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Sample records for diffusion des neutrons

  1. The magnetic diffusion of neutrons; La diffusion magnetique des neutrons

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

    1959-07-01

    The purpose of this report is to examine briefly the diffusion of neutrons by substances, particularly by crystals containing permanent atomic or ionic magnetic moments. In other words we shall deal with ferromagnetic, antiferromagnetic, ferrimagnetic or paramagnetic crystals, but first it is necessary to touch on nuclear diffusion of neutrons. We shall start with the interaction of the neutron with a single diffusion centre; the results will then be applied to the magnetic interactions of the neutron with the satellite electrons of the atom; finally we shall discuss the diffusion of neutrons by crystals. (author) [French] Le but de ce rapport est d'examiner, brievement, la diffusion des neutrons par les substances, et surtout, par des cristaux qui contiennent des moments magnetiques atomiques ou ioniques permanents. C'est-a-dire que nous nous interesserons aux cristaux ferromagnetiques, antiferromagnetiques, ferrimagnetiques ou paramagnetiques; il nous faut cependant rappeler d'abord la diffusion nucleaire des neutrons. Nous commencerons par l'interaction du neutron avec un seul centre diffuseur; puis les resultats seront appliques aux interactions magnetiques du neutron avec les electrons satellites de l'atome; enfin nous discuterons la diffusion des neutrons par les cristaux. (auteur)

  2. Theoretical study of the paramagnetic scattering of neutrons; Etude theorique de la diffusion paramagnetique des neutrons

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

    1962-02-15

    General paramagnetic scattering of neutrons is investigated in the situation where the orbital moment of the magnetic scattering ions is not quenched. The general relevant expression of the cross-section is given. The proceeding results are applied to rare earths and iron group ions. It is shown that if crystalline actions lift the free ion ground state degeneracy the neutron may induce transitions between the various levels, the distances of which are typically of the order of several hundred cm{sup -1}. The various scattering cross-sections (elastic and inelastic) are calculated for the rare earths in a cubic crystal field, for the holmium and erbium sesquioxide and for the anhydrous iron chloride (Cl{sub 2}Fe). These cross-sections are high enough to allow for an experimental detection, thus providing a direct determination of the levels distances through the measurement of the neutron wave-length shift. Moreover it is shown that, for neodymium, holmium, erbium, the total cross-section for neutrons of one angstrom wave length is rather insensitive to the crystal field effects. The results are then compared with the available experimental studies. The influence of the orbital moment on the angular dependence of the scattering for polarised ions is then investigated. The well-known formula: I = I{sub 0}sin{sup 2}{beta} is only an approximation, the validity of which is discussed. (author) [French] On envisage le cas general de la diffusion paramagnetique de neutrons. Le moment orbital des ions magnetiques ne peut etre considere comme bloque. On donne l'expression generale de la section efficace. Les resultats obtenus sont appliques au cas des terres rares et des ions du groupe du fer. On montre que, si les actions cristallines levent la degenerescence du niveau fondamental de l'ion libre, le neutron peut induire des transitions entre les divers sous-niveaux, dont la distance est ordinairement de l'ordre de quelques centaines de cm{sup -1}. La section efficace des

  3. Measurement of the diffusion length of thermal neutrons in the beryllium oxide; Mesure de la longueur de diffusion des neutrons thermiques dans l'oxyde de beryllium

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    Koechlin, J C; Martelly, J; Duggal, V P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    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{sup 3}, the mean density of the massif is 2,92 gr/cm{sup 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) [French] La longueur de diffusion des neutrons thermiques dans l'oxyde de beryllium a ete obtenue en etudiant la repartition spatiale des neutrons dans un massif parallelepipedique de cette matiere placee devant la colonne thermique de la Pile de Saclay. La densite moyenne de l'oxyde de beryllium (BeO) est de 2,95 gr/cm{sup 3}, la densite moyenne du massif de 2,92 gr/cm{sup 3}. La valeur de la longueur de diffusion, deduite des mesures effectuees est: L 32,7 {+-} 0,5 cm (ecart probable). Des remarques sont formulees quant a l'influence de la repartition spectrale du flux de neutrons utilise. (auteurs)

  4. Scattering of Neutrons by Liquid Bromine; Diffusion des neutrons par le brome liquide; R; Dispersion de neutrones por bromo liquido

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    Coote, G E; Haywood, B C [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    1963-01-15

    Neutrons of 0.037 eV and 0.0088 eV were scattered from a thin sample of liquid bromine at room temperature, and the energy distributions of the scattered neutrons measured at angles from 10{sup o} to 160{sup o} by the time-of-flight method. The angular distribution confirms the results of a previous study by Caglioti. Spectra were expressed in the form of the ''scattering law'' S({alpha},{beta}) of Egelstaff and analysed to derive the generalized frequency distribution p({beta}) of atomic motions in the liquid. The function p({beta}) has a broad smooth peak centred at {beta}{approx}0.3, and shows no evidence for discrete levels in the measured range 0<{beta}< 0.8: the general shape of the curve supports the 'quasi-crystalline' picture of the liquid. Difficulty in the absolute normalization of p({beta}) is discussed. (author) [French] Des neutrons de 0,037 eV et 0,0088 eV ont ete dissuses par un echantillon mince de brome liquide a la temperature ambiante; on a mesure, par la methode du temps de vol, les distributions d'energies des neutrons diffuses, pour des angles compris entre 10{sup o} et 160{sup o}. La distribution angulaire confirme les resultats d'une etude anterieure faite par Caglioti. Les spectres ont ete exprimes d'apres la ''loi de dispersion'' S({alpha},{beta}) d'Egelstaff et analyses pour determiner la distribution generalisee des frequences p({beta}) des mouvements atomiques dans le liquide. La fonction p({beta}) comporte un pic large et a faible pente dont le centre est a {beta} {approx} 0,3 et ne presente aucune indication de niveaux discrets dans la gamme mesuree 0 < {beta} < 0,8; la forme generale de la courbe confirme l'image ''quasi cristalline'' du liquide. Les auteurs examinent les difficultes qui s'opposent a la normalisation absolue de p({beta}). (author) [Spanish] Los autores han estudiado la dispersion de neutrones de 0,037 eV y 0,0088 eV en una muestra de bromo liquido de espesor reducido, a temperatura ambiente. La distribucion

  5. Study of the critical scattering of neutrons by iron; Etude de la diffusion critique des neutrons par le fer

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    Galula, M; Jacrot, B; Mangin, J P [Commissariat a l' Energie Atomique, Saclay (France)

    1959-07-01

    The critical scattering of very slow neutrons by iron near critical point is measured by time of flight techniques. The VAN HOVE formula is verified and the geometrical parameters K{sub 1} et r{sub 1} introduced in this theory are determined. (author) [French] On etudie la diffusion critique des neutrons tres lents par le fer dans la region du point de Curie par une methode de temps de vol. On verifie la formule de VAN HOVE et on determine les parametres geometriques K{sub 1} et r{sub 1} introduit par ce dernier. (auteur)

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

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    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)

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

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    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)

  8. 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

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    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

  9. 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

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    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

  10. Study of fast neutron scattering. The displacement cross-section (1962); Etude de la diffusion des neutrons rapides. Section efficace de deplacement (1962)

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

    1962-07-01

    We propose a method for calculating the biological efficiency of fast neutrons emitted by in-pile fission sources. This method justifies the empirical theory of Albert and Welton. In making simple assumptions concerning the cross-sections, we have supposed that the propagation can ben reduced to a mono-kinetic problem. A system of orthonormal functions is then set up making it possible to calculate the flux leaving a planar source. This method generalises the results obtained by Platzek to the case where the elastic cross-sections are not isotropic, and make it possible in particular to define a displacement cross-section: extension of the diffusion coefficient. This method can be generalised to the case of neutron diffraction as a function of time, and to the study of slowing-down. Numerical results are given in an appendix for the following: H{sub 2}O, D{sub 2}O, Fe, Be, Pb, CH, CH{sub 2}. These cross-sections have been verified experimentally in water and in graphite for neutrons of 2.5 and 14 MeV using a SAMES accelerator and a 2 MeV Van De Graaff. (author) [French] Nous proposons une methode permettant de calculer l'efficacite biologique des neutrons rapides issus des sources de fission dans la protection d'une pile. Cette methode justifie la theorie empirique d'Albert et Welton. En faisant des hypotheses simples sur les sections efficaces, nous avons suppose que la propagation pouvait etre ramenee a un probleme monocinetique. Nous construisons alors un systeme de fonctions orthonormales qui permet de calculer le flux issu d'une source plane. Cette methode generalise les resultats obtenus par Platzek au cas ou les sections efficaces elastiques ne sont pas isotropes et en particulier permet de definir une section efficace de deplacement: extension du coefficient de diffusion. Cette methode peut etre generalisee a la diffusion des neutrons en fonction du temps et a l'etude du ralentissement. Les resultats numeriques sont donnes en annexe pour les corps. H{sub 2

  11. Differential Neutron Scattering from Hydrogenous Moderators; Diffusion Differentielle des Neutrons par des Ralentisseurs Hydrogenes; Differentsial'noe rasseyanie nejtronov iz vodorodosoderzhashchikh zamedlitelej; Dispersion Diferencial de Neutrones en Moderadores Hidrogenados

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    Beyster, J. R.; Young, J. C.; Neill, J. M.; Mowry, W. R. [General Atomic Division of General Dynamics Corporation, John Jay Hopkins Laboratory for Pure and Applied Science, San Diego, CA (United States)

    1965-08-15

    experiments, which are sensitive mainly to P{sub 0} scattering. In particular one may test the P{sub 1} scattering kernel appropriate to a given molecular model. Third, the transport cross-section may be calculated directly from the experiments for use in multigroup reactor analysis. (author) [French] On mesure par les methodes du temps de vol les sections efficaces differentielles de diffusion simple (d{sigma}/d {Omega}) pour les ralentisseurs usuels. On utilise des neutrons thermiques produits par une source puisee intense (accelerateur lineaire de la General Atomic) avec un parcours de vol de 12 m a l'extremite duquel est place un echantillon mince du ralentisseur a l*etude. En outre, on a egalement prevu un petit parcours de vol terminal pour les neutrons diffuses, entre l'echantillon et plusieurs detecteurs de neutrons entierement absorbants. Cette methode permet de mesurer simultanement la distribution angulaire de diffusion pour plus de 50 energies des neutrons incidents. Les intensites sont-elevees, le bruit de fond est faible et bien defini et on peut proceder a des mesures rapides pour tous les angles de diffusion compris entre 10 et 155 degres. Les auteurs presentent des mesures de la section efficace differentielle de diffusion pour le vanadium, H{sub 2}O, D{sub 2}O et ZrH. On a etudie le vanadium pour controler le dispositif experimental. On a fait des mesures avec H{sub 2}O pour diverses epaisseurs et orientations des echantillons en vue de les comparer aux calculs fondes sur divers modeles de diffusion par l'hydrogene lie et aux resultats obtenus par Springer et Reinsch. Toutes les mesures experimentales ont ete corrigees pour tenir compte des effets importants de la diffusion multiple dans les echantillons. Pour l'instant, on n'a pas observe experimentalement la variation accusee que fait prevoir le modele de diffusion de Nelkin lorsque l'energie des neutrons est d'environ 0,06 eV; ceci est du au fait que l'hypothese du rotateur unique inhibe a 0,06 eV n'a pas

  12. Determination of the mean free path of the thermal neutrons transport by the measure of a complex diffusion length; Determination du libre parcours moyen de transport des neutrons thermiques par la mesure d'une longueur de diffusion complexe

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

    1955-07-01

    The further method is the outcome of a technique used in the study of neutrons in scattering and slowing-down environment. In this technique, we replace the constant sources used in the classic experiences by modulated sources with a variable frequency. The object of this article is to describe the extension of the method for the mean free path for transport of thermal neutrons and also to indicate the possible applications for other sizes, as the slowing length, or the absolute value of the cross-section of the boron. (M.B.) [French] La methode qui va etre decrite est l'aboutissement d'une technique utilisee dans l'etude des milieux ou diffusent et se ralentissent des neutrons. Dans cette technique, on remplace les sources constantes utilisees dans les experiences classiques par des sources modulees, a frequence variable. L'objet de cet article est de decrire l'extension de la methode a la mesure du libre parcours moyen de transport des neutrons thermiques et egalement d'indiquer les applications possibles a la mesure d'autres grandeurs, telles que la longueur de ralentissement, ou la valeur absolue de la section de capture du bore. (M.B.)

  13. 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}

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    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)

  14. Study by neutron diffusion of magnetic fluctuations in iron in the curie temperature region; Etude des fluctuations d'aimantation dans le fer au voisinage de la temperature de curie par diffusion des neutrons

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

    1958-12-15

    Heinsenberg model and a simple band model; our values in good agreement with those calculated in the Heisenberg model and exclude the band model used. Finally, in chapter IV we recall the chief results obtained. (author) [French] La diffusion critique des neutrons dans le fer est due aux fluctuations d'aimantation qui existent dans les substances ferromagnetiques au voisinage de la temperature de Curie. Les fluctuations peuvent se decrire en termes de correlations; on definit une fonction de correlation: {gamma}{sub R{sub vector}} (t), {gamma}{sub R{sub vector}} (t) , valeur moyenne du produit scalaire d'un spin de reference et d'un spin situe a la distance (R) du premier et considere a l'instant t. Dans le chapitre I nous rappelons les generalites sur les sections efficaces de diffusion des neutrons; nous donnons un bref resume de la theorie de VAN HOVE qui a montre que la section efficace de diffusion magnetique des neutrons est la transformee de FOURIER de la fonction de correlation. Dans le chapitre II nous etudions la dependance spatiale de la fonction de correlation, supposee independante du temps. On peut alors la caracteriser par deux parametres K{sub 1} et r{sub 1} qui permettent de calculer respectivement la portee et l'intensite des correlations. Apres avoir expose le principe de la mesure de ces parametres, nous decrirons l'appareillage experimental. Les valeurs experimentales obtenues sont en bon accord avec celles calculees, l'accord est meilleur si on suppose que ce sont les deuxiemes et non les premiers voisins d'un atome de fer qui sont magnetiquement actifs, ainsi que l'a propose NEEL. Dans le chapitre III nous etudions l'evolution dans le temps de la fonction de correlation; cette evolution est caracterisee par un parametre {lambda} dependant de la temperature, qui intervient dans l'equation de diffusion, a laquelle obeissent les fluctuations d'aimantation {delta}M{sub vector}/{delta}t = {lambda} {nabla

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

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    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.)

  16. Installation for Studying the Scattering of Cold Neutrons; Installation pour l'etude de la diffusion des neutrons thermiques; Ustanovka dlya izucheniya rasseyaniya kholodnykh nejtronov; Instalacion para estudiar la dispersion de neutrones frios

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    Golikov, V V; Shapiro, F L; Shkatula, A; Yanik, E A [Ob' edinennyj institut yadernykh issledovanij, Dubna, Union of Soviet Socialist Republics (Russian Federation)

    1963-01-15

    Using the pulsed fut reactor in the Joint Institute for Nuclear Research, an installation was set up to investigate the spectrometry of cold neutrons. The moderator, adjoining the reactor reflector, and the beryllium filter were at the temperature of liquid nitrogen. The scatterer, located at a distance of about 0. 6 m from the moderator, was irradiated by flashes of cold neutrons, whose duration was determined by the lifetime of the neutrons in the moderator and the dispersion of the times-of-flight over a distance of 0. 6 m. The frequency of the flashes was 8/s. The steep beryllium edge of the spectrum, lying around 200 {mu}s, was used for spectrometry in the quasi-elastic range. The energy of the scattered neutrons was determined by the time-of-flight over the distance between scatterer and detector, which was about 10-40 m. ZnS + 10{sub 2}{sup 10} O{sub 3} scintillation detectors with surfaces of 300 and 2 000 cm{sup 2} were used and the efficiency, for fast neutrons, was about 60%. (author) [French] A l'aide du reacteur a flux pulse de neutrons rapides de l'Institut unifie de recherches nucleaires, les auteurs ont mis au point une installation pour la spectrometrie des neutrons thermiques. Le ralentisseur contigu au reflecteur du reacteur et le filtre deberyllium sont maintenus a la temperature de l'azote liquide. Le diffuseur, place a environ 60 cm du ralentisseur, est irradie par une bouffee de neutrons thermiques dont la duree est determinee par la vie moyenne des neutrons dans le ralentisseur et par l'etalement des temps de vol necessaires pour parcourir la distance de 60 cm. La frequence des bouffees est de 8/s. Pour la spectrometrie dans la region quasi-elastique, on utilise la bande en bordure du spectre qui correspond au beryllium et s'etale sur environ 200 {mu}s. L'energie des neutrons diffuses est determinee par la duree du parcours entre le diffuseur et le detecteur. Les auteurs ont utilise des detecteurs a scintillation a ZnS + 10{sub 2}{sup 10} O

  17. 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

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    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)

  18. Study of the radiation around a high energy accelerator. Production and scattering of cascade neutrons; Etude du rayonnement autour d'un accelerateur de haute energie. Production et diffusion des neutrons de cascade

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-03-01

    The cascade induced in protective screens by a 3 GeV proton beam has been studied using activation detectors; the results have been compared with the cosmic neutron spectrum in the atmosphere. A study of the secondary neutron spectrum has made, it possible to obtain the distribution of the dose and to determine the maximum permissible fluxes expressed in terms of the energy, taking into account all the daughter products present. The dose calculated has been checked experimentally. The proportion of cascade neutrons has been studied using the idea of an imaginary source. The parameters which have to be introduced into the general equations to take into account scattering in the the air have been determined. (author) [French] La cascade induite dans les ecrans de protection par un faisceau de protons de 3 GeV a ete etudiee au moyen de detecteurs a activation et la comparaison a ete faite avec le spectre des neutrons cosmiques dans l'atmosphere. L'etude du spectre des neutrons secondaires a permis de preciser la distribution de la dose et de determiner les flux maximaux admissibles qui sont exprimes en fonction de l'energie, en tenant compte de l'ensemble des descendants presents. La dose calculee a ete verifiee experimentalement. La propagation des neutrons de cascade a ete etudiee en introduisant la notion de source fictive. Les parametres a introduire dans les equations generales pour rendre compte de la diffusion dans l'air ont ete determines. (auteur)

  19. Contribution to the study of the inelastic scattering of neutrons from a to 5 MeV (1961); Contribution a l'etude de la diffusion inelastique des neutrons de 1 a 5 MeV (1961)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    - The aim of this work is to see if this reaction occurs only by compound nucleus formation or involves some contribution of direct interaction. In the first case, the angular distribution of inelastic neutrons is symmetric about 90 degree. In the second case, this distribution must be asymmetric and must change slowly with energy of incident neutrons. The neutrons corresponding at the excitation of a given level of the residual nucleus are selected by their coincidence with the {gamma} rays of deexcitation of this level. From the results of our measurements on iron, iodine and bismuth and of other laboratories on different elements, we can conclude that generally, the inelastic scattering of neutrons of some MeV occurs only by compound nucleus. (author) [French] - Le but de ce travail est de determiner si cette reaction s'effectue uniquement par passage par un noyau compose ou fait intervenir un processus d'interaction directe. Dans le premier cas, la distribution angulaire des neutrons inelastiques est symetrique par rapport a 90 degree. Dans le deuxieme cas, cette distribution doit etre asymetrique et doit varier lentement avec l'energie des neutrons incidents. Les neutrons correspondant a l'excitation d'un niveau determine du residuel sont selectionnes par leur cofncidence avec les rayonnements {gamma} de desexcitation de ce niveau. D'apres les resultats de nos mesures sur le fer, l'iode et le bismuth et de celles des autres laboratoires sur differents elements, nous pouvons conclure que, en general, la diffusion inelastique des neutrons de quelques MeV s'effectue uniquement par noyau compose. (auteur)

  20. Contribution to the experimental study of the critical scattering of cold neutrons in iron; Contriiution a l'etude experimentale de la diffusion critique des neutrons froids par le fer

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-03-15

    The aim of the present work is a study of magnetic fluctuations which are produced in iron in the neighbourhood of the Curie temperature, by neutron scattering. We start by briefly recalling the theory of scattering of neutrons by magnetic substances and Landau's theory of second order phase transitions which enables one to derive the magnetic cross section near the Curie temperature. Following this is a description of the experimental apparatus after which we present the experimental results. The analysis of the results confirms the four-third law obeyed by the magnetic susceptibility near the Curie point, predicted by recent theories based on the Heisenberg model. However, the analysis reveals a non-zero relaxation time for the magnetic fluctuations at the Curie point, which is in disagreement with theoretical conclusions. (author) [French] L'objet du present travail est l'etude des fluctuations d'aimantation qui prennent naissance dans le fer au voisinage de sa temperature de Curie par la diffusion des neutrons. Nous commencons par rappeler brievement les generalites sur la diffusion des neutrons par les substances magnetiques et la theorie de Landau des transitions de phase du second ordre qui permet de deriver une expression de la section efficace magnetique pres de la temperature de Curie. Ensuite, apres la description du dispositif experimental, nous presentons les resultats experimentaux. L'analyse de ces resultats confirme les theories recentes suivant le modele d'Heisenberg en ce qui concerne la 'loi en 4/3' de la susceptibilite magnetique au voisinage du point de Curie; mais par ailleurs elle revele l'existence d'un temps de relaxation des fluctuations d'aimantation non nul en ce point, ce qui est en desaccord avec les previsions theoriques actuelles. (auteur)

  1. 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)

  2. Application des techniques de diffusion de la lumière des rayons X et des neutrons à l'étude des systèmes colloïdaux. Première partie : Présentation théorique des trois techniques Application of Light, X-Ray and Neutron Diffusion Techniques to the Study of Colloidal Systems. Part One : Theoretical Description of Three Techniques

    Directory of Open Access Journals (Sweden)

    Espinat D.

    2006-11-01

    Full Text Available Les techniques de diffusion, des rayons X, des neutrons et de la lumière, jouent un rôle très important pour la compréhension des milieux colloïdaux. Peu d'articles de la littérature s'attachent à présenter conjointement les trois méthodes. Nous avons, dans la première partie de cet article, détaillé les principes théoriques en insistent tout particulièrement sur les spécificités de chacune. Après les rappels concernant la diffusion par les systèmes dilués, nous nous sommes intéressés aux systèmes concentrés pour lesquels les entités diffusantes sont en interaction. Les milieux dispersés montrent souvent une certaine polydispersité que l'on cherche à mesurer; les techniques de diffusion permettent cette mesure. Nous terminerons cette revue par une description des appareillages utilisés. La deuxième partie de cet article concernera une large illustration des possibilités de ces méthodes d'analyse à l'étude des systèmes colloïdaux, sur la base de travaux effectués à I'IFP (Institut Français du Pétrole ou dans de nombreux laboratoires de recherche extérieurs. This article aims to describe X-ray, neutron and light scattering techniques with emphasis on their specific nature and their scope of application. Indeed, whereas light diffusion has been used for a long time in research laboratories, in particular for characterizing polymers in solution, small angle X-ray scattering has been the subject of renewed interest in recent years. Neutron scattering, which is obviously more difficultly accessible, has proven to be extremely useful for studying various systems for which light and X-ray scattering remain relatively powerless. Whereas there is an abundant literature concerning various applications of the three methods, it should be noted that only a few articles have attempted to describe all three techniques at the same time. In this article we have tried to make up for this lack, and as such it was

  3. 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

  4. 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)

  5. Diffusion Parameters of BeO by the Pulsed Neutron Method; Calcul des Parametres de Diffusion de BeO par la Methode des Neutrons Pulses; Opredelenie diffuzionnykh parametrov BeO s pomoshch'yu metoda impul'snykh nejtronov; Determinacion de los Parametros de Difusion en el BeO por el Metodo de los Neutrones Pulsados

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-08-15

    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: {lambda}{sub 0} = absorption constant = 156.02 {+-} 4.37 s{sup -1} D = diffusion coefficient = (1.3334 {+-} 0.0128) x 10{sup 5} cm{sup 2}/s C = diffusion cooling constant = (-4.8758 {+-} 0.5846) x 10{sup 5} cm{sup 4}/s. The effect of neglecting the contribution of the B{sup 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) [French] Les auteurs decrivent l'emploi de la methode des neutrons puises pour la determination precise des parametres de diffusion des ralentisseurs. Ils ont calculi ceux de la glucine par cette methode. Des deuterons puises au moyen d'un accelerateur a cascade produisaient les bouffees de neutrons par la reaction Be (d, n). Le detecteur etait constitue par un compteur proportionnel au BF{sub 3} enrichi. En choisissant convenablement

  6. Applications des techniques de diffusion de la lumière, des rayons X et des neutrons a l'étude des systèmes colloïdaux. Troisième partie : caractérisation des microémulsions et des composés solides. Étude de différents systèmes d'intérêt industriel Application of Light, X-Ray and Neutron Diffusion Techniques to the Study of Colloidal Systems. Part Three: Microemulsions, Solid Materials, Liquids and Miscellaneous Systems with Industrial Uses

    Directory of Open Access Journals (Sweden)

    Espinat D.

    2006-11-01

    Full Text Available Cette troisième partie clôt une revue bibliographique concernant la caractérisation des systèmes colloïdaux par les techniques de diffusion (lumière, rayons X, neutrons (Rev. Inst. Franç. du Pétrole, Vol. 45, No 6, novembre-décembre 1990. Nous poursuivrons cette présentation en nous intéressant aux microémulsions. Après la présentation des informations que l'on peut obtenir sur les systèmes polymériques, les solutions micellaires ou les systèmes fractals, qui ont fait l'objet de multiples travaux, nous reviendrons sur les applications plus restreintes à des systèmes colloïdaux parfois moins bien définis et d'origine industrielle, telles les fractions lourdes des pétroles, les asphaltènes ou encore les additifs de lubrification. Les possibilités de la diffusion des rayons X et des neutrons à l'étude de matériaux solides seront présentées à travers la caractérisation des solides poreux et en particulier des catalyseurs. Nous terminerons cette présentation par les liquides purs ou les mélanges; les techniques de diffusion ont contribué fortement à la description des structures des phases liquides. This article is the second one (the first one was published in Revue de l'Institut Français du Pétrole No. 6, November-December 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 article takes up polymer and colloidal solutions. Particular attention is paid to the importance of scattering techniques for characterizing polymers in solution and micellar solutions. A

  7. 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

  8. Neutron cooling and cold-neutron sources (1962); Refroidissement des neutrons et sources de neutrons froids (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-01

    Intense cold-neutron sources are useful in studying solids by the inelastic scattering of neutrons. The paper presents a general survey covering the following aspects: a) theoretical considerations put forward by various authors regarding thermalization processes at very low temperatures; b) the experiments that have been carried out in numerous laboratories with a view to comparing the different moderators that can be used; c) the cold neutron sources that have actually been produced in reactors up to the present time, and the results obtained with them. (author) [French] Des sources intenses de neutrons froids sont utiles pour l'etude des solides par diffusion inelastique des neutrons. On presente une revue d'ensemble: a) des considerations theoriques faites par divers auteurs sur les processus de thermalisation a tres basse temperature; b) des experiences faites dans de nombreux laboratoires pour comparer les divers moderateurs possibles; c) des sources de neutrons froids effectivement realisees dans des piles a ce jour, et des resultats obtenus avec ces sources. (auteur)

  9. 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)

  10. 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.)

  11. 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.

  12. 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)

  13. Structural and dynamical study about denatured states of yeast phosphoglycerate kinase by neutrons scattering and X-rays; Etude structurale et dynamique des etats denatures de la phosphoglycerate kinase de levure par diffusion des neutrons et des rayons X

    Energy Technology Data Exchange (ETDEWEB)

    Receveur, V

    1997-04-28

    During a long time, the neutron scattering and X-rays techniques have not been used for the studies bearing on the folding of proteins. The compactness and the globularness of a protein are two structural characteristics describing the denatured states and the intermediate states of folding, and the neutrons and x-rays scattering are probably the two techniques the most appropriate to give this kind of information; they are sensible to the spatial extent and to the molecules compactness, and to their general shape. For these three or four last years, the works using these techniques are increasing, giving precious knowledge on the different steps of folding and on the interactions stabilizing the denatured or intermediate states. This thesis falls into this category. (N.C.).

  14. 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

  15. Calculation of the Inelastic Scattering of Neutrons from Polyethylene and Water; Calcul de la diffusion inelastique des neutrons par le polyethylene et l'eau; Raschet neuprugogo rasseyaniya nejtronov poliehtilenom i vodoj; Calculo de la dispersion inelastica de neutrones por polietileno y agua

    Energy Technology Data Exchange (ETDEWEB)

    Goldman, D T; Federighi, F D [Knolls Atomic Power Laboratory, General Electric Company, Schenectady, NY (United States)

    1963-01-15

    A model for the calculation of the scattering of thermal neutrons from chemical system was proposed by Nelkin. This model considered the actual dynamics of the scattering system as composed of a set of oscillatory motions, each describable by a Hamiltonian which commuted with each of the others. It was then possible to express the differential scattering cross-section in closed form. This model has been used to calculate the scattering of neutrons by water. Some care must be taken in performing the numerical integration over angle and energy. The scattering model has been extended to the calculation of neutron scattering from polyethylene C{sub n}H{sub 2n}. Analogous levels of polyethylene can be noted at 0.089 eV, 0.182 eV, 0.354 eV, and 0.533 eV. The differential and total cross-sections have been calculated for the scattering and the latter has been seen to be in reasonable agreement with experiment at room temperature. Scattering kernels have been calculated for a number of temperatures and where possible the results have been compared with experiment. In addition, neutron flux spectra and diffusion lengths have been calculated using the equations of reactor physics. Comparison of these Results with experimental data indicates that such integral measurements are indicative of at least the gross features of the scattering system and should be analysed in conduction with the detailed differential cross-section results. (author) [French] Nelkin a propose un modele pour calculer la diffusion de neutrons thermiques dans des systemes chimiques. Dans ce mod and le on considere que la dynamique reelle du systeme de diffusion se compose d'un ensemble de mouvements oscillatoires, chaque mouvement pouvant 6tre decrit par un hamiltonien commutant avec chacun des autres. Il est alors possible d'exprimer la section efficace differentielle de diffusion sous une forme fermee. Les auteurs ont employe ce modele pour calculer la diffusion des neutrons par l'eau. Il faut prendre

  16. 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

  17. Improvements in fast-neutron spectroscopy methods (1961); Amelioration des methodes de spectrometrie des neutrons rapides (1961)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-02-15

    This research aimed at improving fast-neutron electronic detectors based on n-p elastic scattering. The first part concerns proportional counters; careful constructional methods have made it possible to plot mono-energetic neutron spectra in the range 700 keV - 3 MeV with a resolution of 7 per cent. The second part concerns scintillation counters: an organic scintillator and an inorganic scintillator covered with a thin layer of a scattering agent. An exact study of the types of scintillation has made it possible to develop efficient discriminator circuits. Different neutron spectra plotted in the presence of a strong gamma background are presented. The last part deals with the development of form discrimination methods for the study, in the actual beam, of the elastic scattering of 14.58 MeV electrons. With hydrogen, the distribution f ({phi}) of the recoil protons is f({phi}) = 1 + 0.034 cos {phi} + 0.042 cos{sup 2} {phi}. With tritium the scattering is strongly anisotropic; the curve representing the variation of the differential cross-section for the elastic scattering in the centre of mass system is obtained with a target containing 1 cm{sup 3} of tritium. (author) [French] Le travail a porte sur l'amelioration des detecteurs electroniques de neutrons rapides bases sur la diffusion elastique n-p. La premiere partie est relative aux compteurs proportionnels; des methodes soignees de fabrication ont permis des traces de spectres de neutrons monoenergetiques dans le domaine 700 keV - 3 MeV avec une resolution de 7 pour cent. La deuxieme partie est relative au compteur a scintillations; scintillateur organique et scintillateur mineral recouvert d'un diffuseur mince. Une etude precise des formes de scintillations a permis la mise au point de circuits discriminateurs efficaces. Differents spectres de neutrons traces en presence d'un fond gamma intense sont presentes. La derniere partie est relative a la mise en oeuvre des methodes de discrimination de forme pour l

  18. 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)

  19. 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...

  20. 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)

  1. 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

  2. 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)

  3. Neutron Dosimetry and Irradiation of Solids; Dosimetrie des neutrons et irradiation des solides

    Energy Technology Data Exchange (ETDEWEB)

    Perriot, G; Schmitt, A P [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1962-07-01

    Results of work at C.E.A. from 1958 to 1960 are reviewed. The possibilities offered by classical dosimetry methods are discussed. The tests which led to the utilization, for fast neutron dosimetry, of resistivity variations induced in solid W by such neutrons are described. Experimental W irradiation results led to a definition of neutron efficiency which describes the relations between neutron energy and their effects on materials. Possibilities offered by detectors which make use of radiation damage and are sensitive to neutrons at keV energies were explored. In other work, the principal French reactors were classified according to their ability to produce damage in materials such as W. (authors) [French] Dans ce rapport on a presente les resultats essentiels de travaux qui ont ete effectues de 1958 a 1980 par des chercheurs du CEA issus de differents services. En meme temps qu'une revue des possibilites offertes a l'epoque par les methodes classiques de dosimetrie (utilisation des detecteurs par activation), on a decrit les essais qui devaient permettre d'utiliser, a la dosimetrie les neutrons rapides, les variations de resistivite qu'ils creent dans un corps solide (tungstene). L'irradiation du tungstene a montre l'importance qu'il y avait a definir 'l'efficacite' des neutrons, c'est-a-dire leur aptitude plus ou moins grande, selon leur energie, a creer des defauts dans les materiaux. L'efficacite d'un emplacement d'irradiation se trouvant liee au spectre neutronique, on a vu les difficultes qu'il y avait a utiliser les detecteurs par activation des qu'on n'avait plus affaire a un spectre en 1/E ou de fission et on a pu entrevoir les possibilites offertes par les detecteurs utilisant la creation des defauts qui repondent a tous les neutrons d'energies, superieures a quelques keV. Enfin, on a classe les principaux types de Piles Francaises selon leur aptitude a creer plus ou moins rapidement des dommages dans des materiaux comme le tungstene. (auteur)

  4. Study of the influence of chemical binding on resonant absorption and scattering of neutrons; Etude de l'influence des liaisons chimiques sur l'absorption et la diffusion des neutrons aux energies de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Naberejnev, D.G. [Aix-Marseille-1 Univ., 13 - Marseille (France)

    1999-02-01

    At present time the problem of taking into account of the crystalline binding in the heavy nuclei resonance range is not correctly treated in nuclear data processing codes. The present work deals separately with resonant absorption and scattering of neutrons. The influence of crystalline binding is considered for both types of reactions in the harmonic crystal frame work. The harmonic crystal model is applied to the study of resonant absorption cross sections to show the inconsistency of the free gas model widely in use in reactor neutronics. The errors due to the use of the latter were found to be non negligible. These errors should be corrected by introducing a more elaborated harmonic crystal model in codes for resonances analysis and on the nuclear data processing stage. Currently the influence of crystalline binding on transfer cross section in the resonance domain is taken into account in a naive manner using the model of the free nucleus at rest in the laboratory system. In this work I present a formalism (Uncoupled Phonon Approximation) which permits to consider in more detail the crystalline structure of the nuclear fuel. This formalism shows new features in comparison with the static model. (author)

  5. Study of the influence of chemical binding on resonant absorption and scattering of neutrons; Etude de l'influence des liaisons chimiques sur l'absorption et la diffusion des neutrons aux energies de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Naberejnev, D G [Aix-Marseille-1 Univ., 13 - Marseille (France)

    1999-02-01

    At present time the problem of taking into account of the crystalline binding in the heavy nuclei resonance range is not correctly treated in nuclear data processing codes. The present work deals separately with resonant absorption and scattering of neutrons. The influence of crystalline binding is considered for both types of reactions in the harmonic crystal frame work. The harmonic crystal model is applied to the study of resonant absorption cross sections to show the inconsistency of the free gas model widely in use in reactor neutronics. The errors due to the use of the latter were found to be non negligible. These errors should be corrected by introducing a more elaborated harmonic crystal model in codes for resonances analysis and on the nuclear data processing stage. Currently the influence of crystalline binding on transfer cross section in the resonance domain is taken into account in a naive manner using the model of the free nucleus at rest in the laboratory system. In this work I present a formalism (Uncoupled Phonon Approximation) which permits to consider in more detail the crystalline structure of the nuclear fuel. This formalism shows new features in comparison with the static model. (author)

  6. 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)

  7. 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.)

  8. Conformation and arrangement of polyelectrolytes in semi-diluted solution. A study by small angle neutrons scattering; Conformation et arrangement des polyelectrolytes en solution semi-diluee. Etude par diffusion des neutrons aux petits angles

    Energy Technology Data Exchange (ETDEWEB)

    Spiteri, M N

    1997-03-25

    Polyelectrolytes have particular physical and chemical properties and can thus be used for instance for petroleum production. Some of their microscopic properties have been studied in this work. With the particular zero average contrast technique, the small angle neutron scattering allows to directly know the form factors in semi-diluted solutions of polyelectrolytes where the chains are mixed. Another measure leads to the crystal structure. The electrostatic screen effects when salt is added in aqueous solutions of completely charged PSSNa solutions (f=1) (sodium polystyrene sulfonate) are studied. It seems that the chains take a vermiform conformation. Their persistence length varies as I{sup -1/3} (I is the ionic force). The hydrophobicity effects in partially charged PSSNa solutions (f<1) are given too. They lead to a progressive collapse of the chains when their charge rates decrease. The screen and condensation effects when the charge rate f of the PSSNa (f>f(Manning)) varies in a polar solvent (DMSO) are studied. The vermiform chains have the same persistence length (for each f) which varies as I{sup -1/4}. Lastly, the f variation effects in the case of a weakly charged hydrophilic poly-ion (f

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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)

  18. 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)

  19. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. 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.

  1. Time-of-flight spectrometer for slow neutrons in use at the reactor in Saclay. Its application for the study of the inelastic diffusion of cold neutrons; L'appareillage de spectrometrie a temps-de-vol pour neutrons lents en service a la pile de Saclay. Son application a l'etude de la diffusion inelastique des neutrons froids

    Energy Technology Data Exchange (ETDEWEB)

    Jacsot, B; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Galula, M [Centre National de la Recherche Scientifique (CNRS), 91 - Gif-sur-Yvette (France)

    1955-07-01

    The time-of-flight spectrometers is constituted of a mechanic swivel obturator which absorbs neutrons until energies above 1 KeV, a mechanic filter which allow to retain only high wave length components and a delayed pulses selector with 100 channels. Its main application field is the thermic region where it allowed to measure the inelastic scattering of neutrons using various materials as H{sub 2}O, D{sub 2}O, Be, BeO, etc... (M.P.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. 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)

  4. Simulation of complete neutron scattering experiments: from model systems to liquid germanium; Simulation complete d'une experience de diffusion de neutrons: des systemes modeles au germanium liquide

    Energy Technology Data Exchange (ETDEWEB)

    Hugouvieux, V

    2004-11-15

    In this thesis, both theoretical and experimental studies of liquids are done. Neutron scattering enables structural and dynamical properties of liquids to be investigated. On the theoretical side, molecular dynamics simulations are of great interest since they give positions and velocities of the atoms and the forces acting on each of them. They also enable spatial and temporal correlations to be computed and these quantities are also available from neutron scattering experiments. Consequently, the comparison can be made between results from molecular dynamics simulations and from neutron scattering experiments, in order to improve our understanding of the structure and dynamics of liquids. However, since extracting reliable data from a neutron scattering experiment is difficult, we propose to simulate the experiment as a whole, including both instrument and sample, in order to gain understanding and to evaluate the impact of the different parasitic contributions (absorption, multiple scattering associated with elastic and inelastic scattering, instrument resolution). This approach, in which the sample is described by its structure and dynamics as computed from molecular dynamics simulations, is presented and tested on isotropic model systems. Then liquid germanium is investigated by inelastic neutron scattering and both classical and ab initio molecular dynamics simulations. This enables us to simulate the experiment we performed and to evaluate the influence of the contributions from the instrument and from the sample on the detected signal. (author)

  5. Continuous and discreet methods in the aggregation and des fuzzy stages of a diffuse controller of neutron power; Metodos continuo y discreto en las etapas de agregacion y desdifusificacion de un controlador disfuso de potencia neutronica

    Energy Technology Data Exchange (ETDEWEB)

    Najera H, M.C.; Benitez R, J.S. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    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)

  6. 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

  7. 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.

  8. Characterization of the hydrogen bond in molecular systems of biological interest by neutron scattering; Caracterisation de la liaison hydrogene dans des systemes moleculaires d'interet biologique par diffusion de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Cavillon, F

    2004-10-15

    This work presents a methodology for the analysis of the scattering spectra of neutrons on molecular liquids. This method is based on the adjustment of the molecular form factor concerning great momentum transfer. The subtraction of the intra-molecular contributions gives access to information on inter-molecular interactions such as the hydrogen bond. 3 systems with increasing levels of difficulty have been studied: the ammonia molecule, the N-methyl-formamide (NMF) and the N-methyl-acetamide (NMA). The value we get for the N-D intermolecular distance of the liquid ammonia molecule is 1.7 angstrom, this value is different from the value generally admitted (2.3 angstrom) but we have validated it by studying the isotopic substitution N{sup 14}/N{sup 15}. The adjustment to the NMF is obtained with a good accuracy but the characterization of the hydrogen bound is more delicate to infer. A preliminary study of the NMA molecule shows that this method can give relevant results on complex molecules.

  9. 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)

  10. Small-Angle Scattering of Neutrons by Iron and Nickel Close to the Curie Point; Diffusion aux petits angles des neutrons par le fer et le nickel au voisinage du point de curie; Rasseyanie nejtronov pod malymi uglami zhelezom i nikelem, blizkikh k tochke kyuri; Dispersion de neutrones en angulos pequenos por el hierro y el niquel cerca del punto de curie

    Energy Technology Data Exchange (ETDEWEB)

    Jacrot, B.; Konstantinovic, J.; Parette, G.; Cribier, D. [Centre d' etudes Nucleaires de Saclay (France)

    1963-01-15

    Recent studies on critical scattering by iron have thrown fresh light on the results obtained previously and shown that: (1) At the Curie point, scattering is inelastic. The Lorentzian form of the lines and their angular dependence establish the existence of a scattering process having the diffusion coefficient 2m {Omega}/n = 7. (2) At the Curie point, the coefficient K{sub 1}{sup 2} is less than 5 x 10{sup -5} A{sup -2}. Iron and nickel have very different angular distribution. Small-angle scattering below T{sub c} has also been studied. (author) [French] De nouveaux travaux sur la diffusion critique par le fer ont permis de completer les resultats obtenus anterieurement. Ils etablissent les faits suivants: 1. Au point de Curie la diffusion est inelastique. La forme lorentzienne des raies et leur dependance angulaire etablissent l'existence d*un processus de diffusion avec un coefficient de diffusion 2m{Omega}/n = 7. 2. Au point de Curie le coefficient K{sub 1}{sup 2} est inferieur a 5 x 10{sup -5} A{sup -2}. La distribution angulaire est tres differente dans le cas du fer et du nickel. On a egalement etudie la diffusion aux petits angles en dessous de T{sub c}. (author) [Spanish] Nuevos trabajos sobre la dispersion critica por el hierro han permitido completar los resultados obtenidos anteriormente. Dichos trabajos prueban los hechos siguientes: 1) La dispersion es inelastica en el punto de Curie. La forma lorentziana de las rayas y su dependencia angular demuestran la existencia de un proceso de dispersion con un coeficiente de dispersion igual a 2m {Omega}/n = 7. 2) En el punto de Curie, el coeficiente K{sub 1}{sup 2} es inferior a 5 x 10{sup -5} A{sup -2}. La distribucion angular es muy diferente en el caso del hierro y del ni quel. Se ha estudiado tambien la dispersion en pequenos angulos por debajo de T{sub c}. (author)

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. Study by neutrons diffusion and X-rays of structural and magnetic properties of Bi{sub 2}Sr{sub 2}Ca{sub n-1}Cu{sub n}O{sub 2n+4+{delta}} type superconductive cuprates; Etude par diffusion des neutrons et des rayons X des proprietes structurales et magnetiques des cuprates supraconducteurs de type Bi{sub 2}Sr{sub 2}Ca{sub n-1}Cu{sub n}O{sub 2n+4+{delta}}

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliu-Doloc, L

    1995-09-22

    In this work we have used inelastic and elastic neutron and X-ray scattering techniques for characterizing the incommensurate structural distortions in compounds belonging to the family Bi{sub 2}Sr{sub 2}Ca{sub n-1}Cu{sub n}O{sub 2n+4+{delta}} of high-Tc superconducting cuprates. We have searched the existence of structural instabilities specific of the CuO{sub 2} planes and of magnetic instabilities. The modulated structure of the 2212 phase has been refined from single-crystal neutron diffraction results and importance of distortion of CuO{sub 2} planes has thus been determined. It is shown that the additional oxygen is not ordered three-dimensionally within the modulated structure and that the information about it is contained in diffuse scattering results. A model of the short-range order associated with additional oxygen atoms is proposed and discussed. The temperature studies of the long-range order have shown a great stability of the amplitude and period of the incommensurate distortion wave in the one-layer, as well as in the double-layer compounds, either superconducting or insulating. We find such a behaviour to be highly incompatible with a distortion resulting from a charge-density-wave instability. The results we have obtained indicate that the bismuth-based high-Tc superconducting cuprates have essentially the same physics of the CuO{sub 2} planes as the previous two families, La{sub 2-x}Sr{sub x}CuO{sub 4} and YBa{sub 2}Cu{sub 3}O{sub 6+{delta}}, being at the proximity of three instabilities: a metal-insulator transition, an antiferromagnetic instability and a structural instability specific of the CuO{sub 2} planes. (author).

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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.)

  9. 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

  10. 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.

  11. 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.

  12. 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.)

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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...

  18. 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

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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.

  5. Study of magnetic excitations in the high critical temperature La{sub 2-x}Sr{sub x}CuO{sub 4} superconductor with neutrons inelastic scattering; Etude des excitations magnetiques dans le supraconducteur a haute temperature critique La{sub 2-x}Sr{sub x}CuO{sub 4} par diffusion inelastique des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Petit, S

    1997-11-25

    In this work, we question the relevance of spin correlations in the problem of high-T{sub c} superconductivity. To characterize these correlations, we present a detailed inelastic neutron scattering study of the magnetic excitations spectrum, in the high T{sub c} La{sub 2-x}Sr{sub x}CuO{sub 4} copper oxide superconductor, as a function of temperature and strontium content x. Our study demonstrates the existence of two distinct types of fluctuations. The first type corresponds to two-dimensional isotropic fluctuations peaked at incommensurate points around the antiferromagnetic wavevector. These fluctuations show a gap below T{sub c}. The second type of fluctuations corresponds to two-dimensional very low energy, typically 1 MeV, antiferromagnetic correlations. They arise at low temperatures, for T {<=}T{sub f} {approx} 10 K {<=}T{sub c}. Different scenarios are discussed, namely particle-hole excitation arising because of nesting properties of the Fermi surface, stripes, and localized carriers in the CuO{sub 2} plane because of the La/Sr substitution disorder. Next, aiming to study the interrelation between magnetism and superconductivity, we present a theoretical perturbative treatment of the t-t`-J model. This model is believed to describe the physics of the CuO{sub 2} plane. The considered quasiparticles describe the motion of a doped hole, followed by localized spins on the copper sites. For a given doping x, two phases can exist, containing either x or 1 + x of these quasiparticles. We show at in the second phase, the exchange interaction between localized spins, combined to the intrinsic anharmonicity of these quasiparticles leads to an attractive interaction between them. We study the stability and the symmetry of a superconducting phase based on this mechanism. (author) 215 refs.

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. 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.)

  12. 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

  13. 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.

  14. 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.)

  15. 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

  16. Diffusion-weighted imaging of the pancreas; Diffusionsbildgebung des Pankreas

    Energy Technology Data Exchange (ETDEWEB)

    Gruenberg, K. [Deutsches Krebsforschungszentrum (DKFZ) Heidelberg, Abteilung Radiologie, E010, Heidelberg (Germany); Grenacher, L.; Klauss, M. [Universitaetsklinikum Heidelberg, Abt. Diagnostische und Interventionelle Radiologie, Heidelberg (Germany)

    2011-03-15

    , zwischen normalem, Tumor- und chronisch entzuendlich veraendertem Gewebe zu unterscheiden. In der Literatur werden fuer den so genannten effektiven Diffusionskoeffizienten (''apparent diffusion coefficient'', ADC) fuer gesundes Pankreasgewebe Werte zwischen 1,49 und 1,9 x 10{sup -3} mm{sup 2}/s und fuer Pankreaskarzinome Werte zwischen 1,24 und 1,46 x 10{sup -3} mm{sup 2}/s angegeben. Bei der Autoimmunpankreatitis liegt der mittlere ADC-Wert bei 1,012 x 10{sup -3} mm{sup 2}/s. Die Daten ueber die ADC-Werte zur Differenzierung von chronischer Pankreatitis und Pankreaskarzinomen sind in der Literatur allerdings widerspruechlich. Hier scheint aber die Verwendung des Perfusionsparameters f, der aus den Diffusionsdaten ueber das Intravoxel-incoherent-motion- (IVIM-)Modell berechnet werden kann, den entscheidenden Vorteil zu bringen, sofern er ueber mehrere b-Werte ermittelt wurde. Der Literatur nach liegt der mittlere f-Wert fuer chronisch entzuendlich veraendertes Pankreasgewebe bei ca. 16%, fuer Pankreaskarzinom bei ca. 8% und fuer gesundes Gewebe bei ca. 25%. Bei der Differenzierung der zystischen Pankreastumoren ist die DWI bisher nicht hilfreich. Es gibt derzeit zahlreiche Hinweise mit anderen Tumorentitaeten und im Tiermodell auf einen moeglichen Nutzen der DWI fuer die Verlaufsbeurteilung von Pankreastumoren unter Therapie, jedoch gibt es bei Pankreastumoren hierzu noch keine publizierte Originalarbeit. (orig.)

  17. 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.

  18. 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)

  19. Molecular Dynamics Investigated by Neutron Scattering; Etude de la Dynamique Moleculaire au Moyen de la Diffusion des Neutrons; 0418 0421 0421 041b 0415 0414 041e 0412 0410 041d 0418 0415 041c 041e 041b 0415 041a 0423 041b 042f 0420 041d 041e 0419 0414 0418 041d 0410 041c 0418 041a 0418 0421 041f 041e 041c 041e 0429 042c 042e 0420 0410 0421 0421 0415 042f 041d 0418 042f 041d 0415 0419 0422 0420 041e 041d 041e 0412 ; Estudio de la Dinamica Molecular por Dispersion Neutronica

    Energy Technology Data Exchange (ETDEWEB)

    Janik, J. A. [Institute for Nuclear Physics, Cracow (Poland)

    1965-06-15

    is discussed from the point of view of molecular dynamics. (author) [French] L'auteur fait le point de l'etat actuel des connaissances touchant le probleme de l'applicabilite des theories de Krieger- Nelkin et de Griffing dans le cas des gaz. Ensuite, sur la base d'experiences faites avec du methane liquide, il examine d'une maniere critique l'applicabilite du concept du tenseur de masse a des molecules dans l'etat condense. Les arguments les plus puissants qui militent contre l'application de la theorie de Krieger-Nelkin aux etats condenses sont: a) un deplacement dans la distribution des neutrons en energie, apres diffusion, vers les grandes energies; b) l'absence d'une relation de dependance entre la fraction inelastique et l'angle de diffusion. D'autres parties du memoire traitent de la dynamique de rotation des groupes ammonium dans les composes d'ammonium. L'auteur discute la plupart des donnees d'experience relatives aux halogenures d'ammonium en se fondant sur les experiences de differents auteurs. Far des mesures de neutrons, il a obtenu une liberte de rotation des groupes NH{sub 4} pour certaines substances, tandis que pour d'autres la rotation evolue en vibration de torsion. Dans le dernier cas, il a compare les frequences des vibrations de torsion obtenues par divers auteurs au cours d'experiences sur les neutrons avec celles qui resultent de la spectroscopie infrarouge et des mesures de la chaleur specifique. L'auteur discute egalement l'obstacle qui s'oppose a l'evaluation de la rotation a l'aide de mesures des sections efficaces totales pour les neutrons. En outre, il compare la dynamique de rotation des groupes NH{sub 4} dans NH{sub 4}ClO{sub 4} et des groupes H{sub 3}O dans H{sub 3}OCIO{sub 4}, en se fondant sur des experiences de diffusion inelastique des neutrons. Il a obtenu une rotation libre du groupe NH4 dans le perchlorate d'ammonium, meme a des temperatures ne depassant pas celle de l'azote liquide. Pour H{sub 3}OCIO{sub 4} il a obtenu une

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. A comparison of certain variational solutions of neutron diffusion equation

    International Nuclear Information System (INIS)

    Altiparmakov, D.V.; Milgram, M.S.

    1987-01-01

    Using the R-function theory and the variational method of Kantorovich, an approximate solution of the neutron diffusion equation is constructed for a homogeneous spatial domain of arbitrary shape. Calculations have been carried out by five different types of trial functions for certain characteristic domains of polygonal shape (square, triangle, hexagon, rhombus nad L-shaped domain). In the case of non-convex polygons, the consequence of the R-function solution is very poor and a separate treatment of singularity seems to be necessary. Compared to the R-function solution, the singular function development is mathematically more complicated but superior in respect to convergence rate. (author)

  7. 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

  8. 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).

  9. 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)

  10. 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

  11. Comparison of Spin Wave Theory with Neutron Scattering Results for MnF{sub 2}; Comparaison entre la Theorie de l'Onde de Spin et les Resultats Obtenus par la Diffusion des Neutrons dans MnF{sub 2}; Sravnenie teorij spinovoj volny s rezul'tatami rasseyaniya nejtronov dlya MnF{sub 2}; Estudio Comparativo, en el MnF{sub 2}, de la Teoria Ondulatoria del Spin y de los Resultados de la Dispersion Neutronica

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-04-15

    Recent inelastic neutron scattering experiments carried out by Okazaki and Turberfield have provided a direct measure of spin wave dispersion in MnF{sub 2} both at 4.2 Degree-Sign K and at higher temperatures approaching the Neel point. The low temperature results taken together with other magnetic data allow the various exchange constants in MnF{sub 2} to be evaluated with some precision. Using these constants, the spin wave formalism for anti-ferromagnets described by Oguchi has been employed to predict theoretically the temperature dependence of the spin wave dispersion. No disposable parameters are involved and theoretical dispersion curves may be compared with the experimental results for higher temperatures as a test of the spin wave theory. The spin wave energies were evaluated numerically and integrations over the Brillouin zone were carried out, thus avoiding the use of some form of series expansion in temperature. Account was taken of dynamical interactions between pairs of spin waves but kinematical effects were neglected. Renormalization of the spin wave energies was performed. This is an important factor for T/T{sub N}{>=} Vulgar-Fraction-One-Half The agreement between theory and experiment is in fact remarkably good up to temperatures corresponding to about 90% of the Neel point. It would be dangerous to extrapolate this result to systems of lower spin as kinematical effects are likely to play a much greater part in such cases. For MnF{sub 2} where S= 5/2 a rough estimate suggests that kinematical effects will not be important until M/M{sub O}-20% and this occurs only within a degree or so of the Neel point. A brief discussion of the likely consequences of a biquadratic exchange term in MnF{sub 2} is given and it is concluded that, if present, such exchange is quite small. (author) [French] Des experiences recentes, executees par Okazaki et Turberfield au moyen de la diffusion inelastique des neutrons, ont fourni une mesure directe de la dispersion de

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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).

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. Fourier convergence analysis applied to neutron diffusion Eigenvalue problem

    International Nuclear Information System (INIS)

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

    2004-01-01

    Fourier error analysis has been a standard technique for the stability and convergence analysis of linear and nonlinear iterative methods. Though the methods can be applied to Eigenvalue problems too, all the Fourier convergence analyses have been performed only for fixed source problems and a Fourier convergence analysis for Eigenvalue problem has never been reported. Lee et al proposed new 2-D/1-D coupling methods and they showed that the new ones are unconditionally stable while one of the two existing ones is unstable at a small mesh size and that the new ones are better than the existing ones in terms of the convergence rate. In this paper the convergence of method A in reference 4 for the diffusion Eigenvalue problem was analyzed by the Fourier analysis. The Fourier convergence analysis presented in this paper is the first one applied to a neutronics eigenvalue problem to the best of our knowledge

  11. 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

  12. 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

  13. 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)

  14. Parallel computing for homogeneous diffusion and transport equations in neutronics

    International Nuclear Information System (INIS)

    Pinchedez, K.

    1999-06-01

    Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)

  15. Diffuse neutron scattering study of metallic interstitial solid solutions

    International Nuclear Information System (INIS)

    Barberis, P.

    1991-10-01

    We studied two interstitial solid solutions (Ni-C(1at%) and Nb-O(2at%) and two stabilized zirconia (ZrO2-CaO(13.6mol%) and ZrO2-Y2O3(9.6mol%) by elastic diffuse neutron scattering. We used polarized neutron scattering in the case of the ferromagnetic Ni-based sample, in order to determine the magnetic perturbation induced by the C atoms. Measurements were made on single crystals in the Laboratoire Leon Brillouin (CEA-CNRS, Saclay, France). An original algorithm to deconvolve time-of-flight spectra improved the separation between elastically and inelastically scattered intensities. In the case of metallic solutions, we used a simple non-linear model, assuming that interstitials are isolated and located in octahedral sites. Results are: - in both compounds, nearest neighbours are widely displaced away from the interstitial, while next nearest neighbours come slightly closer. - the large magnetic perturbation induced by carbon in Nickel decreases with increasing distance on the three first neighbour shells and is in good agreement with the total magnetization variation. - no chemical order between solute atoms could be evidenced. Stabilized zirconia exhibit a strong correlation between chemical order and the large displacements around vacancies and dopants. (Author). 132 refs., 38 figs., 13 tabs

  16. 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)

  17. 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.

  18. 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

  19. Basic physics with ultra cold neutrons; Physique fondamentale avec des neutrons ultra froids

    Energy Technology Data Exchange (ETDEWEB)

    Protasov, K. [Laboratoire de Physique Subatomique et de Cosmologie, CNRS-IN2P3, Universite Joseph Fourier, INPG, Grenoble (France)

    2007-07-01

    A short introduction to the physics of Ultra Cold Neutrons (UCN) is given. It covers different aspects from their discovery, their major properties as well as their using in the three experiments of fundamental physics: measurements of the neutron life time and of its electric dipole moment and studies of neutrons quantum states in the Earth's gravitational field. (author)

  20. 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)

  1. 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.)

  2. 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.

  3. Facteurs predictifs de gravite des cellulites cervico-faciales diffuses ...

    African Journals Online (AJOL)

    Objective : Cervicofacial cellulitis is a dermohypodermitis which originates at the hypodermic fat from a burglary of a focus dental, Its gravity lies partly in its diffuse form with a risk of airway obstruction and mediastinitis and secondly in its venture with necrotizing fasciitis and necrotizing gangrene. In the absence of ...

  4. 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.)

  5. Study of the elastic scattering and of the (p,n) charge exchange reaction with neutron-rich light exotic beams; Etude de la diffusion elastique et de la reaction d`echange de charge (p,n) avec des faisceaux exotiques legers riches en neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Cortina Gil, D.

    1996-07-05

    We have measured at GANIL, with the high resolution spectrometer SPEG, the elastic scattering of several neutron rich secondary beams ({sup 6}He, {sup 10}Be and {sup 11}Be) on a polypropylene target and the charge exchange reaction p({sup 6}He, {sup 6}Li)n. These exotic beams were produced by nuclear fragmentation and re-focalized with the SISSI device (superconducting solenoids). The signature of a halo structure in these nuclei has been analysed. Special attention has been paid to several aspects of the associated calculations namely, the proton and neutron density distributions and the small binding energy for the last nucleons in these exotic nuclei. Break-up mechanisms are seen to play an important role in these nuclei. 100 refs.

  6. 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

  7. 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.

  8. Vectorization of three-dimensional neutron diffusion code CITATION

    International Nuclear Information System (INIS)

    Harada, Hiroo; Ishiguro, Misako

    1985-01-01

    Three-dimensional multi-group neutron diffusion code CITATION has been widely used for reactor criticality calculations. The code is expected to be run at a high speed by using recent vector supercomputers, when it is appropriately vectorized. In this paper, vectorization methods and their effects are described for the CITATION code. Especially, calculation algorithms suited for vectorization of the inner-outer iterative calculations which spend most of the computing time are discussed. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. The vectorized CITATION code is executed on the FACOM VP-100 and VP-200 computers, and is found to run over six times faster than the original code for a practical-scale problem. The initial value of the relaxation factor and the number of inner-iterations given as input data are also investigated since the computing time depends on these values. (author)

  9. Cooperative learning of neutron diffusion and transport theories

    International Nuclear Information System (INIS)

    Robinson, Michael A.

    1999-01-01

    A cooperative group instructional strategy is being used to teach a unit on neutron transport and diffusion theory in a first-year-graduate level, Reactor Theory course that was formerly presented in the traditional lecture/discussion style. Students are divided into groups of two or three for the duration of the unit. Class meetings are divided into traditional lecture/discussion segments punctuated by cooperative group exercises. The group exercises were designed to require the students to elaborate, summarize, or practice the material presented in the lecture/discussion segments. Both positive interdependence and individual accountability are fostered by adjusting individual grades on the unit exam by a factor dependent upon group achievement. Group collaboration was also encouraged on homework assignments by assigning each group a single grade on each assignment. The results of the unit exam have been above average in the two classes in which the cooperative group method was employed. In particular, the problem solving ability of the students has shown particular improvement. Further,the students felt that the cooperative group format was both more educationally effective and more enjoyable than the lecture/discussion format

  10. Thermal diffusivity of solids; Diffusivite thermique des solides

    Energy Technology Data Exchange (ETDEWEB)

    Kleimann, H; Fetiveau, Y; Richard, M; Eyraud, L; Eyraud, C; Elston, J [Institut National des Sciences Appliquees, Lyon (France); Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The authors recall the different methods of thermal diffusivity measurements and point out the possible causes of errors. They study in particular the sinusoidal methods of thermal attack using the measurement of either an amplitude or a phase, in which the controlled energy in-put is obtained either by Joule effect or by radiation. By the use of controlled systems, it is possible to apply those methods to the measurement of thermal diffusivity of their samples in a large scale of temperature. The authors describe their experimental achievements and give several results dealing especially with Al{sub 2}O{sub 3}, BeO, MgO. (author) [French] Les auteurs rappellent les diverses methodes de mesure de la diffusivite thermique et indiquent les sources d'erreurs possibles. Ils etudient particulierement les methodes d'attaque thermique sinusoidales utilisant soit une mesure d'amplitude, soit une mesure de phase dans lesquelles l'apport energetique controle est effectue soit par effet Joule, soit par rayonnement. La mise en oeuvre de systemes asservis permet l'utilisation de ces methodes pour la mesure de la diffusivite thermique d'echantillons minces dans une gamme etendue de temperature. Les auteurs decrivent les realisations experimentales et fournissent divers resultats concernant notamment Al{sub 2}O{sub 3}, BeO, MgO. (auteur)

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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).

  16. 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)

  17. 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.

  18. 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)

  19. 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

  20. 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

  1. 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 ...

  2. 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.)

  3. Finite element method for neutron diffusion problems in hexagonal geometry

    International Nuclear Information System (INIS)

    Wei, T.Y.C.; Hansen, K.F.

    1975-06-01

    The use of the finite element method for solving two-dimensional static neutron diffusion problems in hexagonal reactor configurations is considered. It is investigated as a possible alternative to the low-order finite difference method. Various piecewise polynomial spaces are examined for their use in hexagonal problems. The central questions which arise in the design of these spaces are the degree of incompleteness permissible and the advantages of using a low-order space fine-mesh approach over that of a high-order space coarse-mesh one. There is also the question of the degree of smoothness required. Two schemes for the construction of spaces are described and a number of specific spaces, constructed with the questions outlined above in mind, are presented. They range from a complete non-Lagrangian, non-Hermite quadratic space to an incomplete ninth order space. Results are presented for two-dimensional problems typical of a small high temperature gas-cooled reactor. From the results it is concluded that the space used should at least include the complete linear one. Complete spaces are to be preferred to totally incomplete ones. Once function continuity is imposed any additional degree of smoothness is of secondary importance. For flux shapes typical of the small high temperature gas-cooled reactor the linear space fine-mesh alternative is to be preferred to the perturbation quadratic space coarse-mesh one and the low-order finite difference method is to be preferred over both finite element schemes

  4. 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

  5. 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.)

  6. 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)

  7. 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...

  8. Studies of the Solid and Liquid Phases of HF, HCl and HBr by Slow-Neutron Inelastic Scattering; Etude des Phases Solide et Liquide de HF, HCl Et HBr par la Diffusion Inelastique des Neutrons Lents; 0418 0417 0423 0427 0415 041d 0418 0415 0422 0412 0415 0420 0414 042b 0425 0418 0416 0418 0414 041a 0418 0425 0424 0410 0417 HF, HCl I HBr 0421 041f 041e 041c 041e 0429 042c 042e 041d 0415 0423 041f 0420 0423 0413 041e 0413 041e 0420 0410 0421 0421 0415 042f 041d 0418 042f 041c 0415 0414 041b 0415 041d 041d 042b 0425 041d 0415 0419 0422 0420 041e 041d 041e 0412 ; Estudio de las Fases Solida y Liquida del HF, HCl Y HBr por Dispersion Inelastica de Neutrones Lentos

    Energy Technology Data Exchange (ETDEWEB)

    Boutin, H. [Picatinny Arsenal, Dover, NJ (United States); Brookhaven National Laboratory, Upton, NY (United States); Safford, G. J. [Brookhaven National Laboratory, Upton, NY (United States)

    1965-06-15

    be attributed to a weaker degree of association, which is expected from the decreasing electronegativity of the anion. The systematic study of these spectra in the various phases and at variais temperatures allows a distinction to be made between the low frequency motions characteristic of the lattice or of an hydrogen-bonded chain and those low frequency motions which involve nearest-neighbour interactions or short-range forces. The results of these measurements are in agreement with previous infrared data and indicate that hydrogen bonding is an influential factor in explaining the spectra of the solid phases of both HCl and HBr. These data are not in agreement, however, with theories that account for the solid phases of these compounds only in terms of simple dipole arrays or in terms of long-range interactions involving the dipole moment of one single unit interacting with an average electric field. (author) [French] Les auteurs ont etudie les phases liquide et solide de HF, HCl et HBr entre les raies d'absorption 30 et 1200 cm{sup -1} employant la methode de la diffusion inelastique des neutrons froids (0, 005 eV). Ils ont procede a ces mesures pour examiner systematiquement l'influence qu'un accroissement du degre d'electronegativite et la dimension des anions exercent sur les mouvements moleculaires de basse frequence dans la phase solide et la phase liquide. L'etude a d'abord porte sur HF du fait que les donnees obtenues par diffraction des rayons X avaient montre que dans ce compose a l'etat solide apparaissaient de longues chaines en zigzag liees par l'hydrogene. Les spectres des neutrons inelastiquement diffuses par les phases liquide et solide de HF font ressortir le haut degre d'association qui existe dans toute la phase liquide en raison de la liaison hydrogenee. Aux fins d'interpretation, les auteurs comparent les donnees relatives a HF a des donnees analogues relatives aux ions HF{sub 2}{sup -} et H{sub 2}F{sub 3}{sup -} dont les structures sont

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. Studies and modeling of cold neutron sources; Etude et modelisation des sources froides de neutron

    Energy Technology Data Exchange (ETDEWEB)

    Campioni, G

    2004-11-15

    With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources.

  4. 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

  5. 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.)

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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.)

  11. 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

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. Neutron diffusion in spheroidal, bispherical, and toroidal systems

    International Nuclear Information System (INIS)

    Williams, M.M.R.

    1986-01-01

    The neutron flux has been studied around absorbing bodies of spheroidal, bispherical, and toroidal shapes in an infinite nonabsorbing medium. Exact solutions have been obtained by using effective boundary conditions at the surfaces of the absorbing bodies. The problems considered are as follows: 1. Neutron flux and current distributions around prolate and oblate spheroids. It is shown that an equivalent sphere approximation can lead to accurate values for the rate of absorption. 2. Neutron flux and current in a bispherical system of unequal spheres. Three separate situations arise here: (a) two absorbing spheres, (b) two spherical sources, and (c) one spherical source and one absorbing sphere. It is shown how the absorption rate in the two spheres depends on their separation. 3. Neutron flux and current in a toroidal system: (a) an absorbing toroid and (b) a toroidal source. The latter case simulates the flux distribution from a thermonuclear reactor vessel. Finally, a brief description of how these techniques can be extended to multiregion problems is given

  17. Neutron diffusion approximation solution for the the three layer borehole cylindrical geometry. Pt. 1. Theoretical description

    International Nuclear Information System (INIS)

    Czubek, J.A.; Woznicka, U.

    1997-01-01

    A solution of the neutron diffusion equation is given for a three layer cylindrical coaxial geometry. The calculation is performed in two neutron-energy groups which distinguish the thermal and epithermal neutron fluxes in the media irradiated by the fast point neutron source. The aim of the calculation is to define the neutron slowing down and migration lengths which are observed at a given point of the system. Generally, the slowing down and migration lengths are defined for an infinite homogenous medium (irradiated by the point neutron source) as a quotient of the neutron flux moment of the (2n + 2)-order to the moment of the 2n-order. Czubek(1992) introduced in the same manner the apparent neutron slowing down length and the apparent migration length for a given multi-region cylindrical geometry. The solutions in the present paper are applied to the method of semi-empirical calibration of neutron well-logging tools. The three-region cylindrical geometry corresponds to the borehole of radius R 1 surrounded by the intermediate region (e.g. mud cake) of thickness (R 2 -R 1 ) and finally surrounded by the geological formation which spreads from R 2 up to infinity. The cylinders of an infinite length are considered. The paper gives detailed solutions for the 0-th, 2-nd and 4-th neutron moments of the neutron fluxes for each neutron energy group and in each cylindrical layer. A comprehensive list of the solutions for integrals containing Bessel functions or their derivatives, which are absent in common tables of integrals, is also included. (author)

  18. Neutron diffusion approximation solution for the the three layer borehole cylindrical geometry. Pt. 1. Theoretical description

    Energy Technology Data Exchange (ETDEWEB)

    Czubek, J.A.; Woznicka, U. [The H. Niewodniczanski Inst. of Nuclear Physics, Cracow (Poland)

    1997-12-31

    A solution of the neutron diffusion equation is given for a three layer cylindrical coaxial geometry. The calculation is performed in two neutron-energy groups which distinguish the thermal and epithermal neutron fluxes in the media irradiated by the fast point neutron source. The aim of the calculation is to define the neutron slowing down and migration lengths which are observed at a given point of the system. Generally, the slowing down and migration lengths are defined for an infinite homogenous medium (irradiated by the point neutron source) as a quotient of the neutron flux moment of the (2n{sup +}2)-order to the moment of the 2n-order. Czubek(1992) introduced in the same manner the apparent neutron slowing down length and the apparent migration length for a given multi-region cylindrical geometry. The solutions in the present paper are applied to the method of semi-empirical calibration of neutron well-logging tools. The three-region cylindrical geometry corresponds to the borehole of radius R{sub 1} surrounded by the intermediate region (e.g. mud cake) of thickness (R{sub 2}-R{sub 1}) and finally surrounded by the geological formation which spreads from R{sub 2} up to infinity. The cylinders of an infinite length are considered. The paper gives detailed solutions for the 0-th, 2-nd and 4-th neutron moments of the neutron fluxes for each neutron energy group and in each cylindrical layer. A comprehensive list of the solutions for integrals containing Bessel functions or their derivatives, which are absent in common tables of integrals, is also included. (author) 6 refs, 2 figs

  19. 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

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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.

  11. 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.

  12. 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)

  13. Diffusive instability of a kaon condensate in neutron star matter

    International Nuclear Information System (INIS)

    Kubis, Sebastian

    2004-01-01

    The beta equilibrated dense matter with kaon condensate is analyzed with respect to extended stability conditions, including charge fluctuations. This kind of the diffusive instability appeared to be common property in the kaon condensation case. Results for three different nuclear models are presented

  14. 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

  15. 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

  16. 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)

  17. Diffusion of water adsorbed in hydrotalcite: neutron scattering Study

    Energy Technology Data Exchange (ETDEWEB)

    Mitra, S [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai (India); Pramanik, A [Unilever Research India, Bangalore 500 066 (India); Chakrabarty, D [Godrej Sara Lee Limited, Research and Development Centre, Mumbai 400 079 (India); Juranyi, F [Laboratory for Neutron Scattering, ETHZ and PSI, CH-5232 Villigen PSI (Switzerland); Gautam, S [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai (India); Mukhopadhyay, R [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai (India)

    2007-12-15

    Layered double hydroxides (LDH) are a class of ionic lamellar solids with positively charged layers of two kinds of metallic cations and exchangeable hydrated anions. Quasi-elastic neutron scattering (QENS) measurements are performed in this type of LDH structured hydrated hydrotalcite sample to study the dynamical behaviour of the water in geometric confinement within the layers. Dynamical parameters correspond to the confined water molecules revealed that depending on the amount of excess water present, behaves differently and approaches bulk values at high concentration. Both translational and rotational dynamical parameters showed that at very low concentration of excess water, water molecules are attached to the surfaces and show the confinement effect.

  18. 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...

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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...

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    Energy Technology Data Exchange (ETDEWEB)

    Linde, Sven

    1960-06-15

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

  19. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    International Nuclear Information System (INIS)

    Linde, Sven

    1960-06-01

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

  20. A fast nodal neutron diffusion method for cartesian geometry

    International Nuclear Information System (INIS)

    Makai, M.; Maeder, C.

    1983-01-01

    A numerical method based on an analytical solution to the three-dimensional two-group diffusion equation has been derived assuming that the flux is a sum of the functions of one variable. In each mesh the incoming currents are used as boundary conditions. The final equations for the average flux and the outgoing currents are of the response matrix type. The method is presented in a form that can be extended to the general multigroup case. In the SEXI computer program developed on the basis of this method, the response matrix elements are recalculated in each outer iteration to minimize the data transfer between disk storage and central memory. The efficiency of the method is demonstrated for a light water reactor (LWR) benchmark problem. The SEXI program has been incorporated into the LWR simulator SILWER code as a possible option

  1. 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

  2. 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

  3. 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

  4. 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...

  5. 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

  6. 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

  7. 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

  8. 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

  9. On the exact solution for the multi-group kinetic neutron diffusion equation in a rectangle

    International Nuclear Information System (INIS)

    Petersen, C.Z.; Vilhena, M.T.M.B. de; Bodmann, B.E.J.

    2011-01-01

    In this work we consider the two-group bi-dimensional kinetic neutron diffusion equation. The solution procedure formalism is general with respect to the number of energy groups, neutron precursor families and regions with different chemical compositions. The fast and thermal flux and the delayed neutron precursor yields are expanded in a truncated double series in terms of eigenfunctions that, upon insertion into the kinetic equation and upon taking moments, results in a first order linear differential matrix equation with source terms. We split the matrix appearing in the transformed problem into a sum of a diagonal matrix plus the matrix containing the remaining terms and recast the transformed problem into a form that can be solved in the spirit of Adomian's recursive decomposition formalism. Convergence of the solution is guaranteed by the Cardinal Interpolation Theorem. We give numerical simulations and comparisons with available results in the literature. (author)

  10. Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup

    International Nuclear Information System (INIS)

    Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.

    2013-01-01

    In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation

  11. 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).

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. Diffusion-weighted magnetic resonance imaging of the abdomen; Diffusionsgewichtete Magnetresonanztomographie des Abdomens

    Energy Technology Data Exchange (ETDEWEB)

    Schmid-Tannwald, C.; Reiser, M.F.; Zech, C.J. [Klinikum der Ludwig-Maximilians-Universitaet Muenchen, Campus Grosshadern, Institut fuer Klinische Radiologie, Muenchen (Germany)

    2011-03-15

    Diffusion-weighted magnetic resonance imaging (DW-MRI) provides qualitative and quantitative information of tissue cellularity and the integrity of cellular membranes. Since DW-MRI can be performed without ionizing radiation exposure and contrast media application, DW-MRI is a particularly attractive tool for patients with allergies for gadolinium-based contrast agents or renal failure. Recent technical developments have made DW-MRI a robust and feasible technique for abdominal imaging. DW-MRI provides information on the detection and characterization of focal liver lesions and can also visualize treatment effects and early changes in chronic liver disease. In addition DW-MRI is a promising tool for the detection of inflammatory changes in patients with Crohn's disease. (orig.) [German] Die diffusionsgewichtete (DW-)MRT ermoeglicht die Erfassung qualitativer und quantitativer Informationen bzgl. der Gewebezellularitaet und Membranintegritaet. Die DW-MRT ist insbesondere bei Patienten mit einer Allergie gegen gadoliniumhaltige Kontrastmittel oder eingeschraenkter Nierenfunktion attraktiv, da ihr Einsatz nicht mit Strahlenexposition oder Kontrastmittelgabe verbunden ist. Durch technische Weiterentwicklungen ist die robuste Anwendung der DW-MRI in der Bildgebung des Abdomens seit einiger Zeit moeglich geworden. In der Leberdiagnostik lassen sich Zusatzinformationen zur Detektion und Charakterisierung von Leberlaesionen gewinnen, aber auch Therapieerfolge dokumentieren und fruehe chronische Leberveraenderungen visualisieren. Neben ihrer Rolle bei hepatologischen und onkologischen Fragestellungen erscheint der Einsatz der DW-MRT zudem bei entzuendlichen Fragestellungen wie dem Morbus Crohn sehr viel versprechend. (orig.)

  6. Thermal diffuse scattering in time-of-flight neutron diffraction studied on SBN single crystals

    International Nuclear Information System (INIS)

    Prokert, F.; Savenko, B.N.; Balagurov, A.M.

    1994-01-01

    At time-of-flight (TOF) diffractometer D N-2, installed at the pulsed reactor IBR-2 in Dubna, Sr x Ba 1-x Nb 2 O 6 mixed single crystals (SBN-x) of different compositions (0.50 < x< 0.75) were investigated between 15 and 773 K. The diffraction patterns were found to be strongly influenced by the thermal diffuse scattering (TDS). The appearance of the TDS from the long wavelength acoustic models of vibration in single crystals is characterized by the ratio of the velocity of sound to the velocity of neutron. Due to the nature of the TOF Laue diffraction technique used on D N-2, the TDS around Bragg peaks has rather a complex profile. An understanding of the TDS close to Bragg peaks is essential in allowing the extraction of the diffuse scattering occurring at the diffuse ferroelectric phase transition in SBN crystals. 11 refs.; 9 figs.; 1 tab. (author)

  7. 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

  8. 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

  9. 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)

  10. 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.

  11. 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

  12. 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)

  13. Diffusion et réception des chroniques : Chronica Naiarensis, Liber regum, Chronica regum Castellae

    OpenAIRE

    Arizaleta , Amaia

    2006-01-01

    International audience; Il est question dans ces pages des chroniques composées dans l'espace castillan au tournant du XIIe et du XIIIe siècles, afin d'avancer des informations sur le cadre pragmatique de réception des histoires royales qui devaient circuler au palais, ainsi que sur le plaisir de raconter que devaient ressentir les écrivains du roi et les auditeurs de la cour.

  14. Benchmarking a first-principles thermal neutron scattering law for water ice with a diffusion experiment

    Directory of Open Access Journals (Sweden)

    Holmes Jesse

    2017-01-01

    Full Text Available The neutron scattering properties of water ice are of interest to the nuclear criticality safety community for the transport and storage of nuclear materials in cold environments. The common hexagonal phase ice Ih has locally ordered, but globally disordered, H2O molecular orientations. A 96-molecule supercell is modeled using the VASP ab initio density functional theory code and PHONON lattice dynamics code to calculate the phonon vibrational spectra of H and O in ice Ih. These spectra are supplied to the LEAPR module of the NJOY2012 nuclear data processing code to generate thermal neutron scattering laws for H and O in ice Ih in the incoherent approximation. The predicted vibrational spectra are optimized to be representative of the globally averaged ice Ih structure by comparing theoretically calculated and experimentally measured total cross sections and inelastic neutron scattering spectra. The resulting scattering kernel is then supplied to the MC21 Monte Carlo transport code to calculate time eigenvalues for the fundamental mode decay in ice cylinders at various temperatures. Results are compared to experimental flux decay measurements for a pulsed-neutron die-away diffusion benchmark.

  15. 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

  16. 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.

  17. 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

  18. Performance of a parallel algorithm for solving the neutron diffusion equation on the hypercube

    International Nuclear Information System (INIS)

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

    1989-01-01

    The one-group, steady state neutron diffusion equation in two- dimensional Cartesian geometry is solved using the nodal method technique. By decoupling sets of equations representing the neutron current continuity along the length of rows and columns of computational cells a new iterative algorithm is derived that is more suitable to solving large practical problems. This algorithm is highly parallelizable and is implemented on the Intel iPSC/2 hypercube in three versions which differ essentially in the total size of communicated data. Even though speedup was achieved, the efficiency is very low when many processors are used leading to the conclusion that the hypercube is not as well suited for this algorithm as shared memory machines. 10 refs., 1 fig., 3 tabs

  19. 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)

  20. The spectrometer PERKEO III and the decay of free neutrons; Das Spektrometer PERKEO III und der Zerfall des freien Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Maerkisch, B.M.

    2006-10-18

    The Standard Model of particle physics describes neutron decay with three parameters. In contrast to that, a multitude of observables are accessible experimentally. With precision measurements sensitive tests of the Standard Model are thus possible. The neutron decay spectrometer PERKEO III was designed and built as part of this work. It will replace its predecessor PERKEO II, which has reached its systematical and statistical limits. With the new instrument measurements with continuous and pulsed neutron beams become feasible. This either provides an increase in statistics of up to two orders of magnitude, or eliminates the two major instrument specific sources of systematical corrections. In our first measurement in winter 2006/2007, the available event rate will be used to determine weak magnetism from the electron asymmetry A. Previously, this value was not statistically accessible in neutron decay. Systematics are analyzed with the help of our measurement with PERKEO II. For this measurement PERKEO III will be installed at the neutron guide H113 at the Institute Laue-Langevin, Grenoble. The neutron beam from this guide is characterized and a model is given, which allows the rapid calculation of beam profiles and absolute event rates from such a beam. In preparation of a future neutron decay instrument the reflective properties of two non-magnetic neutron mirrors were measured. (orig.)

  1. Study of the neutron strength function and of the diffusion radius as a function of mass number; Etude de la fonction densite de neutron et du rayon de diffusion en fonction du nombre de masse

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

    The strength functions S{sub 0} and S{sub 1}, and the scattering length R' are studied. The values of S{sub 0}, S{sub 1} and R' are obtained by shape analysis of the individual resonances observed in total neutron cross-section measurements. These measurements were performed at the Saclay 45 MeV linac by the time of flight method. The best resolution achieved was equal to 0.19 ns/m. We have particularly studied the nuclei around number of mass A = 50, 90, 140 and 200. The experimental values of S{sub 0}, S{sub 1} and R' are compared with optical model calculations. A fairly good agreement is obtained using an optical potential with surface absorption. (author) [French] Nous etudions les fonctions densite S{sub 0} et S{sub 1} ainsi que le rayon de diffusion R'. Les valeurs de S{sub 0}, S{sub 1} et R' sont determinees a partir de l'analyse de forme des resonances individuelles observees lors de la mesure de la section efficace neutronique totale. Cette mesure a ete faite aupres de l'accelerateur lineaire d'electrons de 45 MeV de Saclay par la methode du temps de vol. La meilleure resolution atteinte etait egale a 0.19 ns/m. Nous avons particulierement etudie les noyaux situes vers les nombres de masse A 50, 90, 140 et 200. Nous comparons ensuite les valeurs experimentales de S{sub 0}, S{sub 1} et R' a differents calculs de modele optique. Un bon accord est obtenu avec un potentiel optique a absorption de surface. (auteur)

  2. Study of the neutron strength function and of the diffusion radius as a function of mass number; Etude de la fonction densite de neutron et du rayon de diffusion en fonction du nombre de masse

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

    The strength functions S{sub 0} and S{sub 1}, and the scattering length R' are studied. The values of S{sub 0}, S{sub 1} and R' are obtained by shape analysis of the individual resonances observed in total neutron cross-section measurements. These measurements were performed at the Saclay 45 MeV linac by the time of flight method. The best resolution achieved was equal to 0.19 ns/m. We have particularly studied the nuclei around number of mass A = 50, 90, 140 and 200. The experimental values of S{sub 0}, S{sub 1} and R' are compared with optical model calculations. A fairly good agreement is obtained using an optical potential with surface absorption. (author) [French] Nous etudions les fonctions densite S{sub 0} et S{sub 1} ainsi que le rayon de diffusion R'. Les valeurs de S{sub 0}, S{sub 1} et R' sont determinees a partir de l'analyse de forme des resonances individuelles observees lors de la mesure de la section efficace neutronique totale. Cette mesure a ete faite aupres de l'accelerateur lineaire d'electrons de 45 MeV de Saclay par la methode du temps de vol. La meilleure resolution atteinte etait egale a 0.19 ns/m. Nous avons particulierement etudie les noyaux situes vers les nombres de masse A 50, 90, 140 et 200. Nous comparons ensuite les valeurs experimentales de S{sub 0}, S{sub 1} et R' a differents calculs de modele optique. Un bon accord est obtenu avec un potentiel optique a absorption de surface. (auteur)

  3. [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

  4. The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1994-01-01

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs

  5. Exact and approximate interior corner problem in neutron diffusion by integral transform methods

    International Nuclear Information System (INIS)

    Bareiss, E.H.; Chang, K.S.J.; Constatinescu, D.A.

    1976-09-01

    The mathematical solution of the neutron diffusion equation exhibits singularities in its derivatives at material corners. A mathematical treatment of the nature of these singularities and its impact on coarse network approximation methods in computational work is presented. The mathematical behavior is deduced from Green's functions, based on a generalized theory for two space dimensions, and the resulting systems of integral equations, as well as from the Kontorovich--Lebedev Transform. The effect on numerical calculations is demonstrated for finite difference and finite element methods for a two-region corner problem

  6. The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

    Energy Technology Data Exchange (ETDEWEB)

    Woznicki, Z I [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1994-12-31

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs.

  7. The numerical analysis of eigenvalue problem solutions in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1995-01-01

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iterations within global iterations. Particular iterative strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 35 figs, 16 tabs

  8. 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)

  9. Hydrogen rotational and translational diffusion in calcium borohydride from quasielastic neutron scattering and DFT

    DEFF Research Database (Denmark)

    Blanchard, Didier; Riktor, M.D.; Maronsson, Jon Bergmann

    2010-01-01

    Hydrogen dynamics in crystalline calcium borohydride can be initiated by long-range diffusion or localized motion such as rotations, librations, and vibrations. Herein, the rotational and translational diffusion were studied by quasielastic neutron scattering (QENS) by using two instruments...... with different time scales in combination with density functional theory (DFT) calculations. Two thermally activated reorientational motions were observed, around the 2-fold (C2) and 3-fold (C3) axes of the BH4− units, at temperature from 95 to 280K. The experimental energy barriers (EaC2 = 0.14 eV and EaC3 = 0...... of the interstitial H2 might come from the synthesis of the compound or a side reaction with trapped synthesis residue leading to the partial oxidation of the compound and hydrogen release....

  10. Quasi-elastic neutron scattering studies of the diffusion of hydrogen in metals

    Energy Technology Data Exchange (ETDEWEB)

    Ross, D K [Birmingham Univ. (UK). School of Physics and Space Research

    1989-01-01

    Quasi-elastic neutron scattering provides a uniquely detailed way of investigating microscopic models for diffusion in lattice gases. In the present paper we discuss extensions of the original Chudley-Elliott model to cover systems containing high concentrations of interacting particles for both the incoherent and coherent cases. In the former case, the peak width is changed by site blocking and by interactions and its shape is altered by correlation effects between successive jumps. In the coherent case, although interactions introduce different correlation effects, the most important changes are due to the short-range order caused by the interactions. A simple Mean Field theory is described which predicts peak narrowing where the diffuse scattering is at a maximum. Experimental tests of both coherent and incoherent theories are described for the case of {alpha}'NbD{sub x}. (orig.).

  11. Quasi-elastic neutron scattering studies of the diffusion of hydrogen in metals

    International Nuclear Information System (INIS)

    Ross, D.K.

    1989-01-01

    Quasi-elastic neutron scattering provides a uniquely detailed way of investigating microscopic models for diffusion in lattice gases. In the present paper we discuss extensions of the original Chudley-Elliott model to cover systems containing high concentrations of interacting particles for both the incoherent and coherent cases. In the former case, the peak width is changed by site blocking and by interactions and its shape is altered by correlation effects between successive jumps. In the coherent case, although interactions introduce different correlation effects, the most important changes are due to the short-range order caused by the interactions. A simple Mean Field theory is described which predicts peak narrowing where the diffuse scattering is at a maximum. Experimental tests of both coherent and incoherent theories are described for the case of α'NbD x . (orig.)

  12. A digital data acquisition system for a time of flight neutron diffuse scattering instrument

    International Nuclear Information System (INIS)

    Venegas, Rafael; Bacza, Lorena; Navarro, Gustavo

    1998-01-01

    Full text. We describe the design of a digital data acquisition system built for acquiring and storing the information produced by a neutron diffuse scattering apparatus. This instrument is based on the analysis of pulsed subthermal neutron which are scattered by a solid or liquid sample, measured as function of the scattered neutron wavelength and momentum direction. The time of flight neutron intensities on 14 different angular detector positions and two fission chambers must be analyzed simultaneously for each neutron burst. A PC controlled data acquisition board system was built based on two parallel multiscannning units, each with its own add-one counting unit, and a common base time generator. The unit plugs onto the ISA bus through an interface card. Two separate counting units were designed, to avoid possible access competition between low counting rate counters at off-axis positions and the higher rate frontal 0 deg and beam monitoring counters. the first unit contains logic for 14 independent and simultaneous multi scaling inputs, with 128 time channels and dwell time per channel of 5, 10 or 20 microseconds. Sweep trigger is synchronized with an electric signal from a coil sensing the rotor. The second unit contains logic for four additional multi scalers using the same external synchronizing signal, similar in all others details to the previously described multi scalers. Basic control routines for the acquisitions were written in C and a program for spectrum display and user interface was written in C ++ for a Windows 3.1 OS. A block diagram of the system is presented

  13. 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

  14. 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

  15. 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)

  16. 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

  17. Study of the diffusion movements of water by quasi-elastic scattering of slow neutrons

    International Nuclear Information System (INIS)

    Yamazaki, Ione Makiko

    1980-01-01

    The diffusion movements of water at three different temperatures in the liquid state have been studied by slow neutron quasi-elastic scattering. The measurements have been performed using the IPEN Triple Axis Spectrometer. Broadening and integrated intensity of the quasi-elastic line have been determined for several momentum transfer (K) in the range 0,7627 ≤ K ≤ 2,993 A -1 . The broadening of the quasi-elastic peaks as function of momentum transfer (K) observed at various temperatures has been interpreted in terms of globular diffusion models. The results obtained at 30 deg C have been explained in a consistent way considering the translational and rotational globular diffusion movements. To describe the results obtained at 55 deg and 70 deg C only the translational globular diffusion model was sufficient. This analysis indicates the existence in water of globules with distance of the farest proton position to the center of gravity of the globule 4,5 A, corroborating the idea of quasi-crystalline structure for water. The Debye-Waller factor has been obtained through the analysis of the integrated intensity of quasi-elastic scattering peaks over the K 2 measured range. From this analysis an estimative of the mean square displacement was obtained. (author)

  18. Solving the neutron diffusion equation on combinatorial geometry computational cells for reactor physics calculations

    International Nuclear Information System (INIS)

    Azmy, Y. Y.

    2004-01-01

    An approach is developed for solving the neutron diffusion equation on combinatorial geometry computational cells, that is computational cells composed by combinatorial operations involving simple-shaped component cells. The only constraint on the component cells from which the combinatorial cells are assembled is that they possess a legitimate discretization of the underlying diffusion equation. We use the Finite Difference (FD) approximation of the x, y-geometry diffusion equation in this work. Performing the same combinatorial operations involved in composing the combinatorial cell on these discrete-variable equations yields equations that employ new discrete variables defined only on the combinatorial cell's volume and faces. The only approximation involved in this process, beyond the truncation error committed in discretizing the diffusion equation over each component cell, is a consistent-order Legendre series expansion. Preliminary results for simple configurations establish the accuracy of the solution to the combinatorial geometry solution compared to straight FD as the system dimensions decrease. Furthermore numerical results validate the consistent Legendre-series expansion order by illustrating the second order accuracy of the combinatorial geometry solution, the same as standard FD. Nevertheless the magnitude of the error for the new approach is larger than FD's since it incorporates the additional truncated series approximation. (authors)

  19. 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

  20. 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

  1. 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.)

  2. Neutron penetrabilities using an optical model potential (1963); Penetrabilites des neutrons selon le modele du potentiel optique (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Mani, G S; Melkanoff, M A; Iori, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    The neutron penetrabilities and reaction cross-sections are presented in these tables for the case of an optical model potential for the nucleus, for nuclei up to mass 230. (authors) [French] Les penetrabilites et les sections efficaces de reaction pour les neutrons sont donnees dans ces tables en utilisant le modele du potentiel optique, pour les noyaux cibles jusqu'a la masse 230. (auteurs)

  3. Neutron penetrabilities using an optical model potential (1963); Penetrabilites des neutrons selon le modele du potentiel optique (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Mani, G.S.; Melkanoff, M.A.; Iori, I. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    The neutron penetrabilities and reaction cross-sections are presented in these tables for the case of an optical model potential for the nucleus, for nuclei up to mass 230. (authors) [French] Les penetrabilites et les sections efficaces de reaction pour les neutrons sont donnees dans ces tables en utilisant le modele du potentiel optique, pour les noyaux cibles jusqu'a la masse 230. (auteurs)

  4. Neutronic measurements of radioactive waste; Les mesures neutroniques des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Perot, B

    1997-12-31

    This document presents the general matters involved in the radioactive waste management and the different non destructive assays of radioactivity. The neutronic measurements used in the characterization of waste drums containing emitters are described with more details, especially the active neutronic interrogation assays with prompt or delayed neutron detection: physical principle, signal processing and evaluation of the detection limit. (author).

  5. Conjugate Gradient like methods and their application to fixed source neutron diffusion problems

    International Nuclear Information System (INIS)

    Suetomi, Eiichi; Sekimoto, Hiroshi

    1989-01-01

    This paper presents a number of fast iterative methods for solving systems of linear equations appearing in fixed source problems for neutron diffusion. We employed the conjugate gradient and conjugate residual methods. In order to accelerate the conjugate residual method, we proposed the conjugate residual squared method by transforming the residual polynomial of the conjugate residual method. Since the convergence of these methods depends on the spectrum of coefficient matrix, we employed the incomplete Choleski (IC) factorization and the modified IC (MIC) factorization as preconditioners. These methods were applied to some neutron diffusion problems and compared with the successive overrelaxation (SOR) method. The results of these numerical experiments showed superior convergence characteristics of the conjugate gradient like method with MIC factorization to the SOR method, especially for a problem involving void region. The CPU time of the MICCG, MICCR and MICCRS methods showed no great difference. In order to vectorize the conjugate gradient like methods based on (M)IC factorization, the hyperplane method was used and implemented on the vector computers, the HITAC S-820/80 and ETA10-E (one processor mode). Significant decrease of the CPU times was observed on the S-820/80. Since the scaled conjugate gradient (SCG) method can be vectorized with no manipulation, it was also compared with the above methods. It turned out the SCG method was the fastest with respect to the CPU times on the ETA10-E. These results suggest that one should implement suitable algorithm for different vector computers. (author)

  6. Artus code. Calculation of the energy transfer from neutrons to a solid; Programme Artus. Calcul des quantites d'energie cedees par des neutrons a un solide

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    1967-07-01

    The described codes have been realized for a particular study concerning the ionization influence of the CO{sub 2} by neutrons on the reaction: CO{sub 2} - graphite. The code hypothesis and formulation are presented and the application to more general physical problems is proposed. (A.L.B.)

  7. Artus code. Calculation of the energy transfer from neutrons to a solid; Programme Artus. Calcul des quantites d'energie cedees par des neutrons a un solide

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    1967-07-01

    The described codes have been realized for a particular study concerning the ionization influence of the CO{sub 2} by neutrons on the reaction: CO{sub 2} - graphite. The code hypothesis and formulation are presented and the application to more general physical problems is proposed. (A.L.B.)

  8. An iterative algorithm for solving the multidimensional neutron diffusion nodal method equations on parallel computers

    International Nuclear Information System (INIS)

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

    1992-01-01

    In this paper the one-group, steady-state neutron diffusion equation in two-dimensional Cartesian geometry is solved using the nodal integral method. The discrete variable equations comprise loosely coupled sets of equations representing the nodal balance of neutrons, as well as neutron current continuity along rows or columns of computational cells. An iterative algorithm that is more suitable for solving large problems concurrently is derived based on the decomposition of the spatial domain and is accelerated using successive overrelaxation. This algorithm is very well suited for parallel computers, especially since the spatial domain decomposition occurs naturally, so that the number of iterations required for convergence does not depend on the number of processors participating in the calculation. Implementation of the authors' algorithm on the Intel iPSC/2 hypercube and Sequent Balance 8000 parallel computer is presented, and measured speedup and efficiency for test problems are reported. The results suggest that the efficiency of the hypercube quickly deteriorates when many processors are used, while the Sequent Balance retains very high efficiency for a comparable number of participating processors. This leads to the conjecture that message-passing parallel computers are not as well suited for this algorithm as shared-memory machines

  9. 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)

  10. Analog simulation of concentration transients in a gaseous diffusion plant (1961); Etude sur simulateur des regimes transitoires des concentrations dans une installation de diffusion gazeuse (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Delarousse, P; Trouve, C; Jacques, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    A finite difference system is used to describe concentration transients in a gaseous diffusion plant for uranium isotope separation. The equipment used in this study is described and examples are given to illustrate the problems which have been solved with it. (authors) [French] Le comportement transitoire d'une cascade de diffusion gazeuse est represente de facon approchee par un systeme differentiel aux differences. On decrit le materiel analogique original qui a permis de simuler ce systeme. Une serie d'exemples illustre les differents problemes qui ont ete resolus au moyen de cet appareil. (auteurs)

  11. 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)

  12. An extension of diffusion theory for thermal neutrons near boundaries; Extension del campo de validez de la teoria de difusion para neutrones termico en las proximidades de bordes

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez Rivas, J L

    1963-07-01

    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 PUGC P/928 (1995), ANL-5872 (1959), DEGR 319 (D) (1961). This model could be extended also to plane geometry. (Author) 19 refs.

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. Study of the neutron slowing-down in graphite; Etude du ralentissement des neutrons dans le graphite

    Energy Technology Data Exchange (ETDEWEB)

    Martelly, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Duggal, V P [Commission a l' Energie atomique indienne (India)

    1955-07-01

    Study of the slowing-down age of neutrons in graphite. In the aim to eliminate the effect of resonance neutrons on the detector, neutrons captured by cadmium are studied with a classic method consisting of calculating the difference between the activity measured with and without screens. This method is described and screening properties of cadmium and gadolinium are compared. The experimental parameters and detectors details are described. The radioactive source is Ra-{alpha}-Be. The experimental results are given and the experimental distribution is compared with theoretical formula. In a second part, the spatial distribution of resonance neutrons in indium is discussed. Finally, the neutron slowing-down between the indium resonance and the thermal equilibrium is discussed as well as the research for the effective value of slowing-down age. The slowing-down age definition is given before calculating its effective value. It compared the slowing-down law with the experiment and the Gurney theory. (M.P.)

  20. Crystal structure and ion-diffusion pathway of inorganic materials through neutron diffraction

    International Nuclear Information System (INIS)

    Yashima, Masatomo

    2012-01-01

    The present brief review describes the application of neutron powder diffractometry and maximum-entropy method to the studies of crystal structure and diffusional pathways of mobile ions in ionic conducting ceramic materials. La 0.62 Li 0.16 TiO 3 and L i0.6 FePO 4 exhibit two- and one-dimensional networks of Li cation diffusional pathways, respectively. In the fluorite-structure ionic conductors such as celia solid solution Ce 0.93 Y 0.07 O 1.96 , bismuth oxide solid solution δ-Bi 1.4 Yb 0.6 O 3 and copper iodide CuI, a similar curved diffusion pathway along the directions is observed. In the cubic ABO 3 perovskite-type ionic conductor, lanthanum gallate solid solution, the mobile ions diffuse along a curved line keeping the interatomic distance between the B cation and O 2- anion. We have experimentally confirmed that the anisotropic thermal motions of the apex O2 atom and the interstitial O3 atoms are essential for the high oxygen permeability of the K 2 NiF 4 -type mixed conductor. Diffusion paths of proton are visualized along c axis in hexagonal hydroxyapatite. (author)

  1. 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)

  2. Analysis of diffuse scattering in neutron powder diagrams. Application to glassy carbon

    International Nuclear Information System (INIS)

    Boysen, H.

    1985-01-01

    From the quantitative analysis of the diffuse scattered intensity in powder diagrams valuable information about the disorder in crystals may be obtained. According to the dimensionality of this disorder (0D, 1D, 2D or 3D corresponding to diffuse peaks, streaks, planes or volume in reciprocal space) a characteristic modulation of the background is observed, which is described by specific functions. These are derived by averaging the appropriate cross sections over all crystallite orientations in the powder and folding with the resolution function of the instrument. If proper account is taken of all proportionality factors different components of the background can be put on one relative scale. The results are applied to two samples of glassy carbon differing in their degree of disorder. The neutron powder patterns contain contributions from 0D (00l peaks due to the stacking of graphitic layers), 1D (hkzeta streaks caused by the random orientation of these layers) and 3D (incoherent scattering, averaged thermal diffuse scattering, multiple scattering). From the fit to the observed data various parameters of the disorder like domain sizes, strains, interlayer distances, amount of incorporated hydrogen, pore sizes etc. are determined. It is shown that the omission of resolution corrections leads to false parameters. (orig.)

  3. Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry

    International Nuclear Information System (INIS)

    Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.

    2009-01-01

    In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)

  4. 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.

  5. 16 Élaboration et caractérisation des sources solides de diffusion ...

    African Journals Online (AJOL)

    master

    Un traitement thermique de séchage et de densification conduit ensuite au matériau désiré. Dans notre travail nous avons utilisé la technique de dépôt par rotation du substrat (spin coating) pour le dépôt des couches sur des substrats de silicium. Après le dépôt, un traitement thermique est réalisé afin d‟obtenir les ...

  6. 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

  7. 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.

  8. 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

  9. Applying the response matrix method for solving coupled neutron diffusion and transport problems

    International Nuclear Information System (INIS)

    Sibiya, G.S.

    1980-01-01

    The numerical determination of the flux and power distribution in the design of large power reactors is quite a time-consuming procedure if the space under consideration is to be subdivided into very fine weshes. Many computing methods applied in reactor physics (such as the finite-difference method) require considerable computing time. In this thesis it is shown that the response matrix method can be successfully used as an alternative approach to solving the two-dimension diffusion equation. Furthermore it is shown that sufficient accuracy of the method is achieved by assuming a linear space dependence of the neutron currents on the boundaries of the geometries defined for the given space. (orig.) [de

  10. Solution of two group neutron diffusion equation by using homotopy analysis method

    International Nuclear Information System (INIS)

    Cavdar, S.

    2010-01-01

    The Homotopy Analysis Method (HAM), proposed in 1992 by Shi Jun Liao and has been developed since then, is based on differential geometry as well as homotopy which is a fundamental concept in topology. It has proved to be useful for obtaining series solutions of many such problems involving algebraic, linear/non-linear, ordinary/partial differential equations, differential-integral equations, differential-difference equations, and coupled equations of them. Briefly, through HAM, it is possible to construct a continuous mapping of an initial guess approximation to the exact solution of the equation of concern. An auxiliary linear operator is chosen to construct such kind of a continuous mapping and an auxiliary parameter is used to ensure the convergence of series solution. We present the solutions of two-group neutron diffusion equation through HAM in this work. We also compare the results with that obtained by other well-known solution analytical and numeric methods.

  11. A numerical study of the eigenvalues in the neutron diffusion theory

    International Nuclear Information System (INIS)

    Lima Bezerra, J. de.

    1982-12-01

    A systematic numerical study for the eigenvalue problem in one dimension was carried out. A computer code RED2G was developed to obtain and to discuss a number of numerical solutions concerning eigenvalues problems originating from the discretization of the two groups neutron diffusion equation in one dimension and steady state. The problem of eigenvalues was created from the discretization by the method of finite differences. The solutions were obtained by four different iterative methods, i.e. Power, Wielandt-1, Wielandt-2 and accelerated Power with the Chebyshev polinomials. The numerical results given by the solution of the two test-problems indicate that the RED2G code is fast and efficient in these calculations and the Wielandt-2 method has been found to be the best both in respect of rapidity of calculations as well as programation effort required. (E.G.) [pt

  12. Use of diffusion bonded SS-Al composite material in the development of neutron detectors

    International Nuclear Information System (INIS)

    Alex, Mary; Prasad, K.R.; Pappachan, A.L.; Grover, A.K.; Krishnan, J.; Derose, D.J.; Bhanumurthy, K.; Kale, G.B.

    2005-01-01

    The present paper describes the development of a SS-Al composite plate in-house at BARC by diffusion bonding technique. Details of the several tests carried out on the composite material and the use of the plate in the development of a boron lined neutron chamber for Dhruva reactor control instrumentation has been described. The bonded sample has withstood tensile strength test, leak test and thermal cycling test and the leak rate was observed to be less than 3 x 10 -10 stdcc/sec. The chamber with the composite material has been installed in Dhruva Basket C location and connected to the log rate safety channel. It has been working successfully for the past two years. The use of SS-Al composite material has improved the reliability and long-term performance of the detector. (author)

  13. Three-dimensional multiple reciprocity boundary element method for one-group neutron diffusion eigenvalue computations

    International Nuclear Information System (INIS)

    Itagaki, Masafumi; Sahashi, Naoki.

    1996-01-01

    The multiple reciprocity method (MRM) in conjunction with the boundary element method has been employed to solve one-group eigenvalue problems described by the three-dimensional (3-D) neutron diffusion equation. The domain integral related to the fission source is transformed into a series of boundary-only integrals, with the aid of the higher order fundamental solutions based on the spherical and the modified spherical Bessel functions. Since each degree of the higher order fundamental solutions in the 3-D cases has a singularity of order (1/r), the above series of boundary integrals requires additional terms which do not appear in the 2-D MRM formulation. The critical eigenvalue itself can be also described using only boundary integrals. Test calculations show that Wielandt's spectral shift technique guarantees rapid and stable convergence of 3-D MRM computations. (author)

  14. HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1984-01-01

    1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    Spatial profiles of neutron emission are routinely obtained at the Joint European Torus (JET) from line-integrated emissivities measured with a multi-channel instrument. It is shown that the manner in which the emission profiles relax following 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%. The radial profiles of χ i for small minor radius (r/a 2 /s for H-mode plasmas with plasma current I p = 3.1 MA and toroidal field B T = 2.3T. The experimental value of χ i is smallest for Z eff = 2.2 and increases weakly with increasing Z eff . The experimental results disagree by two orders of magnitude with predictions from an ion temperature gradient driven turbulence model. (author) 6 refs., 3 figs

  16. The effect of, within the sphere confined, particle diffusion on the line shape of incoherent cold neutron scattering spectra

    International Nuclear Information System (INIS)

    Cvikl, B.; Dahlborg, U.; Calvo-Dahlborg, M.

    1999-01-01

    Based upon the model of particles diffusion within the sphere of partially absorbing boundaries, the possibilities of the detection, by the incoherent cold neutron scattering method, of particle precipitation on the boundary walls, has been investigated. The calculated scattering law as a function of the boundary absorption properties exhibits distinct characteristic which might, under favorable conditions, make such an experimental attempt feasible.(author)

  17. Some reciprocity-like relations in multi-group neutron diffusion and transport theory over bare homogeneous regions

    International Nuclear Information System (INIS)

    Modak, R.S.; Sahni, D.C.

    1996-01-01

    Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)

  18. A method of precise profile analysis of diffuse scattering for the KENS pulsed neutrons

    International Nuclear Information System (INIS)

    Todate, Y.; Fukumura, T.; Fukazawa, H.

    2001-01-01

    An outline of our profile analysis method, which is now of practical use for the asymmetric KENS pulsed thermal neutrons, are presented. The analysis of the diffuse scattering from a single crystal of D 2 O is shown as an example. The pulse shape function is based on the Ikeda-Carpenter function adjusted for the KENS neutron pulses. The convoluted intensity is calculated by a Monte-Carlo method and the precision of the calculation is controlled. Fitting parameters in the model cross section can be determined by the built-in nonlinear least square fitting procedure. Because this method is the natural extension of the procedure conventionally used for the triple-axis data, it is easy to apply with generality and versatility. Most importantly, furthermore, this method has capability of precise correction of the time shift of the observed peak position which is inevitably caused in the case of highly asymmetric pulses and broad scattering function. It will be pointed out that the accurate determination of true time-of-flight is important especially in the single crystal inelastic experiments. (author)

  19. Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

    Energy Technology Data Exchange (ETDEWEB)

    Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Maday, Yvon, E-mail: maday@ann.jussieu.fr [Sorbonne Universités, UPMC Univ Paris 06, UMR 7598, Laboratoire Jacques-Louis Lions and Institut Universitaire de France, F-75005, Paris (France); Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); Brown Univ, Division of Applied Maths, Providence, RI (United States); Riahi, Mohamed Kamel, E-mail: riahi@cmap.polytechnique.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CMAP, Inria-Saclay and X-Ecole Polytechnique, Route de Saclay, 91128 Palaiseau Cedex (France); Salomon, Julien, E-mail: salomon@ceremade.dauphine.fr [CEREMADE, Univ Paris-Dauphine, Pl. du Mal. de Lattre de Tassigny, F-75016, Paris (France)

    2014-12-15

    In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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.)

  5. 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

  6. 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

  7. 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.

  8. 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)

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

    Science.gov (United States)

    Faure, Bastien

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

  10. 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

  11. 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

  12. Determination of the response function for the Portsmouth Gaseous Diffusion Plant criticality accident alarm system neutron detectors

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; Brown, A.S.; Dobelbower, M.C.; Woollard, J.E.

    1997-03-01

    Neutron-sensitive radiation detectors are used in the Portsmouth Gaseous Diffusion Plant's (PORTS) criticality accident alarm system (CAAS). The CAAS is composed of numerous detectors, electronics, and logic units. It uses a telemetry system to sound building evacuation horns and to provide remote alarm status in a central control facility. The ANSI Standard for a CAAS uses a free-in-air dose rate to define the detection criteria for a minimum accident-of-concern. Previously, the free-in-air absorbed dose rate from neutrons was used for determining the areal coverge of criticality detection within PORTS buildings handling fissile materials. However, the free-in-air dose rate does not accurately reflect the response of the neutron detectors in use at PORTS. Because the cost of placing additional CAAS detectors in areas of questionable coverage (based on a free-in-air absorbed dose rate) is high, the actual response function for the CAAS neutron detectors was determined. This report, which is organized into three major sections, discusses how the actual response function for the PORTS CAAS neutron detectors was determined. The CAAS neutron detectors are described in Section 2. The model of the detector system developed to facilitate calculation of the response function is discussed in Section 3. The results of the calculations, including confirmatory measurements with neutron sources, are given in Section 4

  13. Investigation of correlations in light neutron-rich nuclei; Etude des correlations dans les noyaux legers riches en neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Normand, G

    2004-10-01

    Correlations play a crucial role in understanding the structure of light nuclei at and beyond the neutron drip-line. In this context, the two-neutron halo nucleus He{sup 6} and the unbound systems H{sup 5}, He{sup 7,9} and Li{sup 10} have been studied via measurements of the breakup of beams of He{sup 6} and Be{sup 11,12}. The CHARISSA and DEMON detector arrays were employed. The interpretation was facilitated by a simulation code (SILLAGE) which provided for the setup. In the case of He{sup 7}, the existence of an excited state with E{sub r} {approx} 1 MeV and gamma {approx} 0.75 MeV was confirmed. The virtual character of the s-wave ground state of Li{sup 10} was also confirmed and a scattering length of as {approx} -16 fm deduced. The results obtained for He{sup 9} suggest that a virtual s-wave state may exist just above threshold. The study of the three-body breakup of He{sup 6} found that the decay of the first 2+ state is essentially direct, while the decay of the remaining continuum strength is sequential - passage via He{sup 5}. Using the technique of intensity interferometry an rms separation between the halo neutrons of 7.7 +- 0.8 fm was derived. This result was confirmed by a complementary method utilizing Dalitz plots. In the case of H{sup 5}, the invariant mass spectrum was found to exhibit a broad (gamma {approx} 2 MeV) structure some 1.8 MeV above threshold. Comparison with recent three-body model calculations suggest that this corresponds to the predicted 1/2+ ground state. An rms valence neutron separation of some 5.5 fm was estimated. A search was also carried out for the 4n system using the Be{sup 12*} (2 alpha + Xn decay channel). No signal was observed beyond that expected on the basis of the known background processes. (author)

  14. Water diffusion through compacted clays analyzed by neutron scattering and tracer experiments

    International Nuclear Information System (INIS)

    Gonzalez Sanchez, F.

    2007-11-01

    /cm 3 , in order to reduce the pore sizes and to better study the dynamic properties of water close to the water-clay interface. We compared the water dynamics in fully hydrated compacted clays, at two significantly different time-space scales, in an attempt to distinguish the relevant features of the water transport. A fundamental microscopic investigation, tracing down to the atomic level was carried out, by neutron scattering, using time-of-flight and backscattering techniques. A classical macroscopic study was performed by using tracer through-diffusion methods. At the macroscopic level (time/spatial scale of about hours/mm to cm) the water diffusion depends strongly on the clay pore size and arrangement of the particles. However, at the microscopic level (time/spatial scale of about ten to hundred picosecond/10 -8 cm) the diffusion is governed by the local environment, which concerns to cations and clay surfaces and less to the particle arrangement. For a further understanding of this local environment, the water diffusion in clays was also measured at different hydration states, to vary the fraction of interlayer or external layer water, as compared to free pore water. The large difference in the diffusion paths of the two selected techniques makes a direct comparison of water diffusivities impossible. Therefore, two possibilities were established: An indirect comparison by connecting the results for diffusion coefficient at the two different scales through pure geometrical and electrostatic factors; and a direct comparison through the activation energy E a which was estimated from the dependence of the diffusion coefficients on the temperature. In contrast to the macroscopic diffusion coefficients, the activation energy is probably less influenced by geometrical factors and more by microscopic interactions, and thus could possibly be directly compared at the two different scales. The research was accomplished by a detailed characterization of the clay samples

  15. Water diffusion through compacted clays analyzed by neutron scattering and tracer experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Sanchez, F

    2007-11-15

    /cm{sup 3}, in order to reduce the pore sizes and to better study the dynamic properties of water close to the water-clay interface. We compared the water dynamics in fully hydrated compacted clays, at two significantly different time-space scales, in an attempt to distinguish the relevant features of the water transport. A fundamental microscopic investigation, tracing down to the atomic level was carried out, by neutron scattering, using time-of-flight and backscattering techniques. A classical macroscopic study was performed by using tracer through-diffusion methods. At the macroscopic level (time/spatial scale of about hours/mm to cm) the water diffusion depends strongly on the clay pore size and arrangement of the particles. However, at the microscopic level (time/spatial scale of about ten to hundred picosecond/10{sup -8} cm) the diffusion is governed by the local environment, which concerns to cations and clay surfaces and less to the particle arrangement. For a further understanding of this local environment, the water diffusion in clays was also measured at different hydration states, to vary the fraction of interlayer or external layer water, as compared to free pore water. The large difference in the diffusion paths of the two selected techniques makes a direct comparison of water diffusivities impossible. Therefore, two possibilities were established: An indirect comparison by connecting the results for diffusion coefficient at the two different scales through pure geometrical and electrostatic factors; and a direct comparison through the activation energy E{sub a} which was estimated from the dependence of the diffusion coefficients on the temperature. In contrast to the macroscopic diffusion coefficients, the activation energy is probably less influenced by geometrical factors and more by microscopic interactions, and thus could possibly be directly compared at the two different scales. The research was accomplished by a detailed characterization of the clay

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

    Energy Technology Data Exchange (ETDEWEB)

    Wohleber, X

    1997-11-17

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

  17. TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry

    International Nuclear Information System (INIS)

    Kristiansen, G.K.

    1987-01-01

    1 - Nature of physical problem solved: The two-dimensional neutron diffusion equation (x-y, r-z, or r-theta geometry is solved, either in the inhomogeneous (source calculation) or the homogeneous form (K eff calculation or absorber adjustment). The boundary conditions specify each group current as a linear homogeneous function of the group fluxes (gamma matrix concept). For each material, the fission matrix is assumed to by dyadic. 2 - Method of solution: Finite difference formulation (5 point scheme, mesh corner variant) is used. Solution technique: multi-line SOR. Eigenvalue estimate by neutron balance

  18. 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

  19. Fast resolution of the neutron diffusion equation through public domain Ode codes

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, V.M.; Vidal, V.; Garayoa, J. [Universidad Politecnica de Valencia, Departamento de Sistemas Informaticos, Valencia (Spain); Verdu, G. [Universidad Politecnica de Valencia, Departamento de Ingenieria Quimica y Nuclear, Valencia (Spain); Gomez, R. [I.E.S. de Tavernes Blanques, Valencia (Spain)

    2003-07-01

    The time-dependent neutron diffusion equation is a partial differential equation with source terms. The resolution method usually includes discretizing the spatial domain, obtaining a large system of linear, stiff ordinary differential equations (ODEs), whose resolution is computationally very expensive. Some standard techniques use a fixed time step to solve the ODE system. This can result in errors (if the time step is too large) or in long computing times (if the time step is too little). To speed up the resolution method, two well-known public domain codes have been selected: DASPK and FCVODE that are powerful codes for the resolution of large systems of stiff ODEs. These codes can estimate the error after each time step, and, depending on this estimation can decide which is the new time step and, possibly, which is the integration method to be used in the next step. With these mechanisms, it is possible to keep the overall error below the chosen tolerances, and, when the system behaves smoothly, to take large time steps increasing the execution speed. In this paper we address the use of the public domain codes DASPK and FCVODE for the resolution of the time-dependent neutron diffusion equation. The efficiency of these codes depends largely on the preconditioning of the big systems of linear equations that must be solved. Several pre-conditioners have been programmed and tested; it was found that the multigrid method is the best of the pre-conditioners tested. Also, it has been found that DASPK has performed better than FCVODE, being more robust for our problem.We can conclude that the use of specialized codes for solving large systems of ODEs can reduce drastically the computational work needed for the solution; and combining them with appropriate pre-conditioners, the reduction can be still more important. It has other crucial advantages, since it allows the user to specify the allowed error, which cannot be done in fixed step implementations; this, of course

  20. 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

  1. Second Colloquim on the Scattering of Ultrasonic Waves (Deuxieme Colloque sur la Diffusion des Ondes Ultrasonores).

    Science.gov (United States)

    1985-01-29

    8217 eau , Huiti me colloque sur le traitement du signal et ses applications (GRETSI), Nice, 1-5 June 1981, p. 637-644. [16] G. Maze, B. Taconet, and J...December 1984, to appear as a special issue of Traitement du Signal, published by La Direction des Recherches, Etudes et Techniques, 26 blvd. Victor...submerged in water, Huiti~me collogue sur le traitement du signal et ses applications (GRETSI), Nice, 1-5 June 1981, p. 637-644. [19] G. Maze et J

  2. Continuous measurement of neutron flux with the help of a converter; Mesure continue des flux de neutrons a l'aide d'un dispositif convertisseur

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    The device described in this report is working from this following principle: an element, the thermal neutron activate of which has a short period and is emitter of a {beta} particle, is surrounded with a dielectric which is itself surrounded by a conducting metal. {beta} particles are stopped in the conducting metal. Thus an electric current rises between the emitter element and the collector metal. The experimental study has been-done with rhodium as emitter. The linearity as function of thermal flux of the power of the pile, the accuracy, the efficiency of the device have been measured and are satisfactory. Just the stability of the measure equipment is not very good. This method gives, us at moment the instantaneous flux and the integrated flux, is well adapted for the measurements in pile to execute on loops. It is possible to continue to perfect the process. (authors) [French] Le dispositif decrit dans ce rapport fonctionne d'apres le principe suivant: un corps, dont le produit d'activation en neutrons thermiques a une courte periode et est emetteur d'une particule {beta}, est entoure d'un dielectrique lui-meme entoure d'un metal conducteur. Les particules {beta} sont arretees dans le metal conducteur. Il y a ainsi naissance d'un courant electrique entre le corps emetteur et le metal collecteur. L'etude experimentale a ete menee avec le rhodium comme emetteur. La linearite en fonction du flux thermique, de la puissance de la pile, la precision, le rendement du dispositif ont ete mesures et sont satisfaisants. Seule la stabilite de l'appareillage de mesure n'est pas tres bonne. Cette methode, qui peut nous donner a tout instant le flux instantane et le flux integre, est tout a fait adaptee aux mesures en pile a effectuer. sur des boucles. On peut poursuivre la mise au point du procede. (auteurs)

  3. Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry

    International Nuclear Information System (INIS)

    Valdes Parra, J.J.

    1986-01-01

    One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as 'acceleration of synthetic diffusion' which has been applied to solve the neutron transport equation with 'classical schemes of spatial integration' obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author)

  4. A study of the consistent and the lumped source approximations in finite element neutron diffusion calculations

    International Nuclear Information System (INIS)

    Ozgener, B.; Azgener, H.A.

    1991-01-01

    In finite element formulations for the solution of the within-group neutron diffusion equation, two different treatments are possible for the group source term: the consistent source approximation (CSA) and the lumped source approximation (LSA). CSA results in intra-group scattering and fission matrices which have the same nondiagonal structure as the global coefficient matrix. This situation might be regarded as a disadvantage, compared to the conventional (i.e. finite difference) methods where the intra-group scattering and fission matrices are diagonal. To overcome this disadvantage, LSA could be used to diagonalize these matrices. LSA is akin to the lumped mass approximation of continuum mechanics. We concentrate on two different aspects of the source approximations. Although it has been reported that LSA does not modify the asymptotic h 2 convergence behaviour for linear elements, the effect of LSA on convergence of higher degree elements has not been investigated. Thus, we would be interested in determining, p, the asymptotic order of convergence, in: Δk |k eff (analytical) -k eff (finite element)| = Ch p (1) for finite element approximations of varying degree (N) with both of the source approximations. Since (1) is valid in the asymptotic limit, we must use ultra-fine meshes and quadruple precision arithmetic. For our order of convergence study, we used infinite cylindrical geometry with azimuthal symmetry. Hence, the effects of singularities remain uninvestigated. The second aspect we dwell on is the performance of LSA in bilinear 3-D finite element calculations, compared to CSA. LSA has been used quite extensively in 1- and 2-D even-parity transport and diffusion calculations. In this work, we will try to assess the relative merits of LSA and CSA in 3-D problems. (author)

  5. Main results obtained in France in the development of the gaseous diffusion process for uranium isotope separation; Principaux resultats obtenus en France dans les etudes sur la separation des isotopes de l'uranium par diffusion gazeuse

    Energy Technology Data Exchange (ETDEWEB)

    Frejacques, C; Bilous, O; Dixmier, J; Massignon, D; Plurien, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The main problems which occur in the study of uranium isotope separation by the gaseous diffusion process, concern the development of the porous barrier, the corrosive nature of uranium hexafluoride and also the chemical engineering problems related to process design and the choice of best plant and stage characteristics. Porous barriers may be obtained by chemical attack of non porous media or by agglomeration of very fine powders. Examples of these two types of barriers are given. A whole set of measurement techniques were developed for barrier structure studies, to provide control and guidance of barrier production methods. Uranium hexafluoride reactivity and corrosive properties are the source of many difficult technological problems. A high degree of plant leak tightness must be achieved. This necessity creates a special problem in compressor bearing design. Barrier lifetime is affected by the corrosive properties of the gas, which may lead to a change of barrier structure with time. Barrier hexafluoride permeability measurements have helped to make a systematic study of this point. Finally an example of a plant flowsheet, showing stage types and arrangements and based on a minimisation of enriched product costs is also given as an illustration of some of the chemical engineering problems present. (author) [French] Les principaux problemes qui se sont poses dans l'etude de la separation des isotopes de l'uranium par diffusion gazeuse, sont ceux relatifs a l'obtention de barrieres poreuses, ceux lies a l'utilisation de l'hexafluorure d'uranium, enfin les problemes de genie chimique relatifs au procede et a l'agencement optimum des etages et des cascades entre elles. On peut obtenir des barrieres poreuses soit par attaque de membranes pleines, soit par agglomeration de poudres de petites dimensions. Des exemples de ces deux types de barrieres seront donnes. L'etude des proprietes de texture des barrieres obtenues, necessaire pour orienter les recherches de

  6. CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

    International Nuclear Information System (INIS)

    Ahnert, Carol; Aragones, Jose M.

    1983-01-01

    1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

  7. Diffusion

    International Nuclear Information System (INIS)

    Kubaschewski, O.

    1983-01-01

    The diffusion rate values of titanium, its compounds and alloys are summarized and tabulated. The individual chemical diffusion coefficients and self-diffusion coefficients of certain isotopes are given. Experimental methods are listed which were used for the determination of diffusion coefficients. Some values have been taken over from other studies. Also given are graphs showing the temperature dependences of diffusion and changes in the diffusion coefficient with concentration changes

  8. 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

  9. 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.)

  10. Iterative Two- and One-Dimensional Methods for Three-Dimensional Neutron Diffusion Calculations

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Lee, Deokjung; Downar, Thomas J.

    2005-01-01

    Two methods are proposed for solving the three-dimensional neutron diffusion equation by iterating between solutions of the two-dimensional (2-D) radial and one-dimensional (1-D) axial solutions. In the first method, the 2-D/1-D equations are coupled using a current correction factor (CCF) with the average fluxes of the lower and upper planes and the axial net currents at the plane interfaces. In the second method, an analytic expression for the axial net currents at the interface of the planes is used for planar coupling. A comparison of the new methods is made with two previously proposed methods, which use interface net currents and partial currents for planar coupling. A Fourier convergence analysis of the four methods was performed, and results indicate that the two new methods have at least three advantages over the previous methods. First, the new methods are unconditionally stable, whereas the net current method diverges for small axial mesh size. Second, the new methods provide better convergence performance than the other methods in the range of practical mesh sizes. Third, the spectral radii of the new methods asymptotically approach zero as the mesh size increases, while the spectral radius of the partial current method approaches a nonzero value as the mesh size increases. Of the two new methods proposed here, the analytic method provides a smaller spectral radius than the CCF method, but the CCF method has several advantages over the analytic method in practical applications

  11. Boundary element methods applied to two-dimensional neutron diffusion problems

    International Nuclear Information System (INIS)

    Itagaki, Masafumi

    1985-01-01

    The Boundary element method (BEM) has been applied to two-dimensional neutron diffusion problems. The boundary integral equation and its discretized form have been derived. Some numerical techniques have been developed, which can be applied to critical and fixed-source problems including multi-region ones. Two types of test programs have been developed according to whether the 'zero-determinant search' or the 'source iteration' technique is adopted for criticality search. Both programs require only the fluxes and currents on boundaries as the unknown variables. The former allows a reduction in computing time and memory in comparison with the finite element method (FEM). The latter is not always efficient in terms of computing time due to the domain integral related to the inhomogeneous source term; however, this domain integral can be replaced by the equivalent boundary integral for a region with a non-multiplying medium or with a uniform source, resulting in a significant reduction in computing time. The BEM, as well as the FEM, is well suited for solving irregular geometrical problems for which the finite difference method (FDM) is unsuited. The BEM also solves problems with infinite domains, which cannot be solved by the ordinary FEM and FDM. Some simple test calculations are made to compare the BEM with the FEM and FDM, and discussions are made concerning the relative merits of the BEM and problems requiring future solution. (author)

  12. Coarse-grain parallel solution of few-group neutron diffusion equations

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Turinsky, P.J.

    1991-01-01

    The authors present a parallel numerical algorithm for the solution of the finite difference representation of the few-group neutron diffusion equations. The targeted architectures are multiprocessor computers with shared memory like the Cray Y-MP and the IBM 3090/VF, where coarse granularity is important for minimizing overhead. Most of the work done in the past, which attempts to exploit concurrence, has concentrated on the inner iterations of the standard outer-inner iterative strategy. This produces very fine granularity. To coarsen granularity, the authors introduce parallelism at the nested outer-inner level. The problem's spatial domain was partitioned into contiguous subregions and assigned a processor to solve for each subregion independent of all other subregions, hence, processors; i.e., each subregion is treated as a reactor core with imposed boundary conditions. Since those boundary conditions on interior surfaces, referred to as internal boundary conditions (IBCs), are not known, a third iterative level, the recomposition iterations, is introduced to communicate results between subregions

  13. Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

    KAUST Repository

    Wang, Yaqi

    2009-04-01

    Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.

  14. Nodal integral method for the neutron diffusion equation in cylindrical geometry

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1987-01-01

    The nodal methodology is based on retaining a higher a higher degree of analyticity in the process of deriving the discrete-variable equations compared to conventional numerical methods. As a result, extensive numerical testing of nodal methods developed for a wide variety of partial differential equations and comparison of the results to conventional methods have established the superior accuracy of nodal methods on coarse meshes. Moreover, these tests have shown that nodal methods are more computationally efficient than finite difference and finite-element methods in the sense that they require shorter CPU times to achieve comparable accuracy in the solutions. However, nodal formalisms and the final discrete-variable equations they produce are, in general, more complicated than their conventional counterparts. This, together with anticipated difficulties in applying the transverse-averaging procedure in curvilinear coordinates, has limited the applications of nodal methods, so far, to Cartesian geometry, and with additional approximations to hexagonal geometry. In this paper the authors report recent progress in deriving and numerically implementing a nodal integral method (NIM) for solving the neutron diffusion equation in cylindrical r-z geometry. Also, presented are comparisons of numerical solutions to two test problems with those obtained by the Exterminator-2 code, which indicate the superior accuracy of the nodal integral method solutions on much coarser meshes

  15. Short-range order studies in nonstoichiometric transition metal carbides and nitrides by neutron diffuse scattering

    International Nuclear Information System (INIS)

    Priem, Thierry

    1988-01-01

    Short-range order in non-stoichiometric transition metal carbides and nitrides (TiN 0.82 , TiC 0.64 , TiC 0.76 , NbC 0.73 and NbC 0.83 ) was investigated by thermal neutron diffuse scattering on G4-4 (L.L.B - Saclay) and D10 (I.L.L. Grenoble) spectrometers. From experimental measurements, we have found that metalloid vacancies (carbon or nitrogen) prefer the f.c.c. third neighbour positions. Ordering interaction energies were calculated within the Ising model framework by three approximations: mean field (Clapp and Moss formula), Monte-Carlo simulation, Cluster variation Method. The energies obtained by the two latter methods are very close, and in qualitative agreement with theoretical values calculated from the band structure. Theoretical phase diagrams were calculated from these ordering energies for TiN x and TiC x ; three ordered structures were predicted, corresponding to compositions Ti 6 N 5 Ti 2 C and Ti 3 C 2 . On the other hand, atomic displacements are induced by vacancies. The metal first neighbours were found to move away from a vacancy, whereas the second neighbours move close to it. Near neighbour atomic displacements were theoretically determined by the lattice statics formalism with results in good agreement with experiment. (author) [fr

  16. TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron

    International Nuclear Information System (INIS)

    Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.

    1975-01-01

    1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I

  17. 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)); Adam, 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)

    1992-01-01

    Spatial profiles of neutron emission are routinely obtained at the Joint European Torus (JET) from line-integrated emissivities measured with a multi-channel instrument. It is shown that the manner in which the emission profiles relax following 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%. The radial profiles of [chi][sub i] for small minor radius (r/a < 0.6) were found to be flat and to take values between 0.3 and 1.1 m[sup 2]/s for H-mode plasmas with plasma current I[sub p] = 3.1 MA and toroidal field B[sub T] = 2.3T. The experimental value of [chi][sub i] is smallest for Z[sub eff] = 2.2 and increases weakly with increasing Z[sub eff]. The experimental results disagree by two orders of magnitude with predictions from an ion temperature gradient driven turbulence model. (author) 6 refs., 3 figs.

  18. On progress of the solution of the stationary 2-dimensional neutron diffusion equation: a polynomial approximation method with error analysis

    International Nuclear Information System (INIS)

    Ceolin, C.; Schramm, M.; Bodmann, B.E.J.; Vilhena, M.T.

    2015-01-01

    Recently the stationary neutron diffusion equation in heterogeneous rectangular geometry was solved by the expansion of the scalar fluxes in polynomials in terms of the spatial variables (x; y), considering the two-group energy model. The focus of the present discussion consists in the study of an error analysis of the aforementioned solution. More specifically we show how the spatial subdomain segmentation is related to the degree of the polynomial and the Lipschitz constant. This relation allows to solve the 2-D neutron diffusion problem for second degree polynomials in each subdomain. This solution is exact at the knots where the Lipschitz cone is centered. Moreover, the solution has an analytical representation in each subdomain with supremum and infimum functions that shows the convergence of the solution. We illustrate the analysis with a selection of numerical case studies. (author)

  19. On progress of the solution of the stationary 2-dimensional neutron diffusion equation: a polynomial approximation method with error analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ceolin, C., E-mail: celina.ceolin@gmail.com [Universidade Federal de Santa Maria (UFSM), Frederico Westphalen, RS (Brazil). Centro de Educacao Superior Norte; Schramm, M.; Bodmann, B.E.J.; Vilhena, M.T., E-mail: celina.ceolin@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

    2015-07-01

    Recently the stationary neutron diffusion equation in heterogeneous rectangular geometry was solved by the expansion of the scalar fluxes in polynomials in terms of the spatial variables (x; y), considering the two-group energy model. The focus of the present discussion consists in the study of an error analysis of the aforementioned solution. More specifically we show how the spatial subdomain segmentation is related to the degree of the polynomial and the Lipschitz constant. This relation allows to solve the 2-D neutron diffusion problem for second degree polynomials in each subdomain. This solution is exact at the knots where the Lipschitz cone is centered. Moreover, the solution has an analytical representation in each subdomain with supremum and infimum functions that shows the convergence of the solution. We illustrate the analysis with a selection of numerical case studies. (author)

  20. On the group approximation errors in description of neutron slowing-down at large distances from a source. Diffusion approach

    International Nuclear Information System (INIS)

    Kulakovskij, M.Ya.; Savitskij, V.I.

    1981-01-01

    The errors of multigroup calculating the neutron flux spatial and energy distribution in the fast reactor shield caused by using group and age approximations are considered. It is shown that at small distances from a source the age theory rather well describes the distribution of the slowing-down density. With the distance increase the age approximation leads to underestimating the neutron fluxes, and the error quickly increases at that. At small distances from the source (up to 15 lengths of free path in graphite) the multigroup diffusion approximation describes the distribution of slowing down density quite satisfactorily and at that the results almost do not depend on the number of groups. With the distance increase the multigroup diffusion calculations lead to considerable overestimating of the slowing-down density. The conclusion is drawn that the group approximation proper errors are opposite in sign to the error introduced by the age approximation and to some extent compensate each other

  1. 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.

  2. Investigation of the diffusion {proportional_to} brasses using methods depending on the evaporation or condensation of zinc; Etude de la diffusion dans les laitons {proportional_to} au moyen des methodes d'evaporation ou de condensation du zinc

    Energy Technology Data Exchange (ETDEWEB)

    Accary, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Centre d' Etudes de Chimie Metallurgique du CNRS (France)

    1959-07-01

    Diffusion in {alpha} brasses has been investigated using methods involving the evaporation and the condensation of zinc. Having shown that at sufficiently high temperatures intergranular diffusion has no effect, it was then proved that the rate of evaporation or of condensation can only be defined if the mechanical treatment of the test piece before diffusion, the direction of the diffusion and the nature of the impurities present are also defined. The coefficient of diffusion D is then given by the equation D ({pi}/4t){rho}{sup 2}{sub 0} where t is the duration of the diffusion; {rho}{sub 0} is the extrapolated value of {rho} = ({delta}m)/({delta}C) for a zero value of the variation of concentration ({delta}m in is the change in weight of the test piece per unit surface; {delta}C is the difference between the concentration at the surface and the initial concentration of the test piece). This method has been used to study the effect of the direction of the diffusion on the coefficient of diffusion. The coefficient for diffusion which decreases the concentration of zinc is 5 times greater than that for diffusion which increases the quantity of zinc in the metal; an interpretation of this phenomena based on the mechanism of diffusion vacancies in the structure has been proposed. By means of micrographic investigation and by weighing it has been shown that the presence of certain impurities, such as phosphorous, arsenic, antimony, silicon, and aluminium can result in a marked increase of the rate of diffusion: the effect of these impurities on the coefficient of diffusion has been related to their valency and atomic weight. (author) [French] La diffusion dans les laitons {alpha} a ete etudiee au moyen des methodes d'evaporation et de condensafion du zinc. Apres avoir montre qu'aux temperatures suffisamment elevees, la diffusion intergranulaire ne jouait aucun role, l'auteur a prouve que la vitesse d'evaporation ou de condensation n'est definie que dans la mesure ou

  3. 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

  4. 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

  5. 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

  6. Space synthesis: an application of synthesis method to two and three dimensional multigroup neutron diffusion equations; Synthese spatiale: une application de la methode de synthese aux equations de diffusion neutronique multigroupe a deux et trois dimensions

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    In order to reduce computing time, two and three-dimensional multigroup neutron diffusion equations in cylindrical, rectangular (X, Y), (X, Y, Z) and hexagonal geometries are solved by the method of synthesis using an appropriate variational principle (stationary principle). The basic idea is to reduce the number of independent variables by constructing two or three-dimensional solutions from solutions of fewer variables, hence the name 'synthesis method'. Whatever the geometry, we are led to solve a system of ordinary differential equations with matrix coefficients to which one can apply well-known numerical methods: CHEBYSHEV's polynomial method, Gaussian elimination. Numerical results furnished by synthesis programs written for the IBM 7094, the IBM 360-75 and the CDC 6600 computers, are confronted with those which are given by programs employing the classical finite difference method. [French] En vue de reduire le-temps de calcul, les equations de diffusion neutronique, multigroupe, a deux et trois dimensions d'espace dans les geometries cylindrique, rectangulaire (X, Y), (X, Y, Z) et hexagonale sont resolues par la methode de synthese utilisant un principe variationnel approprie (principe stationnaire). L'idee consiste a reduire le nombre de variables independantes par construction d'une solution bi ou tridimensionnelle au moyen de solutions dependant d'un nombre inferieur de variables, d'ou le nom de la methode. Dans tous les cas de geometrie, nous sommes conduits a resoudre un systeme d'equations differentielles a coefficients matriciels auquel peuvent s'appliquer les methodes numeriques courantes; methode polynomiale de TCHEBYCHEFF et methode d'elimination de GAUSS. Les resultats numeriques obtenus par nos codes de synthese programmes sur IBM 7094, IBM 360-75 et CDC 6600, sont confrontes avec ceux que fournissent les programmes adoptant la methode classique des differences finies. (auteur)

  7. 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

  8. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  9. 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)

  10. 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.

  11. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    International Nuclear Information System (INIS)

    Fanaro, L.C.C.B.

    1984-01-01

    It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

  12. LA DIFFUSION DES INNOVATIONS MANAGERIALES EN CONTROLE DE GESTION, UNE APPROCHE RHETORIQUE

    OpenAIRE

    Eric Maton; Dragos Zelinschi

    2012-01-01

    International audience; The purpose of this paper is to explain how a rhetoric approach with several dimensions can help to better understand the diffusion of management control innovations. We use a dynamic conception of rhetoric and a rhetorical analysis of graphical representations. We apply a rhetoric framework of analysis to two innovations (activity-based costing and balanced scorecard) and to their description in one review, the Harvard Business Review, with an objective of a wide diff...

  13. 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

  14. Numerical solution of the equation of neutrons transport on plane geometry by analytical schemes using acceleration by synthetic diffusion

    International Nuclear Information System (INIS)

    Alonso-Vargas, G.

    1991-01-01

    A computer program has been developed which uses a technique of synthetic acceleration by diffusion by analytical schemes. Both in the diffusion equation as in that of transport, analytical schemes were used which allowed a substantial time saving in the number of iterations required by source iteration method to obtain the K e ff. The program developed ASD (Synthetic Diffusion Acceleration) by diffusion was written in FORTRAN and can be executed on a personal computer with a hard disc and mathematical O-processor. The program is unlimited as to the number of regions and energy groups. The results obtained by the ASD program for K e ff is nearly completely concordant with those of obtained utilizing the ANISN-PC code for different analytical type problems in this work. The ASD program allowed obtention of an approximate solution of the neutron transport equation with a relatively low number of internal reiterations with good precision. One of its applications would be in the direct determinations of axial distribution neutronic flow in a fuel assembly as well as in the obtention of the effective multiplication factor. (Author)

  15. 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

  16. 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)

  17. Iterative schemes for obtaining dominant alpha-modes of the neutron diffusion equation

    International Nuclear Information System (INIS)

    Singh, K.P.; Modak, R.S.; Degweker, S.B.; Singh, Kanchhi

    2009-01-01

    Two new methods of obtaining dominant prompt alpha-modes (sometimes referred to as time-eigenfunctions) of the multigroup neutron diffusion equation are discussed. In the first of these, we initially compute the dominant K-eigenfunctions and K-eigenvalues (denoted by λ 1 , λ 2 , λ 3 ...λ 1 being equal to the K eff ) for the given nuclear reactor model, by existing method based on sub-space iteration (SSI) which is an improved version of power iteration method. Subsequently, a uniformly distributed (positive or negative) 1/v absorber of sufficient concentration is added so as to make a particular eigenvalue λ i equal to unity. This gives ith alpha-mode. This procedure is repeated to find all the required alpha-modes. In the second method, we solve the alpha-eigenvalue problem directly by SSI method. This is clearly possible for a sub-critical reactor for which the inverse of the dominant alpha-eigenvalues are also the largest in magnitude as required by the SSI method. Here, the procedure is made applicable even to a super-critical reactor by making the reactor model sub-critical by the addition of a 1/v absorber. Results of these calculations for a 3-D two group PHWR test-case are given. These results are validated against the results as obtained by a completely different approach based on Orthomin(1) algorithm published earlier. The direct method based on the sub-space iteration strategy is found to be a simple and reliable method for obtaining any number of alpha-modes. Also comments have been made on the relationship between fundamental α and k values.

  18. Two-dimensional finite element neutron diffusion analysis using hierarchic shape functions

    International Nuclear Information System (INIS)

    Carpenter, D.C.

    1997-01-01

    Recent advances have been made in the use of p-type finite element method (FEM) for structural and fluid dynamics problems that hold promise for reactor physics problems. These advances include using hierarchic shape functions, element-by-element iterative solvers and more powerful mapping techniques. Use of the hierarchic shape functions allows greater flexibility and efficiency in implementing energy-dependent flux expansions and incorporating localized refinement of the solution space. The irregular matrices generated by the p-type FEM can be solved efficiently using element-by-element conjugate gradient iterative solvers. These solvers do not require storage of either the global or local stiffness matrices and can be highly vectorized. Mapping techniques based on blending function interpolation allow exact representation of curved boundaries using coarse element grids. These features were implemented in a developmental two-dimensional neutron diffusion program based on the use of hierarchic shape functions (FEM2DH). Several aspects in the effective use of p-type analysis were explored. Two choices of elemental preconditioning were examined--the proper selection of the polynomial shape functions and the proper number of functions to use. Of the five shape function polynomials tested, the integral Legendre functions were the most effective. The serendipity set of functions is preferable over the full tensor product set. Two global preconditioners were also examined--simple diagonal and incomplete Cholesky. The full effectiveness of the finite element methodology was demonstrated on a two-region, two-group cylindrical problem but solved in the x-y coordinate space, using a non-structured element grid. The exact, analytic eigenvalue solution was achieved with FEM2DH using various combinations of element grids and flux expansions

  19. Vectorized and multitasked solution of the few-group neutron diffusion equations

    International Nuclear Information System (INIS)

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

    1989-01-01

    A numerical algorithm with parallelism was used to solve the two-group, multidimensional neutron diffusion equations on computers characterized by shared memory, vector pipeline, and multi-CPU architecture features. Specifically, solutions were obtained on the Cray X/MP-48, the IBM-3090 with vector facilities, and the FPS-164. The material-centered mesh finite difference method approximation and outer-inner iteration method were employed. Parallelism was introduced in the inner iterations using the cyclic line successive overrelaxation iterative method and solving in parallel across lines. The outer iterations were completed using the Chebyshev semi-iterative method that allows parallelism to be introduced in both space and energy groups. For the three-dimensional model, power, soluble boron, and transient fission product feedbacks were included. Concentrating on the pressurized water reactor (PWR), the thermal-hydraulic calculation of moderator density assumed single-phase flow and a closed flow channel, allowing parallelism to be introduced in the solution across the radial plane. Using a pinwise detail, quarter-core model of a typical PWR in cycle 1, for the two-dimensional model without feedback the measured million floating point operations per second (MFLOPS)/vector speedups were 83/11.7. 18/2.2, and 2.4/5.6 on the Cray, IBM, and FPS without multitasking, respectively. Lower performance was observed with a coarser mesh, i.e., shorter vector length, due to vector pipeline start-up. For an 18 x 18 x 30 (x-y-z) three-dimensional model with feedback of the same core, MFLOPS/vector speedups of --61/6.7 and an execution time of 0.8 CPU seconds on the Cray without multitasking were measured. Finally, using two CPUs and the vector pipelines of the Cray, a multitasking efficiency of 81% was noted for the three-dimensional model

  20. 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

  1. 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)

  2. 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

  3. Sensitivity analysis of the Galerkin finite element method neutron diffusion solver to the shape of the elements

    Energy Technology Data Exchange (ETDEWEB)

    Hosseini, Seyed Abolfaz [Dept. of Energy Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of)

    2017-02-15

    The purpose of the present study is the presentation of the appropriate element and shape function in the solution of the neutron diffusion equation in two-dimensional (2D) geometries. To this end, the multigroup neutron diffusion equation is solved using the Galerkin finite element method in both rectangular and hexagonal reactor cores. The spatial discretization of the equation is performed using unstructured triangular and quadrilateral finite elements. Calculations are performed using both linear and quadratic approximations of shape function in the Galerkin finite element method, based on which results are compared. Using the power iteration method, the neutron flux distributions with the corresponding eigenvalue are obtained. The results are then validated against the valid results for IAEA-2D and BIBLIS-2D benchmark problems. To investigate the dependency of the results to the type and number of the elements, and shape function order, a sensitivity analysis of the calculations to the mentioned parameters is performed. It is shown that the triangular elements and second order of the shape function in each element give the best results in comparison to the other states.

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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$^{o}$

  10. 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)

  11. Two-dimensional geometrical corner singularities in neutron diffusion. Part 2: Application to the SNR-300 benchmark

    International Nuclear Information System (INIS)

    Cacuci, D.G.; Univ. of Karlsruhe; Kiefhaber, E.; Stehle, B.

    1998-01-01

    The explicit solution developed by Cacuci for the multigroup neutron diffusion equation at interior corners in two-dimensional two-region domains has been applied to the SNR-300 fast reactor prototype design to obtain the exact behavior of the multigroup fluxes at and around typical corners arising between absorber/fuel and follower/fuel assemblies. The calculations have been performed in hexagonal geometry using four energy groups, and the results clearly show that the multigroup fluxes are finite but not analytical at interior corners. In particular, already the first-order spatial derivatives of the multigroup fluxes become unbounded at the corners between follower and fuel assemblies. These results highlight the need to treat properly the influence of corners, both for the direct calculation and for the reconstruction of pointwise neutron flux and power distributions in heterogeneous reactor cores

  12. Solution of the neutron diffusion equation to study the 3D distribution of power, applied to nuclear reactors

    International Nuclear Information System (INIS)

    Costa, Danilo Leite

    2013-01-01

    This work aims to present a study about the power distribution behavior in a PWR type reactor, considering both intensity and migration of power peaks due to insertion of control rods into the core. Employing the multidimensional steady-state neutron diffusion equation in order to simulate the neutron flux, and using the Finite Difference Method. Furthermore, based on the axial power distribution on the largest heat flux rod, is carried out thermal analysis of this rod and associated coolant channel. For this purpose is employed the FueLRod 3 D code, it uses the Finite Element Method to model the fuel rod and the associated coolant channel, allowing the thermohydraulics simulation of a single rod discretized in three dimensions, considering the heat flux from the pellet, crossing the gap and the cladding until it reaches the coolant. (author)

  13. Investigation of the response of a neutron moisture meter using a multigroup, two-dimensional diffusion theory code

    International Nuclear Information System (INIS)

    Ritchie, A.I.M.; Wilson, D.J.

    1984-12-01

    A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated

  14. Scintillation Counters for Neutron Scattering Experiments; Compteurs a scintillations pour les experiences de diffusion neutronique; Stsintillyatsionnye schetchiki dlya ehksperimentov s rasseyaniem nejtronov; Contadores de centelleo para experimentos de dispersion neutronica.

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D; Duffil, C [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Wraight, L A [Aere, Harwell, Didcot, Berks (United Kingdom)

    1963-01-15

    Scintillation counters discussed in this paper are of two types : (i) UP or B{sub 2}O{sub 3} fixed with zinc sulphide in varving composition, (ii) Li containing glasses of varying composition. The ideal composition of each rype for neutron scattering experiments and their relative sensitivity tae neutrons and {gamma}-rays are considered. The ZnS type can be used with a {gamma}-ray pulse shape discriminator and practical experience of its use in a multi-counter time-of-flight experiment is described. The Li glass has higher {gamma}-background but also higher neutron efficiency. Performance figures of a scintillator containing 25% by wt. Li{sub 2}O and 1 mm thick will be given. (author) [French] Les compteurs a scintillations etudies sont de deux types : i) LiF ou B{sub 2}O{sub 3} fixe par du sulfure de zinc en composition variable, ii) verres contenant du lithium en composition variable. Les auteurs examinent la composition ideale de chaque type de compteur pour les experiences de diffusion de neutrons, ainsi que leur sinsibilite rrelative aux neutrons et aux rayons gamma. On peut utiliser le compteur a ZnS avec un discriminates de forme pour rayons gamma; les auteurs decrivent l'experience qu'ils ont de son emploi dans une experience de temps de vol a plusieurs compteurs. Le compteur verre-Li a un mouvement propre plus eleve pour les rayons gamma mais une meilleure efficacite vis-a-vis des neutrons. Les auteurs donnent quelques chiffres concernant le fonctionnement d'un scintillateur contenant 25% en poids de Li{sub 2}O et ayant une epaisseur de un millimetre. (author) [Spanish] Los contadores de centelleo examinados en esta memoria son de dos tipos : a) de LiF o B{sub 2}O{sub 3} mezclado con sulfuro de cinc en proporciones variables; b) de vidrios litiados de diversas composiciones. Los autores estudian la composicion ideal de cada tipo para experimentos de dispersion neutronica, asi como su sensibilidad relativa a los neutrones y a los rayos gamma. El tio de Zn

  15. Mass measurement project by determination of Q{sub {beta}} for neutron-rich nuclei; Projet de mesure des masses par determination des Q{sub {beta}} pour des noyaux tres riches en neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Pautrat, M; Lagrange, J M; Petizon, L [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire; Vanhorenbeeck, J; Duhamel, P [Brussels Univ. (Belgium). Inst. d` Astronomie et d` Astrophysique; Binon, F [Universite Libre de Bruxelles (Belgium)

    1994-12-31

    The aim of the project described hereafter is to collect new data on the exotic neutron rich nuclei of the Fe to Zn region, and in particular to determine their masses, for both nuclear physics and astrophysics purposes. These isotopes will be produced through projectile fragmentation at the GANIL facility and selected by the LISE3 spectrometer. Their half-lives will be measured as well as the energy of their main {gamma} rays; {gamma} - {gamma} coincidences will then allow to build a preliminary level scheme. The analysis of {beta} spectra and {beta} - {gamma} coincidences will finally provide the maximum {beta} decay energies of the studied nuclei leading to their masses. The difficulties arising from the low production rates, the {beta} detection, the data handling are discussed together with the solutions proposed to overcome them. (authors). 17 refs.

  16. FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix

    International Nuclear Information System (INIS)

    Misfeldt, I.B.

    1987-01-01

    1 - Nature of physical problem solved: The two-dimensional neutron diffusion equation (xy geometry) is solved in the homogeneous form (K eff calculation). The boundary conditions specify each group current as a linear homogeneous function of the group fluxes (gamma matrix concept). For each material, the fission matrix is assumed to be dyadic. 2 - Method of solution: Finite element formulation with Lagrange type elements. Solution technique: SOR with extrapolation. 3 - Restrictions on the complexity of the problem: Maximum order of the Lagrange elements is 6

  17. Intramolecular diffusive motion in alkane monolayers studied by high-resolution quasielastic neutron scattering and molecular dynamics simulations

    DEFF Research Database (Denmark)

    Hansen, Flemming Yssing; Criswell, L.; Fuhrmann, D

    2004-01-01

    Molecular dynamics simulations of a tetracosane (n-C24H50) monolayer adsorbed on a graphite basal-plane surface show that there are diffusive motions associated with the creation and annihilation of gauche defects occurring on a time scale of similar to0.1-4 ns. We present evidence...... that these relatively slow motions are observable by high-energy-resolution quasielastic neutron scattering (QNS) thus demonstrating QNS as a technique, complementary to nuclear magnetic resonance, for studying conformational dynamics on a nanosecond time scale in molecular monolayers....

  18. 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

  19. Neutron Diffusion in a Space Lattice of Fissionable and Absorbing Materials

    Science.gov (United States)

    Feynman, R. P.; Welton, T. A.

    1946-08-27

    Methods are developed for estimating the effect on a critical assembly of fabricating it as a lattice rather than in the more simply interpreted homogeneous manner. An idealized case is discussed supposing an infinite medium in which fission, elastic scattering and absorption can occur, neutrons of only one velocity present, and the neutron m.f.p. independent of position and equal to unity with the unit of length used.

  20. 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

  1. Boron Neutron Capture Therapy activity of diffused tumors at TRIGA Mark II in Pavia

    Energy Technology Data Exchange (ETDEWEB)

    Bortolussi, S.; Stella, S.; De Bari, A.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy)]|[National Institute of Nuclear Physics (INFN), Pavia (Italy); Bruschi, P.; Bakeine, J.G. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Clerici, A.; Ferrari, C.; Zonta, C.; Zonta, A. [Department of Surgery, University of Pavia, Pavia (Italy); Nano, R. [Department of Animal Biology, University of Pavia, Pavia (Italy)

    2008-10-29

    The Boron neutron Capture Therapy research in Pavia has a long tradition: it begun more than 20 years ago at the TRIGA Mark II reactor of the University. A technique for the treatment of the hepatic metastases was developed, consisting in explanting the liver treated with {sup 10}B, irradiating it in the thermal column of the reactor, and re-implanting the organ in the patient. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research will be presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,{alpha}) spectroscopy and neutron autoradiography. (authors)

  2. Boron Neutron Capture Therapy activity of diffused tumors at TRIGA Mark II in Pavia

    International Nuclear Information System (INIS)

    Bortolussi, S.; Stella, S.; De Bari, A.; Altieri, S.; Bruschi, P.; Bakeine, J.G.; Clerici, A.; Ferrari, C.; Zonta, C.; Zonta, A.; Nano, R.

    2008-01-01

    The Boron neutron Capture Therapy research in Pavia has a long tradition: it begun more than 20 years ago at the TRIGA Mark II reactor of the University. A technique for the treatment of the hepatic metastases was developed, consisting in explanting the liver treated with 10 B, irradiating it in the thermal column of the reactor, and re-implanting the organ in the patient. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research will be presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,α) spectroscopy and neutron autoradiography. (authors)

  3. 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)

  4. 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

  5. Parallel computing for homogeneous diffusion and transport equations in neutronics; Calcul parallele pour les equations de diffusion et de transport homogenes en neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Pinchedez, K

    1999-06-01

    Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)

  6. The application of isogeometric analysis to the neutron diffusion equation for a pincell problem with an analytic benchmark

    International Nuclear Information System (INIS)

    Hall, S.K.; Eaton, M.D.; Williams, M.M.R.

    2012-01-01

    Highlights: ► Isogeometric analysis used to obtain solutions to the neutron diffusion equation. ► Exact geometry captured for a circular fuel pin within a square moderator. ► Comparisons are made between the finite element method and isogeometric analysis. ► Error and observed order of convergence found using an analytic solution. -- Abstract: In this paper the neutron diffusion equation is solved using Isogeometric Analysis (IGA), which is an attempt to generalise Finite Element Analysis (FEA) to include exact geometries. In contrast to FEA, the basis functions are rational functions instead of polynomials. These rational functions, called non-uniform rational B-splines, are used to capture both the geometry and approximate the solution. The method of manufactured solutions is used to verify a MatLab implementation of IGA, which is then applied to a pincell problem. This is a circular uranium fuel pin within a square block of graphite moderator. A new method is used to compute an analytic solution to a simplified version of this problem, and is then used to observe the order of convergence of the numerical scheme. Comparisons are made against quadratic finite elements for the pincell problem, and it is found that the disadvantage factor computed using IGA is less accurate. This is due to a cancellation of errors in the FEA solution. A modified pincell problem with vacuum boundary conditions is then considered. IGA is shown to outperform FEA in this situation.

  7. SYNTH-C, Steady-State and Time-Dependent 3-D Neutron Diffusion with Thermohydraulic Feedback

    Energy Technology Data Exchange (ETDEWEB)

    Brega, E [ENEL-CRTN, Bastioni di Porta Volta 10, Milan (Italy); Salina, E [A.R.S. Spa, Viale Maino 35, Milan (Italy)

    1980-04-01

    1 - Description of problem or function: SYNTH-C-STEADY and SYNTH-C- TRANS solve respectively the steady-state and time-dependent few- group neutron diffusion equations in three dimensions x,y,z in the presence of fuel temperature and thermal-hydraulic feedback. The neutron diffusion and delayed precursor equations are approximated by a space-time (z,t) synthesis method with axially discontinuous trial functions. Three thermal-hydraulic and fuel heat transfer models are available viz. COBRA-3C/MIT model, lumped parameter (WIGL) model and adiabatic fuel heat-up model. 2 - Method of solution: The steady-state and time-dependent synthesis equations are solved respectively by the Wielandt's power method and by the theta-difference method (in time), both coupled with a block factorization technique and double precision arithmetic. The thermal-hydraulic model equations are solved by fully implicit finite differences (WIGL) or explicit-implicit difference techniques with iterations (COBRA-EC/MIT). 3 - Restrictions on the complexity of the problem: Except for the few- group limitation, the programs have no other fixed limitation so the ability to run a problem depends only on the available computer storage.

  8. Simulation and optimization of a new focusing polarizing bender for the diffuse neutrons scattering spectrometer DNS at MLZ

    International Nuclear Information System (INIS)

    Nemkovski, K; Ioffe, A; Su, Y; Babcock, E; Schweika, W; Brückel, Th

    2017-01-01

    We present the concept and the results of the simulations of a new polarizer for the diffuse neutron scattering spectrometer DNS at MLZ. The concept of the polarizer is based on the idea of a bender made from the stack of the silicon w