures que pour un petit nombre de champs. On a notamment realise des progres considerables en ce qui concerne les defauts magnetiques dans les ferro-aimants ou l'on peut faire varier la section efficace magnetique en modifiant la direction de magnetisation par rapport au vecteur de diffusion des neutrons. Cto parvient ainsi a separer la diffusion consideree des effets non magnetiques sous la forme d'une difference entre les deux mesures d'intensite. Un phenomene qui contribue au bruit de fond de diffusion et qui possede un caractere partiellement magnetique et par consequent n'est pas elimine par la technique susmentionnee, est la diffusion multiple de Bragg dans un polycristal

Performance modeling of parallel algorithms for solving neutron diffusion problems

International Nuclear Information System (INIS)

Azmy, Y.Y.; Kirk, B.L.

1995-01-01

Neutron diffusion calculations are the most common computational methods used in the design, analysis, and operation of nuclear reactors and related activities. Here, mathematical performance models are developed for the parallel algorithm used to solve the neutron diffusion equation on message passing and shared memory multiprocessors represented by the Intel iPSC/860 and the Sequent Balance 8000, respectively. The performance models are validated through several test problems, and these models are used to estimate the performance of each of the two considered architectures in situations typical of practical applications, such as fine meshes and a large number of participating processors. While message passing computers are capable of producing speedup, the parallel efficiency deteriorates rapidly as the number of processors increases. Furthermore, the speedup fails to improve appreciably for massively parallel computers so that only small- to medium-sized message passing multiprocessors offer a reasonable platform for this algorithm. In contrast, the performance model for the shared memory architecture predicts very high efficiency over a wide range of number of processors reasonable for this architecture. Furthermore, the model efficiency of the Sequent remains superior to that of the hypercube if its model parameters are adjusted to make its processors as fast as those of the iPSC/860. It is concluded that shared memory computers are better suited for this parallel algorithm than message passing computers

Diffusion weighted magnetic resonance imaging: ischemic and traumatic injury of the central nervous system; Diffusionsgewichtete MRI: ischaemische und traumatische Verletzungen des Zentralnervensystems

Energy Technology Data Exchange (ETDEWEB)

Huisman, T.A.G.M.; Sorensen, A.G. [Massachusetts General Hospital and Harvard Medical School, Boston (United States). MGH-NMR Center; Hawighorst, H.; Benoit, C.H. [Swiss Paraplegic Center Nottwil (Switzerland). Inst. of Radiology

2001-12-01

Routine hat sich die DWI bei der Diagnostik des akuten Schlaganfalls und des Traumas bewaehrt. durch die Moeglichkeiten zwischen Laesionen mit zytotoxischem Oedem (verminderte Diffusion) und Laesionen mit vasogenem Oedem (vermehrte Diffusion) zu unterscheiden. Cerebrale Verletzungen koennen so frueher nachgewiesen werden. Die Messung der Diffusion in verschiedenen Raumrichtungen erlaubt es eine Vielzahl funktionaler Karten zu erstellen. Die am haeufigsten verwendeten Karten sind die des apparent diffusion coefficients (ADC) und der isotropen Diffusion. Zusaetzlich koennen Karten ueber anisotrope Diffusion berechnet werden. Diese sollen Auskunft ueber die Integritaet und Lokalisation von Nervenbahnen geben. Diese funktional-anatomische Information wird wahrscheinlich in der Fuehdiagnostik von primaeren und sekundaeren Gewebeverletzungen unterschiedlicher Ursachen eine immer wichtigere Rolle spielen und koennte bestehende und zukuenftige neuroprotektive Behandlungen leiten und validieren. (orig.)

The Role of Diffusion Measurements in the Study of Crystal Lattice Defects; Role des Mesures de la Diffusion dans l'Etude des Defauts des Reseaux Cristallins; Rol' diffuzionnykh izmerenij v izuchenii defektov kristallicheskikh reshetok.rol' diffuzionnykh izmerenij v izuchenii defektov kristallicheskikh reshetok; Las Mediciones de la Difusion en el Estudio de Defectos de los Reticulados Cristalinos

Energy Technology Data Exchange (ETDEWEB)

Kidson, G. V. [Chalk River Nuclear Laboratories, ON (Canada)

1966-02-15

transport atomique dans les solides presentent souvent un interet direct pour les ingenieurs specialises dans l'etude, la mise au point et l'utilisation de materiaux. Actuellement on s'efforce de plus en plus d'aboutir a la comprehension de ces proprietes en fonction de l'apparition et la nature de defauts ponctuels et lineaires dans les cristaux. Le memoire fait le point de certaines etudes de diffusion effectuees recemment a Chalk River; ces travaux non seulement fournissent des resultats importants pour la technologie nucleaire, mais permettent egalement de mieux comprendre la nature des defauts de reseaux que l'on rencontre dans les materiaux. Les questions etudiees sont: a) l'autodiffusion dans la phase de zirconium pur a haute temperature; b) la diffusion d'un solute dans le plomb; c) l'interdiffusion de l'aluminium et du zirconium. L'auteur discute brievement les resultats assez surprenants et encore imparfaitement compris qui sont decrits sous a). Certaines indications donnent a penser que la diffusion se produit soit par un reseau serre de dislocation resultant d'une transformation de la phase martensitique, soit par un excedent de lacunes provoquees dans le cristal par des impuretes. Sous b), l'auteur discute la diffusion extraordinairement rapide des solutes de metaux nobles dans des monocristaux de plomb de grande purete selon l'etat de solution des atomes du solute. Il montre que Ton peut comprendre leur comportement de diffusion si l'on admet qu'une fraction f Inverted-Question-Mark des atomes dissous du solute occupe des sites interstitiels. Le coefficient de diffusion mesure D{sub m} et le coefficient de diffusion interstitielle sont lies par la relation D{sub m} = f{sub i}D{sub i}. Sous c), l'auteur decrit la formation et la croissance rapide d'un compose intermetallique unique ZrAl{sub 3} dans la zone de diffusion formee entre le zirconium et l'aluminium purs et il interprete le mecanisme de diffusion en fonction de la structure du reseau compose. Il

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-02-15

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

Natural equilibria in steady-state neutron diffusion with temperature feedback

International Nuclear Information System (INIS)

Pounders, J. M.; Ingram, R.

2013-01-01

The critical diffusion equation with feedback is investigated within the context of steady-state multiphysics. It is proposed that for critical configurations there is no need to include the multiplication factor k in the formulation of the diffusion equation. This is notable because exclusion of k from the coupled system of equations precludes the mathematically tenuous notion of a nonlinear eigenvalue problem. On the other hand, it is shown that if the factor k is retained in the diffusion equation, as is currently common practice, then the resulting problem is equivalent to the constrained minimization of a functional representing the critical equilibrium of neutron and temperature distributions. The unconstrained solution corresponding to k = 1 represents the natural equilibrium of a critical system at steady-state. Computational methods for solving the constrained problem (with k) are briefly reviewed from the literature and a method for the unconstrained problem (without k) is outlined. A numerical example is studied to examine the effects of the constraint in the nonlinear system. (authors)

Experimental Study of the Process of Neutron Thermalization in Time in Aqueous Moderators; Etude Experimentale du Processus de Thermalisation des Neutrons en Fonction du Temps dans les Ralentisseurs Contenant de l'Eau; Ehksperimental'noe izuchenie protsessa termalizatsii nejtronov vo vremeni v vodorodosoderzhashchikh zamedlitelyakh; Estudio Experimental del Proceso de Termalizacion de Neutrones en Funcion del Tiempo en Moderadores Acuosos

Energy Technology Data Exchange (ETDEWEB)

Ishmaev, S. N.; Mostovoj, V. I.; Sadikov, I. P.; Chernyshov, A. A. [Gosudarstvennyj Komitet po Ispol' zovaniju Atomnoj Ehnergii SSSR, Institut Atomnoj Ehnergii Im. I.V.Kurchatova, Moskva, SSSR (Russian Federation)

1965-08-15

Using a method described in another paper presented at this Symposium, the authors measured non-stationary neutron spectra in zirconium hydride and in water at room temperature as a function of slowing-down time, with a resolution of 3.5 {mu}s. The spectra measured in a large ZrH prism (B{sup 2} = 3.8 x 10{sup -2} cm{sup 2}) belong to a slowing-down time range (t > 30 {mu}s) in which the mean neutron energy is less than 0.13 eV and the energy exchange between neutron gas and the medium is caused mainly by the excitation of acoustic lattice vibrations. In this range thermalization proceeds at a comparatively slow rate and the time required to establish an equilibrium neutron spectrum is of the order of 400 {mu}s. In a small ZrH assembly (B{sup 2} Asymptotically-Equal-To 20 x 10{sup -2} cm{sup 2}) a strong diffusion-cooling effect is observed. The spectra measured at long slowing-down times are found to differ considerably from the Maxwellian distribution with the temperature of the medium. The time required to establish an equilibrium energy distribution increases with decreasing system size. The neutron energy spectra in water (B{sup 2} = 2.56 x 10{sup -2}cm{sup 2}) are presented for slowing-down times of over 13 {mu}s, where distortions from resolution are insignificant. The shape of the spectrum is only fully consistent with the Maxwellian distribution at room temperature after about 100 {mu}s following a fast neutron burst. The time required to establish an equilibrium neutron spectrum in this experiment is three to four times longer than the values calculated by various other authors. The possible reasons for this discrepancy are discussed. (author) [French] A l'aide de la methode que les auteurs ont decrite dans un autre memoire presente au Colloque, ils ont mesure, en fonction du temps de ralentissement, les spectres non stationnaires des neutrons dans l'hydrure de zirconium et dans l'eau a la temperature ambiante, avec un pouvoir de resolution de 3,5 {mu

Study by neutron diffusion of local order liquid sulfur around the polymerization transition

International Nuclear Information System (INIS)

Descotes, L.

1994-05-01

We studied the liquid sulfur according to the temperature. The sulfur is one of the most complicated elementary liquid. We experimented the neutron diffusion by the powder orthorhombic sulfur. The complexity at the polymerization transition are only accompanied by weak local structural transfer. 231 refs., 48 figs., 8 tabs., 3 annexes

Automation on computer of the partial area method in the analysis of resonances induced by 'S' neutrons 2. with an interference term and extension of the method to the treatment of multi resonances (1963); Automatisation sur ordinateur de la methode des aires partielles dans l'analyse des resonances induites par les neutrons ''S''. 2, avec terme d'interference et extension de la methode au traitement des multiresonances (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

This report deals with the numerical analysis on an I.B.M. 7090 computer of transmission resonances induced by 's' wave neutrons in time of flight experiments. The analysis method used is the partial area one. In this second part the interference term is taken into account. Modifications have been made in the programs and subroutines described in the first part, to determine the resonant transmissions from experimental raw data, and the relating partial areas. Also programs and subroutines are thoroughly described, which estimate the resonance parameters. The field of the partial area method has been extended to cover the case where several resonances have to be treated simultaneously, provided they do not interfere. (authors) [French] Le pretent rapport a pour objet l'analyse numerique sur ordinateur I.B.M. 7090 des resonances dues aux neutrons ''s'' dans les experiences de transmission par temps de vol, la methode d'analyse utilisee etant la methode dea aires partielles. Dans cette deuxieme partie il a ete tenu compte du terme d'interference. On y trouvera une description des amenagements apportes aux programmes et sous-programmes decrits dans la premiere partie pour determiner les transmissions interfero-resonnantes a partir des donnees experimentales brutes et les aires partielles afferentes. Sont egalement decrits les programmes et sous-programmes necessaires au calcul des parametres caracteristiques des resonances. Le domaine d'application de la methode a ete etendu au traitement simultane de plusieurs resonances groupees n'interferant pas entre elles. (auteurs)

Neutron Depolarization in Superconductors

Science.gov (United States)

Zhuchenko, N. K.

1995-04-01

The dependences of neutron depolarization on applied magnetic field are deduced along the magnetization hysteresis loop in terms of the Bean model of the critical state. The depolarization in uniaxial superconductors with the reversible magnetization, including uniaxial magnetic superconductors, is also considered. A strong depolarization is expected if the neutrons travel along the vortex lines. On calcule la dépendance en champ magnétique de la dépolarisation des neutrons le long du cycle d'hystérésis en termes du modèle critique de Bean. On considère aussi la dépolarisation dans les supraconducteurs uniaxiaux en fonction de l'aimantation réversible, y compris pour les supraconducteurs magnétiques. On attend une forte dépolarisation si les neutrons se propagent le long des vortex.

Fuel material neutron crystallography texture and structure determinations (1960); Etudes sur l'evolution des structures cristallines de l'uranium par neutrocristallographie (1960)

Energy Technology Data Exchange (ETDEWEB)

Laniesse, J; Englander, M; Meriel, P [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

{beta} {yields} {alpha} transformation at 333 deg. K of a bulk U alloy containing 1000 ppm chromium. (author) [French] La methode utilisee permet d'obtenir d'une maniere plus rapide et plus statistique que par difraction de rayons X des informations sur la texture d'agregats polycristallins d'uranium-metal et sur les cours de l'evolution de leurs structures crystallines pendant leurs transformations allotropiques. Elle emploie un faisceau neutronique sortant du reacteur EL3 au CEN Saclay, monochromatise a {lambda} = 1,143 A {+-} 0,030 Angstrom; l'axe de ce faisceau rigoureusement collimate, est dirige perpendiculairement a l'axe de fibre de l'echantillon de forme cylindrique et de dimensions optimales. Les neutrons diffuses sont recus dans un compteur BF3 qui peut decrire un cercle centre sur l'echantillon (R = 1,30 m), dans un plan horizontal passant par l'axe du faisceau; de plus, le dispositif de fixation de l'echantillon permet la rotation de celui-ci autour de son axe longitudinal. Le spectre de reference des raies de diffraction neutronique de la phase {alpha} de l'uranium pur est ainsi obtenu a la temperature ordinaire, en utilisant un echantillon pour lequel il est certain qu'il n'existe aucune texture parasite ou residuaire. Il en est de meme pour la phase {beta} d'une solution solide metastable d'uranium contenant la concentration nominale d'impuretes metalliques telles que le Cr ou le Si. Ces deux spectres sont compares aux spectres theoriques ainsi qu'aux spectres de diffraction roentgenographiques. En choisissant ensuite convenablement un certain nombre de ces raies de diffraction, les evaluations d'orientations predominantes eventuelles peuvent etre suivies qualitativement et semi-quantitativement, en fonction des differents traitements mecaniques et/ou thermiques subis simultanement ou non par l'echantillon. De meme en enregistrant d'une maniere continue l'intensite d'une raie mixte judicieusement placee par rapport aux raies des spectres {alpha} et {beta

Study by neutron diffusion of local order liquid sulfur around the polymerization transition; Etude par diffusion de neutrons de l`ordre local du soufre liquide autour de la transition de polymerisation

Energy Technology Data Exchange (ETDEWEB)

Descotes, L

1994-05-01

We studied the liquid sulfur according to the temperature. The sulfur is one of the most complicated elementary liquid. We experimented the neutron diffusion by the powder orthorhombic sulfur. The complexity at the polymerization transition are only accompanied by weak local structural transfer. 231 refs., 48 figs., 8 tabs., 3 annexes.

Development of the hierarchical domain decomposition boundary element method for solving the three-dimensional multiregion neutron diffusion equations

International Nuclear Information System (INIS)

Chiba, Gou; Tsuji, Masashi; Shimazu, Yoichiro

2001-01-01

A hierarchical domain decomposition boundary element method (HDD-BEM) that was developed to solve a two-dimensional neutron diffusion equation has been modified to deal with three-dimensional problems. In the HDD-BEM, the domain is decomposed into homogeneous regions. The boundary conditions on the common inner boundaries between decomposed regions and the neutron multiplication factor are initially assumed. With these assumptions, the neutron diffusion equations defined in decomposed homogeneous regions can be solved respectively by applying the boundary element method. This part corresponds to the 'lower level' calculations. At the 'higher level' calculations, the assumed values, the inner boundary conditions and the neutron multiplication factor, are modified so as to satisfy the continuity conditions for the neutron flux and the neutron currents on the inner boundaries. These procedures of the lower and higher levels are executed alternately and iteratively until the continuity conditions are satisfied within a convergence tolerance. With the hierarchical domain decomposition, it is possible to deal with problems composing a large number of regions, something that has been difficult with the conventional BEM. In this paper, it is showed that a three-dimensional problem even with 722 regions can be solved with a fine accuracy and an acceptable computation time. (author)

The determination by irradiation with a pulsed neutron generator and delayed neutron counting of the amount of fissile material present in a sample; Determination de la quantite de matiere fissile presente dans un echantillon par irradiation au moyen d'une source pulsee de neutrons et comptage des neutrons retardes

Energy Technology Data Exchange (ETDEWEB)

Beliard, L; Janot, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

A preliminary study was conducted to determine the amount of fissile material present in a sample. The method used consisted in irradiating the sample by means of a pulsed neutron generator and delayed neutron counting. Results show the validity of this method provided some experimental precautions are taken. Checking on the residual proportion of fissile material in leached hulls seems possible. (authors) [French] Ce rapport rend compte d'une etude preliminaire effectuee en vue de determiner la quantite de matiere fissile presente dans un echantillon. La methode utilisee consiste a irradier l'echantillon considere au moyen d'une source puisee de neutrons et a compter les neutrons retardes produits. Les resultats obtenus permettent de conclure a la validite de la methode moyennant certaines precautions. Un controle de la teneur residuelle en matiere fissile des gaines apres traitement semble possible. (auteurs)

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

Assembly Discontinuity Factors for the Neutron Diffusion Equation discretized with the Finite Volume Method. Application to BWR

International Nuclear Information System (INIS)

Bernal, A.; Roman, J.E.; Miró, R.; Verdú, G.

2016-01-01

Highlights: • A method is proposed to solve the eigenvalue problem of the Neutron Diffusion Equation in BWR. • The Neutron Diffusion Equation is discretized with the Finite Volume Method. • The currents are calculated by using a polynomial expansion of the neutron flux. • The current continuity and boundary conditions are defined implicitly to reduce the size of the matrices. • Different structured and unstructured meshes were used to discretize the BWR. - Abstract: The neutron flux spatial distribution in Boiling Water Reactors (BWRs) can be calculated by means of the Neutron Diffusion Equation (NDE), which is a space- and time-dependent differential equation. In steady state conditions, the time derivative terms are zero and this equation is rewritten as an eigenvalue problem. In addition, the spatial partial derivatives terms are transformed into algebraic terms by discretizing the geometry and using numerical methods. As regards the geometrical discretization, BWRs are complex systems containing different components of different geometries and materials, but they are usually modelled as parallelepiped nodes each one containing only one homogenized material to simplify the solution of the NDE. There are several techniques to correct the homogenization in the node, but the most commonly used in BWRs is that based on Assembly Discontinuity Factors (ADFs). As regards numerical methods, the Finite Volume Method (FVM) is feasible and suitable to be applied to the NDE. In this paper, a FVM based on a polynomial expansion method has been used to obtain the matrices of the eigenvalue problem, assuring the accomplishment of the ADFs for a BWR. This eigenvalue problem has been solved by means of the SLEPc library.

Diffusion in membranes: Toward a two-dimensional diffusion map

Directory of Open Access Journals (Sweden)

Toppozini Laura

2015-01-01

Full Text Available For decades, quasi-elastic neutron scattering has been the prime tool for studying molecular diffusion in membranes over relevant nanometer distances. These experiments are essential to our current understanding of molecular dynamics of lipids, proteins and membrane-active molecules. Recently, we presented experimental evidence from X-ray diffraction and quasi-elastic neutron scattering demonstrating that ethanol enhances the permeability of membranes. At the QENS 2014/WINS 2014 conference we presented a novel technique to measure diffusion across membranes employing 2-dimensional quasi-elastic neutron scattering. We present results from our preliminary analysis of an experiment on the cold neutron multi-chopper spectrometer LET at ISIS, where we studied the self-diffusion of water molecules along lipid membranes and have the possibility of studying the diffusion in membranes. By preparing highly oriented membrane stacks and aligning them horizontally in the spectrometer, our aim is to distinguish between lateral and transmembrane diffusion. Diffusion may also be measured at different locations in the membranes, such as the water layer and the hydrocarbon membrane core. With a complete analysis of the data, 2-dimensional mapping will enable us to determine diffusion channels of water and ethanol molecules to quantitatively determine nanoscale membrane permeability.

Macro-méthodologie et didactique des mathématiques

OpenAIRE

Fluckiger, Annick

2017-01-01

La didactique des mathématiques née, dans les années soixante, d’une réflexion menée sur les moyens d’améliorer l’enseignement des mathématiques, a eu l’ambition de se constituer en science (science des conditions spécifiques de la diffusion des connaissances mathématiques utiles au fonctionnement des institutions humaines selon Brousseau). Elle a alors été amenée à se doter de moyens d’études appropriés pour englober l’ensemble des recherches sur la diffusion des connaissances. La diversific...

Study of accelerated diffusion in gold and aluminium under neutron irradiation

International Nuclear Information System (INIS)

Acker, Denis.

1977-09-01

The speed-up of diffusion under neutron irradiation was studied. The experiments concern the self-diffusion of gold as a function of temperature and the heterodiffusion of copper and gold in aluminium against flux and temperature. In each of these systems the coefficients measured were 10 6 times higher than the expected extra-irradiation values for a flux of 6.10 12 n/cm 2 /s and at a temperature 0.33 Tsub(f), Tsub(f) being the matting point of the matrix expressed in Kelvins. The results obtained can be explained satisfactorily by assuming that, under irradiation: the activation energy of the diffusion coefficient is equal to half the hole migration energy (corrected for the hole-impurity interaction terms in the case of heterodiffusion); the diffusion coefficient under irradiation varies with the square root of the flux; defect wells eliminate interstitials much more efficient by than holes. The first two points agree well with theoretical predictions if the holes and interstitials are assumed to disappear essentially by mutual recombination, whereas the third can be interpreted in terms of a low efficiency of wells for holes and by supposing that the interstitial elimination reaction is limited only by the diffusion rate of these interstitials [fr

Non-destructive analysis of major components in plant materials by mean of 14-MeV neutrons; Analyse ''non destructrice'' des principaux constituants de la matiere vegetale apres irradiation aux neutrons de 14 MeV

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

Although reactors are an important source of neutrons for activation analysis, it is sometimes convenient to have recourse to other, less expensive neutron sources. The Centre d'Etudes Nucleaires at Grenoble has small electrostatic accelerators which generate fast neutrons of 14 MeV energy. These SAMES-type generators are used for directing a deuton beam at 150 kV onto a tritiated target, the resulting flux of monoenergetic neutrons attaining 5 x 10{sup 11} ns{sup -1} in 4 {pi} geometry by the {sup 3}H(d,n){sup 4}He reaction. Numerous elements found in plant material can be activated in this flux, mainly by (n,p), (n,{alpha}) and (n,2n) reactions. Current research is directed towards making use of the entire gamma spectrum of activated plant matter. A computer is used to break down the spectrum into seven main spectral regions by the least-squares method. As a first approximation, these regions are those obtained from aluminium, calcium, potassium, magnesium, phosphorus, silicon and chlorine standards in standard activation and radioactive decay conditions. The wanted advantage of this way of analysis is not acute sensibility, but great fastness. Therefore radioactivation with 14 MeV neutrons is particularly well adapted to quick and simultaneous dosages of useful elements in agronomy. (author) [French] Bien que les reacteurs constituent une source de neutrons importante en analyse par activation, il est parfois commode de recourir a d'autres sources neutronigenes exigeant un investissement moindre. Le Centre d'Etudes Nucleaires de Grenoble dispose de petits accelerateurs electrostatiques generateurs de neutrons rapides de 14 MeV. Ces generateurs, de type SAMES, accelerent un faisceau de deutons sous 150 kV vers une cible tritiee. Celle-ci fournit un flux de neutrons mono-energetiques atteignant 5 x 10{sup 11} n/s/4 {pi} par la reaction {sup 3}H(d,n){sup 4}He. De nombreux elements composant les matieres vegetales s'activent dans ce flux principalement par reactions

Measurement of the fission cross-section of {sup 235}U and {sup 239}Pu for thermal neutrons; Mesures des sections de fission de {sup 235}U et de {sup 239}Pu en neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Fraysse, G; Prosdocimi, A; Netter, F; Samour, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

Improved techniques of fast detection have been applied for determining the fission cross-sections of {sup 235}U and {sup 239}Pu with reference to the absorption cross-section of Boron. Monochromatic neutron beams of 0.0322 eV, 0.0626 eV and 0.275 eV have been employed. Use has been made of a Xe-filled gaseous scintillator and of a low-geometry solid state ion chamber. Both measured alpha and fission rates. The results at the reference energy of 0.0253 eV are: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (authors) [French] Des techniques avancees de comptage rapide ont ete mise en oeuvre pour determiner la section efficace de fission de {sup 235}U et de {sup 239}Pu par rapport a celle d'absorption du bore. Des faisceaux de neutrons monochromatiques de 0,0322 eV, 0,0626 eV et 0,275 eV ont ete employes. Les detecteurs utilises sont un scintillateur gazeux rempli de xenon et une chambre d'ionisation a etat solide a basse geometrie. Les deux ont mesure les taux des desintegrations alpha et des fissions. Les resultats a l'energie de reference de 0,0253 eV sont: ({sigma}{sub F}){sub 0} {sup 235}U = 588 {+-} 10 barns ({sigma}{sub F}){sub 0} {sup 239}Pu = 738 {+-} 7 barns. (auteurs)

NodHex3D: An application for solving the neutron diffusion equations in hexagonal-Z geometry and steady state

International Nuclear Information System (INIS)

Esquivel E, J.; Del Valle G, E.

2014-10-01

The system called NodHex3D is a graphical application that allows the solution of the neutron diffusion equation. The system considers fuel assemblies of hexagonal cross section. This application arose from the idea of expanding the development of neutron own codes, used primarily for academic purposes. The advantage associated with the use of NodHex3D, is that the kernel configuration and fuel batches is dynamically without affecting directly the base source code of the solution of the neutron diffusion equation. In addition to the kernel configuration to use, specify the values for the cross sections for each batch of fuel used, these values are: diffusion coefficient, removal cross section, absorption cross section, fission cross section and dispersion cross section. Important also, considering that the system is able to perform calculations for various energy groups. As evidence of the operation of NodHex3D, was proposed to model three-dimensional core of a nuclear reactor VVER-1000, based on the reference problem AER-FCM-101. The configuration of the reactor core consists of fuel assemblies (25 batches), composed of seven distinct materials, one of which reflector material, vacuum boundary conditions on the surface delimiting the reactor core. The diffusion equation for two energy groups solves, obtaining the value of the effective neutron multiplication factor. The obtained results are compared to those documented in the reference problem and by 3-DNT codes. (Author)

A diffuse neutron scattering study of clustering kinetics in Cu-Ni alloys

International Nuclear Information System (INIS)

Vrijen, J.; Radelaar, S.; Schwahn, D.

1977-01-01

Diffuse scattering of thermal neutrons was used to investigate the kinetics of clustering in Cu-Ni alloys. In order to optimize the experimental conditions the isotopes 65 Cu and 62 Ni were alloyed. The time evolution of the diffuse scattered intensity at 400 0 C has been measured for eight Cu-Ni alloys, varying in composition between 30 and 80 at. pour cent Ni. The relaxation of the so called null matrix, containing 56.5 at. pour cent Ni has also been investigated at 320, 340, 425 and 450 0 C. Using Cook's model from all these measurements information has been deduced about diffusion at low temperatures and about thermodynamic properties of the Cu-Ni system. It turns out that Cook's model is not sufficiently detailed for an accurate description of the initial stages of these relaxations

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1977-11-01

The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently

Fluctuations Magnetiques des Gaz D'electrons Bidimensionnels: Application AU Compose Supraconducteur LANTHANE(2-X) Strontium(x) Cuivre OXYGENE(4)

Science.gov (United States)

Benard, Pierre

Nous presentons une etude des fluctuations magnetiques de la phase normale de l'oxyde de cuivre supraconducteur La_{2-x}Sr _{x}CuO_4 . Le compose est modelise par le Hamiltonien de Hubbard bidimensionnel avec un terme de saut vers les deuxiemes voisins (modele tt'U). Le modele est etudie en utilisant l'approximation de la GRPA (Generalized Random Phase Approximation) et en incluant les effets de la renormalisation de l'interaction de Hubbard par les diagrammes de Brueckner-Kanamori. Dans l'approche presentee dans ce travail, les maximums du facteur de structure magnetique observes par les experiences de diffusion de neutrons sont associes aux anomalies 2k _{F} de reseau du facteur de structure des gaz d'electrons bidimensionnels sans interaction. Ces anomalies proviennent de la diffusion entre particules situees a des points de la surface de Fermi ou les vitesses de Fermi sont tangentes, et conduisent a des divergences dont la nature depend de la geometrie de la surface de Fermi au voisinage de ces points. Ces resultats sont ensuite appliques au modele tt'U, dont le modele de Hubbard usuel tU est un cas particulier. Dans la majorite des cas, les interactions ne determinent pas la position des maximums du facteur de structure. Le role de l'interaction est d'augmenter l'intensite des structures du facteur de structure magnetique associees a l'instabilite magnetique du systeme. Ces structures sont souvent deja presentes dans la partie imaginaire de la susceptibilite sans interaction. Le rapport d'intensite entre les maximums absolus et les autres structures du facteur de structure magnetique permet de determiner le rapport U_ {rn}/U_{c} qui mesure la proximite d'une instabilite magnetique. Le diagramme de phase est ensuite etudie afin de delimiter la plage de validite de l'approximation. Apres avoir discute des modes collectifs et de l'effet d'une partie imaginaire non-nulle de la self-energie, l'origine de l'echelle d'energie des fluctuations magnetiques est examinee

Methods for determining fast neutron spectra using threshold detectors; Les methodes de determination des spectres de neutrons rapides a l'aide de detecteurs a seuil

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

We propose to examine all the methods by which fast neutron spectra can be determined using the response of threshold detectors (activation or fission chamber detectors). Most of these methods have been proposed and often even used by various authors of which a list will be found in the bibliography. The aim of the present report is thus not to present original work but rather to gather into a single article and to present in a rational form a whole series of methods which have already been described in articles scattered throughout the specialised literature. Up to the present, each author has in general studied one or two methods and no comparative study of all the possible methods seems to have been made. The most comprehensive study on this topic is that of P.M. UTHE from whose article much has been borrowed. We have tried here to develop a useful tool which should facilitate the systematic experimental study leading to the recognition of the respective merits of the methods proposed. (author) [French] On se propose d'examiner l'ensemble des methodes permettant de determiner les spectres de neutrons rapides a partir des reponses de detecteurs a seuil (detecteurs par activation ou chambre a fission). La plupart de ces methodes ont deja ete proposees, et souvent meme utilisees, par differents auteurs dont on trouvera la liste en bibliographie. Le but du present rapport n'est donc pas de faire oeuvre originale mais plutot de rassembler dans un meme document et de presenter de maniere homogene toute une serie de methodes qui ont deja fait l'objet d'articles disperses dans la litterature specialisee. Jusqu'a present, chaque auteur s'est en general limite a l'etude experimentale d'une ou deux methodes et aucune etude comparative de l'ensemble des methodes possibles ne semble avoir ete faite. Le rapport le plus complet a ce sujet est celui de P.M. UTHE auquel de larges emprunts ont ete faits. On s'est efforce ici d'elaborer un outil de travail commode qui devrait

Determination of the ion thermal diffusivity from neutron emission profiles in decay

International Nuclear Information System (INIS)

Sasao, M.; Adams, J.M.; Conroy, S.; Jarvis, O.N.; Marcus, F.B.; Sadler, G.; Belle, P. van

1994-01-01

Spatial profiles of the neutron emission from deuterium plasmas are routinely obtained at the Joint European Torus (JET) using the line-integrated signals measured with a multichannel instrument. It is shown that the manner in which these profiles relax following the termination of strong heating with neutral beam injection (NBI) permits the local thermal diffusivity (χ i ) to be obtained with an accuracy of about 20%. (author)

Study of neutrons resonances as a function of spin for the nuclei I = 1/2 (silver, thulium, platinum) between 1 and 800 eV; Etude des resonnances de neutrons en foncton de leur spin pour les noyaux I = 1/2 (argent, thulium et platine) entre 1 et 800 eV

Energy Technology Data Exchange (ETDEWEB)

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

1966-04-01

The assignment of spins has been made for 120 levels of 239 detected. From these data we have deduced that: - The ratio of average spacing of each state of spin obeys the law of 2 J + 1. - The values of the strength function of {sup 195}Pt and {sup 169}Tm are in agreement with the model of Chase et Al and has the same values for the two states of spin. - The fluctuations of total radiation widths of resonances are observed in the case of {sup 195}Pt for the two spin-states. - The neutron widths are governed by the distributions of Porter and Thomas with one degree of freedom; the observed spacings are consistent with the Wigner distribution for two populations. (author) [French] L'attribution de spin a ete faite pour 120 niveaux sur 239 detectes. On en deduit: - Le rapport des espacements moyens pour chaque etat de spin qui obeit a la loi en (2j + 1). - Les valeurs des fonctions densites de {sup 195}Pt et {sup 169}Tm sont en accord avec le modele du noyau deforme de Chase et al. et ont la meme valeur pour les deux etats de spin. - Des fluctuations, dans les largeurs radiatives totales des resonances sont observees dans le cas de {sup 195}Pt pour les deux etats de spin. - Les largeurs de diffusion suivent les distributions de Porter et Thomas a un degre de liberte; les espacements celles de Wigner a deux populations. (auteur)

Influence des défauts de la structure du verre sur la résistance mécanique des fibres optiques

Science.gov (United States)

Chmel, A.; Baptizmanski, V. V.; Kharshak, A. A.

1992-12-01

For silica (pure and doped) optical fibers prepared from preforms irradiated by thermal neutrons and Ar^+ ions, the measurements of lifetime under transverse stress were carried out. It is found that the neutron bombardment leads to decrease of the lifetime and the strength of fiber while the ion implantation results in increasing of these parameters. The influence of particle irradiation on the glass strength is explained by the generation of structural defects which were observed by the infrared an Raman spectroscopy methods. The direction of the change of mechanical properties of fibers is determined by the type of defects and their distribution in the cross-section of fiber. Des préformes de fibres optiques de silice ont été exposées à un bombardement par des faisceaux de neutrons thermiques et d'ions Ar^+ à énergie de 40 keV afin d'engendrer des défauts de la structure respectivement dans le volume ou dans la couche superficielle des échantillons. Les fibres obtenues des préformes irradiées et non irradiées ont été mises en charge de flexion. On a déterminé leur temps de rupture sous diverses contraintes de traction sur la surface extérieure de la fibre en flexion. On a observé une diminution du temps de rupture des fibres après le bombardement neutronique et une augmentation après l'implantation ionique. L'analyse de la nature des défauts en faisant appel à la spectroscopie infrarouge et Raman a fait ressortir que l'irradiation aux ions entraînait essentiellement des liaisons chimiques dans une fine couche superficielle de l'échantillon alors que l'irradiation aux neutrons occasionnait une déformation des zones du réseau silicique dans toute la section de la préforme et de la fibre. L'amélioration de la résistance mécanique s'explique par une augmentation de la mobilité des éléments structuraux de la matrice de verre par suite d'une rupture partielle des liaisons chimiques et sa diminution par l'apparition de concentrateurs

POW3D-Neutron diffusion module of the AUS system. A user's manual

International Nuclear Information System (INIS)

Harrington, B.V.; Pollard, J.P.; Barry, J.M.

1996-11-01

POW3D is a three-dimensional neutron diffusion module of the AUS modular neutronics code system. It performs eigenvalue, source of feedback-free kinetics calculations. The module includes general criticality search options and extensive editing facilities including perturbation calculations. Output options include flux or reaction rate plot files. The code permits selection from one of a variety of different solution methods (MINI, ICCG or SLOR) for inner iterations with region re balance to enhance convergence. A MINI accelerated Gauss-Siedel method is used for upscatter iterations with group rebalance to enhance a convergence. Chebyshev source extrapolation is applied for outer iterations. A detailed index is included

The Measurement of Reactivity In Multiregion Subcritical Systems by the Pulsed Neutron Technique; Mesure de la Reactivite dans les Systemes Sous-Critiques a Plusieurs Regions par la Methode des Neutrons Pulses; Izmerenie reaktivnosti v mnogozonnykh podkriticheskikh sistemakh metodom impul'snykh nejtronov; Mediciones de la Reactividad en Sistemas Subcriticos de Varias Regiones Mediante la Tecnica de los Neutrones Pulsados

Energy Technology Data Exchange (ETDEWEB)

Sherwin, J.; Leng, J. H. [United Kingdom Atomic Energy Authority, Windscale Works, Cumberland (United Kingdom)

1965-10-15

The prompt decay of a thermalized burst of neutrons in a multiregion subcritical system is examined using a two-group diffusion theory. It is shown that a relationship can be established between the prompt decay constant of the fundamental mode and the effective multiplication factor of the system in terms of two parameters, defined, for the purposes of this paper, as the reactor response coefficient and the prompt decay correction and which depend to a large extent on the spatial flux patterns within the system. For the uniform bare system, the response coefficient can be identified with the mean neutron lifetime in the system; for a multiregion system it represents a compound of the lifetimes in each region weighted by perturbation type integrals. The second parameter, the decay correction, can have no physical meaning in that it arises from an attempt to relate the two scales of reactivity involved; that is the scale using the prompt decay constant and the scale derived using the effective multiplication factor. The properties of these parameters are examined here with reference to an enriched uranium graphitemoderated reactor consisting of uniform core and reflector and it is shown that the two parameters are not uniquely defined by the reactivity of the system but depend on the method chosen to make the system subcritical. Two sets of measurements are treated by the theory. In one, the Windscale Advanced Gas-Cooled Reactor was shut down by a uniform poison. In the other the reactivity of a core in the zero-energy reactor HERO was varied by altering the loaded radius. The neutron pulse measurements are all shown to be in good agreement with more conventional methods of reactivity determination. Some discussion of the experimental techniques and the difficulties encountered in the graphite-moderated systems is presented. (author) [French] Les auteurs etudient la decroissance des neutrons instantanes d'une bouffee thermalisee de neutrons dans un systeme sous

Parallel preconditioned conjugate gradient algorithm applied to neutron diffusion problem

International Nuclear Information System (INIS)

Majumdar, A.; Martin, W.R.

1992-01-01

Numerical solution of the neutron diffusion problem requires solving a linear system of equations such as Ax = b, where A is an n x n symmetric positive definite (SPD) matrix; x and b are vectors with n components. The preconditioned conjugate gradient (PCG) algorithm is an efficient iterative method for solving such a linear system of equations. In this paper, the authors describe the implementation of a parallel PCG algorithm on a shared memory machine (BBN TC2000) and on a distributed workstation (IBM RS6000) environment created by the parallel virtual machine parallelization software

Effets des rayonnements de haute energie sur le cristallin de la souris

CERN Document Server

Di Paola, M; Bianchi, M; Bianchi, M no 1; Baarli, J no 1; Di Paola, M no 1

1973-01-01

L'opacification du cristallin après irradiation par des neutrons et d'autres rayonnements ionisants a été largement étudiée, vu l'importance de telles recherches en radioprotection et la possibilité offerte par ce système d'analyser les effets produits par de petites doses de rayonnement. Les neutrons de haute énergie sont d'un intéret particulier pour la radioprotection près des accélérateurs et dans l'espace, mais jusqu'à présent les études ont été limitées aux neutrons d'énergie inférieur à 14 MeV. L'opacification du cristallin chez la souris, après irradiation par des neutrons de 400 MeV produits par le Synchro-Cyclotron du CERN, a été étudiée. Le valeurs d'E.B.R. ont été déterminées par comparaison avec des rayons X de 250 kV. Une breve discussion des résultats obtenus est inclue dans la présentation.

Analysis and development of spatial hp-refinement methods for solving the neutron transport equation; Analyse et developpement de methodes de raffinement hp en espace pour l'equation de transport des neutrons

Energy Technology Data Exchange (ETDEWEB)

Fournier, D.

2011-10-10

flux behaviour is very different depending on the energy, there is no reason to use the same spatial discretization. Such an approach implies to modify the initial estimators in order to take into account the coupling between groups. This study is done from a theoretical as well as from a numerical point of view. (author) [French] Pour la conception des coeurs de reacteurs de 4eme generation, une precision accrue est requise pour les calculs des differents parametres neutroniques. Les ressources memoire et le temps de calcul etant limites, une solution consiste a utiliser des methodes de raffinement de maillage afin de resoudre l'equation de transport des neutrons. Le flux neutronique, solution de cette equation, depend de l'energie, l'angle et l'espace. Les differentes variables sont discretisees de maniere successive. L'energie avec une approche multigroupe, considerant les differentes grandeurs constantes sur chaque groupe, l'angle par une methode de collocation, dite approximation SN. Apres discretisation energetique et angulaire, un systeme d'equations hyperboliques couplees ne dependant plus que de la variable d'espace doit etre resolu. Des elements finis discontinus sont alors utilises afin de permettre la mise en place de methodes de raffinement dite hp. La precision de la solution peut alors etre amelioree via un raffinement en espace (h-raffinement), consistant a subdiviser une cellule en sous-cellules, ou en ordre (p-raffinement) en augmentant l'ordre de la base de polynomes utilisee. Dans cette these, les proprietes de ces methodes sont analysees et montrent l'importance de la regularite de la solution dans le choix du type de raffinement. Ainsi deux estimateurs d'erreurs permettant de mener le raffinement ont ete utilises. Le premier, suppose des hypotheses de regularite tres fortes (solution analytique) alors que le second utilise seulement le fait que la solution est a variations bornees. La

The determination of sulphur in materials of high neutron absorption cross-section by fast-neutron activation analysis; Determination du soufre dans les matieres de forte section efficace d'absorption neutronique, au moyen d'une analyse par activation avec des neutrons rapides; Opredelenie sery v materialakh s bol'shim secheniem pogloshcheniya nejtronov metodom aktivatsionnogo analiza bystrykh nejtronov; Determinacion del azufre en sustancias de elevada seccion eficaz de absorcion neutronica mediante analisis por activacion con neutrones rapidos

Energy Technology Data Exchange (ETDEWEB)

Gibbons, D; Simpson, H [Wantage Research Laboratory (A.E.R.E.), Wantage, Berks. (United Kingdom)

1962-01-15

The production of radiophosphorus by the fast neutron (n, p) reaction on sulphur is used for the determination of sulphur by activation analysis. Using this technique it is possible to determine sulphur in materials of high thermal-neutron absorption cross-section because the absorption cross-sections for fast neutrons are, in general, much lower than those for slow neutrons, so that self-shielding errors are considerably reduced. Interference due to the slow neutron (n, {gamma}) reaction on neutral phosphorus is eliminated by a double irradiation technique involving irradiation in regions of differing slow/fast flux-ratios. The method has been applied to the determination of sulphur in chromium and arsenic. (author) [French] La production de radiophosphore par reaction (n, p) de neutrons rapides sur le soufre est utilisee pour la determination du soufre au moyen d'une analyse par activation. Cette methode permet la determination du soufre dans les matieres de forte section efficace d'absorption des neutrons thermiques; en effet, les sections efficaces d'absorption des neutrons rapides sont, d'une maniere generale, beaucoup plus faibles que pour les neutrons lents, de sorte que les erreurs d'autoprotection se trouvent considerablement reduites. L'interference due a la reaction (n, {gamma}) des neutrons lents sur le phosphore naturel est eliminee par une double irradiation, qui implique une irradiation dans les regions ou le rapport du flux de neutrons lents au flux de neutrons rapides a des valeurs differentes. La methode a ete appliquee a la determination du soufre dans le chrome et l'arsenic. (author) [Spanish] La formacion de fosforo radiactivo por la accion de los neutrones rapidos sobre el azufre, reaccion (n, p), puede servir para determinar este elemento mediante analisis por activacion. El empleo de esta tecnica permite efectuar determinaciones de azufre en materiales de elevada seccion eficaz de absorcion de neutrones termicos porque, en general, las

A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

International Nuclear Information System (INIS)

Ozgener, B.

1998-01-01

A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

High order backward discretization of the neutron diffusion equation

Energy Technology Data Exchange (ETDEWEB)

Ginestar, D.; Bru, R.; Marin, J. [Universidad Politecnica de Valencia (Spain). Departamento de Matematica Aplicada; Verdu, G.; Munoz-Cobo, J.L. [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear; Vidal, V. [Universidad Politecnica de Valencia (Spain). Departamento de Sistemas Informaticos y Computacion

1997-11-21

Fast codes capable of dealing with three-dimensional geometries, are needed to be able to simulate spatially complicated transients in a nuclear reactor. We propose a new discretization technique for the time integration of the neutron diffusion equation, based on the backward difference formulas for systems of stiff ordinary differential equations. This method needs to solve a system of linear equations for each integration step, and for this purpose, we have developed an iterative block algorithm combined with a variational acceleration technique. We tested the algorithm with two benchmark problems, and compared the results with those provided by other codes, concluding that the performance and overall agreement are very good. (author).

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)

Computational modeling for the angular reconstruction of monoenergetic neutron flux in non-multiplying slabs using synthetic diffusion approximation

International Nuclear Information System (INIS)

Mansur, Ralph S.; Barros, Ricardo C.

2011-01-01

We describe a method to determine the neutron scalar flux in a slab using monoenergetic diffusion model. To achieve this goal we used three ingredients in the computational code that we developed on the Scilab platform: a spectral nodal method that generates numerical solution for the one-speed slab-geometry fixed source diffusion problem with no spatial truncation errors; a spatial reconstruction scheme to yield detailed profile of the coarse-mesh solution; and an angular reconstruction scheme to yield approximately the neutron angular flux profile at a given location of the slab migrating in a given direction. Numerical results are given to illustrate the efficiency of the offered code. (author)

Determination of the ion thermal diffusivity from neutron emission profiles in decay

Energy Technology Data Exchange (ETDEWEB)

Sasao, M. (National Inst. for Fusion Science, Nagoya (Japan)); Adams, J.M. (AEA Industrial Technology, Harwell (United Kingdom)); Conroy, S.; Jarvis, O.N.; Marcus, F.B.; Sadler, G.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking)

1994-01-01

Spatial profiles of the neutron emission from deuterium plasmas are routinely obtained at the Joint European Torus (JET) using the line-integrated signals measured with a multichannel instrument. It is shown that the manner in which these profiles relax following the termination of strong heating with neutral beam injection (NBI) permits the local thermal diffusivity ([chi][sub i]) to be obtained with an accuracy of about 20%. (author).

Solution of the neutron diffusion equation at two groups of energy by method of triangular finite elements

International Nuclear Information System (INIS)

Correia Filho, A.

1981-04-01

The Neutron Diffusion Equation at two groups of energy is solved with the use of the Finite - Element Method with first order triangular elements. The program EFTDN (Triangular Finite Elements on Neutron Diffusion) was developed using the language FORTRAN IV. The discrete formulation of the Diffusion Equation is obtained with the application of the Galerkin's Method. In order to solve the eigenvalue - problem, the Method of the Power is applied and, with the purpose of the convergence of the results, Chebshev's polynomial expressions are applied. On the solution of the systems of equations Gauss' Method is applied, divided in two different parts: triangularization of the matrix of coeficients and retrosubstitution taking in account the sparsity of the system. Several test - problems are solved, among then two P.W.R. type reactors, the ZION-1 with 1300 MWe and the 2D-IAEA - Benchmark. Comparision of results with standard solutions show the validity of application of the EFM and precision of the results. (Author) [pt

Numeric algorithms for parallel processors computer architectures with applications to the few-groups neutron diffusion equations

International Nuclear Information System (INIS)

Zee, S.K.

1987-01-01

A numeric algorithm and an associated computer code were developed for the rapid solution of the finite-difference method representation of the few-group neutron-diffusion equations on parallel computers. Applications of the numeric algorithm on both SIMD (vector pipeline) and MIMD/SIMD (multi-CUP/vector pipeline) architectures were explored. The algorithm was successfully implemented in the two-group, 3-D neutron diffusion computer code named DIFPAR3D (DIFfusion PARallel 3-Dimension). Numerical-solution techniques used in the code include the Chebyshev polynomial acceleration technique in conjunction with the power method of outer iteration. For inner iterations, a parallel form of red-black (cyclic) line SOR with automated determination of group dependent relaxation factors and iteration numbers required to achieve specified inner iteration error tolerance is incorporated. The code employs a macroscopic depletion model with trace capability for selected fission products' transients and critical boron. In addition to this, moderator and fuel temperature feedback models are also incorporated into the DIFPAR3D code, for realistic simulation of power reactor cores. The physics models used were proven acceptable in separate benchmarking studies

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1977-11-01

The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.

Localizing by autoradiography at -195 deg radioactive areas in rats exposed to a high flux of thermal neutrons, importance of phosphorus 32 in consecutive internal irradiation; Localisation par autoradiographie a -195 deg des zones radioactives chez le rat expose a un haut flux de neutrons thermiques, importance du phosphore 32 dans l'irradiation interne consecutive

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

Rats weighing 25 g were exposed for 5 mn to a flux of 6.10{sup 12} thermal neutrons/cm{sup 2}/s. Anatomical autoradiography at -195 deg. C has enabled the radioactive organs to be easily localised, to follow in course of time the decrease of radioactivity, and from it to deduce the probable nature of the numerous emitters in question. In particular, the phosphorus 32 has thus appeared to be one of those responsible for internal irradiation, general, on the one hand, by activating cellular phosphorus, local, on the other, by activating bony phosphates. Owing to this, an accidental irradiation by neutrons might have consequences that are both somatic (elective irradiation of the bone marrow) and genetic (activation of nucleic acids). The gamma spectrometry has confirmed the nature of certain other emitters. (author) [French] Des rats de 25 g ont ete exposes pendant 5 mn a un flux de 6.10{sup 12} neutrons thermiques/cm{sup 2}/s. L'autoradiographie anatomique a -195 deg. C a permis de localiser facilement les organes radioactifs, de suivre dans le temps la decroissance de la radioactivite, et d'en deduire la nature probable des nombreux emetteurs en cause. En particulier, le phosphore 32 est ainsi apparu comme l'un des responsables de l'irradiation interne, d'une part generale par activation du phosphore cellulaire, d'autre part locale par activation des phosphates osseux. Une irradiations accidentelle par neutrons aurait, de ce fait, des consequences a la fois somatiques (irradiation elective de la moelle osseuse) et genetiques (activation des acides nucleiques). La spectrometrie gamma a confirme la nature de certains autres emetteurs. (auteur)

New apparatus for discriminating shapes - application to the study of neutron-proton elastic scattering at 14.6 MeV; Nouveau dispositif de discrimination de formes - Application a l'etude de la diffusion elastique neutron-proton a 14,6 MeV; Novyj diskriminator po forme - primenenie k izucheniyu uprugogo rasseyaniya nejtronproton s 14,6 MeV; Nuevo dispositivo para discriminacion de formas - Aplicacion al estudio de la dispersion elastica neutron-proton a 14,6 MeV

Energy Technology Data Exchange (ETDEWEB)

Crettez, J P; Cambou, F; Ambrosino, G [Laboratoire Maurice de Broglie, Paris (France)

1962-04-15

The mean life of scintillations in caesium iodide is dependent on the nature of the ionizing particle, the shortest life corresponding to the highest ionization density. This property is utilized for distinguishing different particles producing scintillations of similar amplitude. The apparatus described is a shape discriminator. It measures the time required by the scintillation to fall from its maximum to an adjustable fraction thereof. A time-amplitude converter provides a pulse the height of which is proportional to the time so measured. A comparison is then made between the shapes of the scintillations produced by alpha particles, protons and electrons, and the results obtained are shown. Application of this method to the measurement of recoil protons set in motion in a thin hydrogenated diffuser by a neutron flux is also described. Suppression of the pulses produced by gamma rays makes it possible to deduce from the spectra obtained the variation of the differential elastic scattering cross-section in terms of the angle, for d-t reaction neutrons. (author) [French] La vie moyenne des scintillations dans l'iodure de cesium varie suivant la nature de la particule ionisante. A la plus grande densite d'ionisation correspond la vie la plus courte. Cette propriete est employee pour distinguer differentes particules produisant des scintillations d'amplitude analogue. L'appareil decrit est un discriminateur de forme. Il mesure le temps mis par la scintillation pour passer de son maximum a une fraction reglable de cette amplitude. Un convertisseur temps-amplitude fournit une impulsion dont la hauteur est proportionnelle au temps ainsi mesure. Les auteurs comparent les formes des scintillations produites par les particules a, les protons, les electrons; ils presentent les resultats obtenus. Les auteurs decrivent l'application de cette methode a la mesure des protons de recul qu'un flux de neutrons met en mouvement dans un diffuseur hydrogene mince. La suppression des

The inelastic scattering of medium energy {alpha} particles; Sur la diffusion inelastique des particules {alpha} a moyenne energie

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

The aim of this work is to find out what are the properties of the so-called 'anomalous states' in medium weight nuclei. These states preferentially excited in the inelastic scattering of medium energy charged particles have an excitation energy at about 4 MeV for nuclei with Z {<=} 29 and in the range 2-3 MeV for high Z nuclei. From a combination of angular distribution data in the elastic and inelastic scattering of 30 MeV {alpha} particles, and correlation data between inelastic {alpha} particles and deexcitation {gamma} rays, we show that for even-even nuclei, we can attribute spin 3 and parity minus to these 'anomalous states'. This is quite in agreement with the interpretation of these levels suggested by Lane as due to collective octupole oscillations. We give a resume of the theories used in the analysis of the data and a description of the experimental set-up. (author) [French] Le but de cette etude est de determiner les proprietes des niveaux dits 'anormalement excites' lors de la diffusion inelastique des particules chargees de moyenne energie sur des noyaux de masse moyenne et lourde. L'energie de ces niveaux est de l'ordre de 4 MeV pour les noyaux avec Z {<=} 29 et de 2 a 3 MeV pour les noyaux de Z plus eleve. De l'examen des courbes de distribution angulaire des particules {alpha} de 30 MeV diffusees elastiquement et inelastiquement, et de la correlation angulaire entre {alpha} excitant ces niveaux 'anormaux' et {gamma} de desexcitation, on deduit que, dans le cas des pair-pair, on peut attribuer a ces niveaux spin 3 et parite moins. Ceci renforce l'hypothese emise par Lane qui attribue ces niveaux a des oscillations octupolaires de la surface du noyau. On donne un apercu des theories utilisees dans l'analyse des resultats et une description des dispositifs experimentaux. (auteur)

The diffusion coefficient as a function of energy for usual moderators; Le coefficient de diffusion en fonction de l'energie des thermaliseurs usuels

Energy Technology Data Exchange (ETDEWEB)

Cadilhac, M; Livolant, M; Pillard, D; Soule, J -L [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-07-01

After a review of the definition of the energy function to be taken for the diffusion coefficient used in calculations of reactor cells, in the elementary theory of diffusion, various possible approximations of this function are given. These approximations are shown to be accurate enough for practical uses. Methods for the direct calculations of these approximations for the currently used 'normal' models (methods avoiding the calculation of differential cross sections) are described; these methods are applied to several usual moderators at various temperatures according to different models. (authors) [French] On rappelle la definition de la fonction de l'energie que l'on doit prendre pour le coefficient de diffusion dans les calculs de cellule de reacteur en theorie elementaire de la diffusion. Puis on indique diverses approximations possibles de cette fonction. On montre que ces approximations ont en pratique une precision suffisante. Enfin on decrit des methodes de calcul direct de ces approximations pour les modeles 'normaux' couramment employes (methodes qui evitent d'avoir a calculer les sections efficaces differentielles), et on les applique a plusieurs thermaliseurs usuels a differentes temperatures (selon differents modeles). (auteurs)

The KASY synthesis programme for the approximative solution of the 3-dimensional neutron diffusion equation

International Nuclear Information System (INIS)

Buckel, G.; Wouters, R. de; Pilate, S.

1977-01-01

The synthesis code KASY for an approximate solution of the three-dimensional neutron diffusion equation is described; the state of the art as well as envisaged program extensions and the application to tasks from the field of reactor designing are dealt with. (RW) [de

Positron annihilation in solids: positronium diffusion; Annihilation du positon dans les solides diffusion du positonium

Energy Technology Data Exchange (ETDEWEB)

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

1969-04-01

The existence of two slow components in life-time spectrum of positron annihilation in silicium, aluminium and alkaline-earth oxides powders is established. These two long mean-lives {approx_equal} 10{sup -9} s and {approx_equal} 10{sup -7} s result from annihilation, inside and outside the grains respectively, of ortho-positronium formed in defects present in ionic crystals investigated. Dynamic behaviour of Ps, so revealed, is analyzed in terms of diffusion in excellent agreement with experiment. Diffusion constants of the order of 10{sup -4} cm{sup 2} sec{sup -1} and mean path before annihilation from 50 to 300 Angstrom are measured. From 100 to 500 K the temperature influence upon diffusion process is effective only in SiO{sub 2} where activation energy is found about 10{sup -2} eV. The p-Ps zero point energy evaluated by angular correlation gives the order of magnitude for defects dimensions and diffusion mean-time. Finally, o-Ps behaviour in space between grains, where its interaction with atmospheric gases can be only detected, is analysed. (author) [French] Nous mettons en evidence l'existence de deux composantes lentes dans le spectre de temps de vie du positon avant annihilation dans des poudres d'oxydes alcalinoterreux d'alumine et de silice. Ces deux longues vies moyennes {approx_equal} 10{sup -9} s et {approx_equal} 10{sup -7} s resultent respectivement de l'annihilation a l'interieur et a l'exterieur des grains de l'ortho-positonium forme dans certains defauts presents dans les cristaux ioniques etudies. L'analyse des proprietes dynamiques du Ps ainsi revelees, est effectuee en termes de diffusion en excellent accord avec l'experience. Des constantes de diffusion de l'ordre de 10{sup -4} cm{sup 2} sec{sup -1} et des parcours moyens avant annihilation variant de 50 a 300 Angstrom sont ainsi mesures. Entre 100 et 500 K l'influence de la temperature sur le processus de diffusion n'est sensible que dans SiO{sub 2} ou l'energie d'activation est trouvee

Étude des interactions hadroniques à deux corps chargés, induites sur des protons par un faisceau de $\\pi^{-}, K^{-}$, et $\\overline{p}$ entre 3 et 12 GeV/c, avec un angle de diffusion dans le centre de masse voisin de 90$^{o}$

CERN Document Server

Billoir, Pierre

1983-01-01

Étude des interactions hadroniques à deux corps chargés, induites sur des protons par un faisceau de $\\pi^{-}, K^{-}$, et $\\overline{p}$ entre 3 et 12 GeV/c, avec un angle de diffusion dans le centre de masse voisin de 90$^{o}$

Derivation of Inter-Atomic Force Constants of Cu2O from Diffuse Neutron Scattering Measurement

Directory of Open Access Journals (Sweden)

T. Makhsun

2013-04-01

Full Text Available Neutron scattering intensity from Cu2O compound has been measured at 10 K and 295 K with High Resolution Powder Diffractometer at JRR-3 JAEA. The oscillatory diffuse scattering related to correlations among thermal displacements of atoms was observed at 295 K. The correlation parameters were determined from the observed diffuse scattering intensity at 10 and 295 K. The force constants between the neighboring atoms in Cu2O were estimated from the correlation parameters and compared to those of Ag2O

Study of {gamma}'s in Naiade; Etude des gamma de Naiade

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

Following a study of the gamma sources, the flux of gamma of different energies in the swimming pool is investigated. The biological dose can thus be obtained by calculation, and compared with the results given by photographic plates. The influence of photoneutrons is estimated by calculation, and research is being carried out on their influence on the thermal neutron flux curve on the axis of the uranium plate, with the plate emitting neutrons and with the plate protected by boral. (author) [French] Apres l'etude des sources de gamma, l'on etudie le flux de gamma de differentes energies dans la piscine. La dose biologique peut etre obtenue ainsi par le calcul et comparee avec les resultats donnes par les plaques photographiques. L'influence des photoneutrons est estimee par le calcul et l'on recherche leur influence sur la courbe de flux de neutrons thermiques sur l'axe de la plaque d'uranium, la plaque emettant des neutrons et la plaque protegee par du boral. (auteur)

Study of water diffusion on single-supported bilayer lipid membranes by quasielastic neutron scattering

DEFF Research Database (Denmark)

Bai, M.; Miskowiec, A.; Hansen, F. Y.

2012-01-01

High-energy-resolution quasielastic neutron scattering has been used to elucidate the diffusion of water molecules in proximity to single bilayer lipid membranes supported on a silicon substrate. By varying sample temperature, level of hydration, and deuteration, we identify three different types...... of diffusive water motion: bulk-like, confined, and bound. The motion of bulk-like and confined water molecules is fast compared to those bound to the lipid head groups (7-10 H2O molecules per lipid), which move on the same nanosecond time scale as H atoms within the lipid molecules. Copyright (C) EPLA, 2012...

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

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

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

Diffusion of Hydrogen in the beta-Phase of Pd-H Studied by Small Energy Transfer Neutron Scattering

Energy Technology Data Exchange (ETDEWEB)

Nelin, G; Skoeld, K

1974-07-01

The diffusion of hydrogen in beta-PdH has been studied by quasielastic neutron scattering. It is shown that the diffusion occurs through jumps between adjacent octahedral interstitial sites. The observed integrated quasielastic intensities cannot be described by a simple Debye-Waller factor. The phase transition from the beta-phase to the alpha-phase has also been studied. No dramatic changes in the scattering patterns were observed. It is concluded that the diffusion mechanism is remarkably similar between the low concentration alpha-phase and the high concentration beta-phase

LES ZONES HUMIDES DE FONDS DE VALLEE ET LA REGULATION DES POLLUTIONS AZOTEES DIFFUSES

OpenAIRE

Clement , Jean-Christophe

2001-01-01

Depuis une vingtaine d'années, les zones humides de bas-fonds, véritables interfaces entre les bassins versants agricoles et les cours d'eau, sont connues pour être capables de réguler les pollutions azotées diffuses. Cependant, cette capacité est fonction de bon nombre de paramètres qui jouent sur l'efficacité réelle de ces zones tampons. De nombreuses tentatives d'aménagements se sont d'ailleurs soldées par des échecs en partie à raison de la méconnaissance de ces différents facteurs. En Br...

Neutron scattering for investigation into the connection between phonons and diffusion in metallic systems

International Nuclear Information System (INIS)

Herzig, C.

1995-01-01

For examining the connection between the diffusion systematics and the lattice dynamics of the body-centered cubic metals, the temperature dependence of the self-diffusion (radiotracer technique) and the phonon dispersion (neutron scattering) have been measured in selected systems. In continuation of previous studies, the goal of the examinations reported was to put the earlier developed phonon-related diffusion model on a broader experimental basis, in order to perform verifying analyses. The phonon dispersion of the group 5 metal Nb has been measured up to high temperatures. In contrast to the values measured for the group 4 (β-Zr) and group 6 (Cr) metals, the dispersion in Nb revealed an only very weak temperature dependence. The exceptional case of the bcc β-Tl has been examined by measuring the diffusion and the dispersion in the β-T 83 In 17 alloy. Significant deviations from the conditions in the bcc transition metals have been found. Self-diffusion has been measured for the first time in Ba and β-Sc. Their diffusion systematics correlate with electron configuration. The influence of the d-electron concentration on the diffusion systematics has been measured in Ti-Mo and Hf-Nb alloys, the results backing the predictions of the phonon-related diffusion model. (orig.) [de

Monte Carlo simulation of neutron transport phenomena

International Nuclear Information System (INIS)

Srinivasan, P.

2009-01-01

Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux

Contribution to the study of the desorption of fission gases formed in irradiated uranium oxide; Contribution a l'etude de la desorption des gaz de fission formes dans l'oxyde d'uranium irradie

Energy Technology Data Exchange (ETDEWEB)

Berry, J -L; Darras, R; Roger, B [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-09-01

The release of {sup 133}Xe from irradiated UO{sub 2} has been studied in the temperature range 1300 to 1900 deg C, using various monocrystalline or sintered samples. Up to 1600 deg C, this release is proportional to the square root of the time and thus occurs essentially by diffusion. The apparent diffusion constant D' decreases and the activation energy of the corresponding process increases as the integrated neutron flux received by the fuel increases. As the density of the sintered samples decreases however, the activation energy of the release also decreases, so that the difference between D' values for sintered samples of different densities decreases as the temperature rises. Finally, above 1600 deg C, the fission gas release phenomenon is governed by UO{sub 2} evaporation, and all the different types of oxide studied have similar behaviors, characterized by poor retention of these gases. (authors) [French] La desorption du xenon 133 forme dans le bioxyde d'uranium irradie a ete etudiee dans l'intervalle de 1300 a 1900 C, a l'aide de differents echantillons monocristallins ou frittes. Jusqu'a 1600 C, elle s'effectue proportionnellement a la racine carree du temps, donc essentiellement par diffusion. La pseudo-constante de desorption D' decroit et l'energie d'activation du processus correspondant croit lorsque le flux de neutrons integre recu par le combustible augmente. Cependant, lorsque la densite des frittes diminue, l'energie d'activation de desorption diminue egalement, de sorte que l'ecart entre les valeurs de D' relatives a des frittes de densites differentes se restreint lorsque, la temperature s'eleve. Finalement, au-dessus de 1600 C, l'evaporation de l'UO{sub 2} regit le phenomene de liberation des gaz de fission, et toutes les qualites d'oxyde etudiees presentent alors des comportements voisins a cet egard, caracterises par une mediocre retention de ces gaz. (auteurs)

Contribution to the study of diffusion in poly-phase system; Contribution a l'etude de la diffusion en systeme polyphase

Energy Technology Data Exchange (ETDEWEB)

Adda, Y; Philibert, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Institut de Recherches de la Siderurgie Francaise (IRSID), 78 - Saint-Germain-en-Laye (France)

1959-07-01

this latter result and of the anomalies in the coefficient of diffusion should probably take into account the accumulation of reticular defects which would appear after displacement of the phase boundaries. (author) [French] Apres diffusion chimique entre deux metaux, a des temperatures auxquelles le diagramme d'equilibre indique l'existence de plusieurs phases, on observe, sur une section perpendiculaire au front de diffusion, des bandes paralleles correspondant a ces diverses phases. On sait que, dans ce cas, la courbe concentration-penetration presente alors une discontinuite a chaque interface. Les concentrations a l'aplomb de ces discontinuites donnent les limites de solubilites respectives des phases en presence. Au cours de nos essais relatifs au couple uranium-zirconium, nous avons verifie que ces concentrations ne varient pas avec la duree de diffusion et que, donc, les conditions de l'equilibre thermodynamique sont bien remplies. On dispose alors d'une methode interessante pour etablir le diagramme d'equilibre a l'etat solide, methode que nous avons appliquee au systeme U-Zr. Les etudes cinetiques de la diffusion polyphasee sont encore assez rares, par suite des difficultes experimentales. Aussi diverses methodes sont-elles proposees pour evaluer le coefficient de diffusion a partir d'etudes purement micrographiques (mesure des epaisseurs des phases intermediaires). Nos resultats experimentaux montrent que les hypotheses sur lesquelles sont basees ces methodes sont tres rarement valables. En effet, nous avons etabli les courbes concentration-penetration dans les systemes U-Zr (entre 590 deg. C et 950 deg. C), et U-MO (entre 800 deg. C et 1050 deg. C). Celles-ci montrent souvent des courbures tres accentuees, qui traduisent des variations du coefficient de diffusion qui ne peuvent pas s'exprimer suivant des relations simples. Enfin, nous avons observe certaines anomalies au voisinage des interfaces entre phases adjacentes. En outre, nous avons etudie l

Study by neutron diffusion of magnetic fluctuations in iron in the curie temperature region

International Nuclear Information System (INIS)

Ericson-Galula, M.

1958-12-01

The critical diffusion of neutrons in iron is due to the magnetisation fluctuations which occur in ferromagnetic substances in the neighbourhood of the Curie temperature. The fluctuations can be described in correlation terms; a correlation function γ R vector (t) is defined, γ R vector (t) = 0 vector (0) S R vector (t)> mean value of the scalar product of a reference spin and a spin situated at a distance (R) from the first and considered at the instant t. In chapter I we recall the generalities on neutron diffusion cross-sections; a brief summary is given of the theory of VAN HOVE, who has shown that the magnetic diffusion cross section of neutrons is the Fourier transformation of the correlation function. In chapter Il we study the spatial dependence of the correlation function, assumed to be independent of time. It can then be characterised by two parameters K 1 and r 1 , by means of which the range and intensity of the correlations can be calculated respectively. After setting out the principle of the measurement of these parameters, we shall describe the experimental apparatus. The experimental values obtained are in good agreement with the calculations, and the agreement is better if it is supposed that the second and not the first neighbours of an iron atom are magnetically active, as proposed by Neel. In chapter III we study the evolution with time of the correlation function; this evolution is characterised by a parameter Λ depending on the temperature, which occurs in the diffusion equation obeyed by the magnetisation fluctuations: δM vector /δt = Λ ∇ 2 M vector . The principle of the measurement of Λ is given, after which the modifications carried out on the experimental apparatus mentioned in chapter II are described. The results obtained are then discussed and compared with the theoretical forecasts of De Gennes, mode by using the Heinsenberg model and a simple band model; our values in good agreement with those calculated in the Heisenberg

Rotor calculations for neutron spectroscopy; Calculs des rotors de spectrometres a neutrons

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-01

The determination of stress in a rotating disk plane of symmetry normal to the axis of rotation has been studied by a number of investigators. In a recent paper Reich gives an operating process for an analytical solution in an asymmetric rotating disk. In the report we give the calculation of finite difference stress solutions applicable to the two rotating disks. The equations are then programmed for the 360.75 computer by Fortran methods concerning the rotors of choppers. (author) [French] La determination des contraintes dans les disques symetriques, en rotation a ete etudiee par de nombreux auteurs. Dans un recent rapport, Reich donne une solution pour le calcul des disques asymetriques. Ce rapport concerne l'application du calcul des contraintes par differences finies aux deux types de rotors. Les equations ecrites en langage Fortran pour l'ordinateur 360.75 concerne les rotors de choppers. (auteur)

Study of the fluctuations in neutrons density in a homogeneous plutonium pile; Etude de la fluctuation de la population de neutrons sur une pile homogene au plutonium

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Clouet d' Orval, C; Deilgat, E; Labbe, J; Molbert, M; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1962-07-01

The variations in neutron density in the centre of a reactor core are subjected to a statistical fluctuation phenomenon. Because of the correlations existing between the neutrons, their distribution does not follow Poisson's law. It diverges from it by a quantity depending on various parameter such as the lifetime and the effective fraction of the slowed-down neutrons. Experiments have been carried out designed to demonstrate this divergence and also, more particularly, the correlation between unaffected and slowed-down neutrons; these have been carried out with the help of the reactor 'Proserpine', a critical homogeneous plutonium assembly. (author) [French] L'evolution de la population neutronique au sein d'un milieu multiplicateur est soumise a un phenomene de fluctuations statistiques. En raison des correlations qui existent entre les neutrons, leur distribution ne suit pas une loi de Poisson. Elle s'en ecarte d'une quantite qui depend de divers parametres tels que le temps de vie, la proportion effective des neutrons retardes. Des experiences, destinees a mettre en evidence cet ecart, et plus particulierement les correlations entre neutrons prompts et neutrons retardes, ont ete realisees a l'aide de la pile 'Proserpine', experience critique homogene au plutonium. (auteur)