Pellet by pellet neutron flux calculations coupled with nodal expansion method

International Nuclear Information System (INIS)

Aldo, Dall'Osso

2003-01-01

We present a technique whose aim is to replace 2-dimensional pin by pin de-homogenization, currently done in core reactor calculations with the nodal expansion method (NEM), by a 3-dimensional finite difference diffusion calculation. This fine calculation is performed as a zoom in each node taking as boundary conditions the results of the NEM calculations. The size of fine mesh is of the order of a fuel pellet. The coupling between fine and NEM calculations is realised by an albedo like boundary condition. Some examples are presented showing fine neutron flux shape near control rods or assembly grids. Other fine flux behaviour as the thermal flux rise in the fuel near the reflector is emphasised. In general the results show the interest of the method in conditions where the separability of radial and axial directions is not granted. (author)

A new calculation of atmospheric neutrino flux: the FLUKA approach

International Nuclear Information System (INIS)

Battistoni, G.; Bloise, C.; Cavalli, D.; Ferrari, A.; Montaruli, T.; Rancati, T.; Resconi, S.; Ronga, F.; Sala, P.R.

1999-01-01

Preliminary results from a full 3-D calculation of atmospheric neutrino fluxes using the FLUKA interaction model are presented and compared to previous existing calculations. This effort is motivated mainly by the 3-D capability and the satisfactory degree of accuracy of the hadron-nucleus models embedded in the FLUKA code. Here we show examples of benchmarking tests of the model with cosmic ray experiment results. A comparison of our calculation of the atmospheric neutrino flux with that of the Bartol group, for E ν > 1 GeV, is presented

Calculation of conventional and prompt lepton fluxes at very high energy

CERN Document Server

Fedynitch, Anatoli; Gaisser, Thomas K; Riehn, Felix; Stanev, Todor

2015-01-01

An efficient method for calculating inclusive conventional and prompt atmospheric leptons fluxes is presented. The coupled cascade equations are solved numerically by formulating them as matrix equation. The presented approach is very flexible and allows the use of different hadronic interaction models, realistic parametrizations of the primary cosmic-ray flux and the Earth's atmosphere, and a detailed treatment of particle interactions and decays. The power of the developed method is illustrated by calculating lepton flux predictions for a number of different scenarios.

3-D flux distribution and criticality calculation of TRIGA Mark-II

International Nuclear Information System (INIS)

Can, B.

1982-01-01

In this work, the static calculation of the (I.T.U. TRIGA Mark-II) flux distribution has been made. The three dimensional, r-θ-z, representation of the core has been used. In this representation, for different configuration, the flux distribution has been calculated depending on two group theory. The thermal-hydraulics, the poisoning effects have been ignored. The calculations have been made by using the three dimensional and multigroup code CAN. (author)

Calculation of neutron flux in the presence of a source

International Nuclear Information System (INIS)

Planchard, J.

1993-09-01

Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs

Calculation of conventional and prompt lepton fluxes at very high energy

Directory of Open Access Journals (Sweden)

Fedynitch Anatoli

2015-01-01

Full Text Available An efficient method for calculating inclusive conventional and prompt atmospheric leptons fluxes is presented. The coupled cascade equations are solved numerically by formulating them as matrix equation. The presented approach is very flexible and allows the use of different hadronic interaction models, realistic parametrizations of the primary cosmic-ray flux and the Earth's atmosphere, and a detailed treatment of particle interactions and decays. The power of the developed method is illustrated by calculating lepton flux predictions for a number of different scenarios.

Câbles electriques - Calcul du courant admissible - Partie 1: Equations de l'intensité du courant admissible (facteur de charge 100%) et calcul des pertes - Section 2: Facteurs de pertes par courants de Foucault dans les gaines dans le cas de deux circuits disposés en nappe

CERN Document Server

International Electrotechnical Commission. Geneva

1993-01-01

Câbles electriques - Calcul du courant admissible - Partie 1: Equations de l'intensité du courant admissible (facteur de charge 100%) et calcul des pertes - Section 2: Facteurs de pertes par courants de Foucault dans les gaines dans le cas de deux circuits disposés en nappe

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

Energy Technology Data Exchange (ETDEWEB)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

International Nuclear Information System (INIS)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-07-01

This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-06-15

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

A finite element calculation of flux pumping

Science.gov (United States)

Campbell, A. M.

2017-12-01

A flux pump is not only a fascinating example of the power of Faraday’s concept of flux lines, but also an attractive way of powering superconducting magnets without large electronic power supplies. However it is not possible to do this in HTS by driving a part of the superconductor normal, it must be done by exceeding the local critical density. The picture of a magnet pulling flux lines through the material is attractive, but as there is no direct contact between flux lines in the magnet and vortices, unless the gap between them is comparable to the coherence length, the process must be explicable in terms of classical electromagnetism and a nonlinear V-I characteristic. In this paper a simple 2D model of a flux pump is used to determine the pumping behaviour from first principles and the geometry. It is analysed with finite element software using the A formulation and FlexPDE. A thin magnet is passed across one or more superconductors connected to a load, which is a large rectangular loop. This means that the self and mutual inductances can be calculated explicitly. A wide strip, a narrow strip and two conductors are considered. Also an analytic circuit model is analysed. In all cases the critical state model is used, so the flux flow resistivity and dynamic resistivity are not directly involved, although an effective resistivity appears when J c is exceeded. In most of the cases considered here is a large gap between the theory and the experiments. In particular the maximum flux transferred to the load area is always less than the flux of the magnet. Also once the threshold needed for pumping is exceeded the flux in the load saturates within a few cycles. However the analytic circuit model allows a simple modification to allow for the large reduction in I c when the magnet is over a conductor. This not only changes the direction of the pumped flux but leads to much more effective pumping.

COSANI-2, Gamma Doses from SABINE Calculation, Activity from ANISN Flux Calculation

International Nuclear Information System (INIS)

Dupont, C.

1975-01-01

1 - Nature of physical problem solved: Retrieval of SABINE and/or ANISN results. Calculates in case of SABINE results the individual contributions of capture gamma rays in each region to the total gamma dose and to the total gamma heating may calculate in case of ANISN new activity rates starting from ANISN flux saved on tape and activity cross sections taken on an ANISN binary library tape. The program can draw on a BENSON plotter any of the following quantities: - group flux; - activity rates; - dose rates; - neutron spectra for SABINE; - neutron or gamma direct or adjoint spectra for ANISN; - gamma heating and dose rate for SABINE including individual contributions from each region. Several ANISN and/or SABINE cases can be drawn on the same graph for comparison purposes. 2 - Restrictions on the complexity of the problem: Maximum number of: - tapes containing ANISN and/or SABINE results: 5; - curves per graph: 3; - regions: 40; - points per curve: 500; - energy groups: 200

Radiative flux calculations at UV and visible wavelengths

International Nuclear Information System (INIS)

Grossman, A.S.; Grant, K.E.; Wuebbles, D.J.

1993-10-01

A radiative transfer model to calculate the short wavelength fluxes at altitudes between 0 and 80 km has been developed at LLNL. The wavelength range extends from 175--735 nm. This spectral range covers the UV-B wavelength region, 250--350 nm, with sufficient resolution to allow comparison of UV-B measurements with theoretical predictions. Validation studies for the model have been made for both UV-B ground radiation calculations and tropospheric solar radiative forcing calculations for various ozone distributions. These studies indicate that the model produces results which agree well with respect to existing UV calculations from other published models

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

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

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

International Nuclear Information System (INIS)

Kurosawa, M.

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

Science.gov (United States)

Kurosawa, Masahiko

2005-01-01

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

26 Calcul multi-caractéristique du coût du non-qualité via la fonction ...

African Journals Online (AJOL)

PR BOKO

Le coût du non - qualité (CNQ) est un indicateur permettant l'estimation de la marge ..... multicritères de Taguchi en fonction de l'évolution du poids et capacité du ... A Scheduling Example; International Journal of Information and Management.

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

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

Error estimates for ice discharge calculated using the flux gate approach

Science.gov (United States)

Navarro, F. J.; Sánchez Gámez, P.

2017-12-01

Ice discharge to the ocean is usually estimated using the flux gate approach, in which ice flux is calculated through predefined flux gates close to the marine glacier front. However, published results usually lack a proper error estimate. In the flux calculation, both errors in cross-sectional area and errors in velocity are relevant. While for estimating the errors in velocity there are well-established procedures, the calculation of the error in the cross-sectional area requires the availability of ground penetrating radar (GPR) profiles transverse to the ice-flow direction. In this contribution, we use IceBridge operation GPR profiles collected in Ellesmere and Devon Islands, Nunavut, Canada, to compare the cross-sectional areas estimated using various approaches with the cross-sections estimated from GPR ice-thickness data. These error estimates are combined with those for ice-velocities calculated from Sentinel-1 SAR data, to get the error in ice discharge. Our preliminary results suggest, regarding area, that the parabolic cross-section approaches perform better than the quartic ones, which tend to overestimate the cross-sectional area for flight lines close to the central flowline. Furthermore, the results show that regional ice-discharge estimates made using parabolic approaches provide reasonable results, but estimates for individual glaciers can have large errors, up to 20% in cross-sectional area.

Etat Du Magnesium Dans Quelques Sols Sales Du Sud Et Du ...

African Journals Online (AJOL)

étude a été réalisée sur 86 échantillons de sols provenant du Centre et du Sud de l´Irak. L\\'expérimentation a consisté, d´une part, à calculer les différents sels dominants dans les sols salés par la méthode de combinaison hypothétique et, ...

Chain Rule Approach for Calculating the Time-Derivative of Flux

Energy Technology Data Exchange (ETDEWEB)

Langenbrunner, James R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Booker, Jane M. [Booker Scientific, Fredericksburg, TX (United States)

2017-10-03

The reaction history (gamma-flux observable) is mathematically studied by using the chain rule for taking the total-time derivatives. That is, the total time-derivative of flux is written as the product of the ion temperature derivative with respect to time and the derivative of the flux with respect to ion temperature. Some equations are derived using the further simplification that the fusion reactivity is a parametrized function of ion temperature, T. Deuterium-tritium (D-T) fusion is used as the application with reactivity calculations from three established reactivity parametrizations.

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

Calcul du flux thermique a travers la liaison batiment- Sol | Diao ...

African Journals Online (AJOL)

In this work, we propose analytical models in two dimensions (2D) for the calculation of the thermal transfers through the connection building and ground in steady, witch we determine the linear thermal coefficient, and dynamic method. The methods used hold parameters geometrical and thermal having a great influence on ...

MURALB - a programme for calculating neutron fluxes in many groups

International Nuclear Information System (INIS)

MacDougall, J.

1977-09-01

The program MURALB solves the multi-group transport equation (with no upscatter) in many equal lethargy groups to produce neutron fluxes in these groups. The code has been made very flexible by confining the spatial flux solution to a single subroutine which takes as input the cross section data and source for a single group and calculates the flux for that group. In this way by supplying different versions of this routine different geometries and methods of solution of the transport equation may be treated. At present plane, cylindrical and spherical diffusion theory and collision probability solutions are available, together with a two region collision probability solution for a rod in a square cell. There is no basic restriction to one dimension but the practical size of problem tends to be limited to about 30 spatial regions by core storage requirements. In addition to the flux solution, the code calculates neutron balance, reaction rates and few groups cross sections for each mesh region, together with the values averaged over the system (cell or reactor). The program is available both as a stand-alone code and integrated into the COSMOS system. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

Calculation of the thermal neutron flux depression in the loop VISA-1

International Nuclear Information System (INIS)

Martinc, R.

1961-01-01

Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well [sr

Guide du calcul avec les logiciels libres XCAS, Scilab, Bc, Gp, GnuPlot, Maxima, MuPAD

CERN Document Server

Connan, Guillaume

2008-01-01

Le Guide du Calcul avec les logiciels libres vous propose de maîtriser les bases de plusieurs logiciels gratuits de calcul formel et numérique, et de géométrie dynamique : XCAS, Scilab, Bc, Gp, GnuPlot, Maxima, Octave, Yacas, et également MuPAD (non libre). Ces logiciels sont multiplateformes (Windows, Mac, Linux) et s'installent rapidement. La grande originalité de cet ouvrage est de détailler pas à pas l'utilisation de plusieurs logiciels pour plusieurs types de calcul. Il propose ainsi des applications en mathématiques, en physique et en sciences de l'ingénieur (par exemple : équations différentielles utilisées en physique, traitement du son...). Il vous aidera à savoir utiliser le bon logiciel au bon moment pour résoudre le bon problème. Cet ouvrage s'adresse à toute personne voulant utiliser le calcul en mathématiques à l'aide de logiciels libres : étudiants en mathématiques ou en informatique, professeurs de lycée souhaitant utiliser ces outils dans le cadre de leur cours et jusqu'...

Calculation of atmospheric neutrino flux using the interaction model calibrated with atmospheric muon data

International Nuclear Information System (INIS)

Honda, M.; Kajita, T.; Kasahara, K.; Midorikawa, S.; Sanuki, T.

2007-01-01

Using the 'modified DPMJET-III' model explained in the previous paper [T. Sanuki et al., preceding Article, Phys. Rev. D 75, 043005 (2007).], we calculate the atmospheric neutrino flux. The calculation scheme is almost the same as HKKM04 [M. Honda, T. Kajita, K. Kasahara, and S. Midorikawa, Phys. Rev. D 70, 043008 (2004).], but the usage of the 'virtual detector' is improved to reduce the error due to it. Then we study the uncertainty of the calculated atmospheric neutrino flux summarizing the uncertainties of individual components of the simulation. The uncertainty of K-production in the interaction model is estimated using other interaction models: FLUKA'97 and FRITIOF 7.02, and modifying them so that they also reproduce the atmospheric muon flux data correctly. The uncertainties of the flux ratio and zenith angle dependence of the atmospheric neutrino flux are also studied

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-06-15

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

A diffusion-theoretical method to calculate the neutron flux distribution in multisphere configurations

International Nuclear Information System (INIS)

Schuerrer, F.

1980-01-01

For characterizing heterogene configurations of pebble-bed reactors the fine structure of the flux distribution as well as the determination of the macroscopic neutronphysical quantities are of interest. When calculating system parameters of Wigner-Seitz-cells the usual codes for neutron spectra calculation always neglect the modulation of the neutron flux by the influence of neighbouring spheres. To judge the error arising from that procedure it is necessary to determinate the flux distribution in the surrounding of a spherical fuel element. In the present paper an approximation method to calculate the flux distribution in the two-sphere model is developed. This method is based on the exactly solvable problem of the flux determination of a point source of neutrons in an infinite medium, which contains a spherical perturbation zone eccentric to the point source. An iteration method allows by superposing secondary fields and alternately satisfying the conditions of continuity on the surface of each of the two fuel elements to advance to continually improving approximations. (orig.) 891 RW/orig. 892 CKA [de

SPACETRAN, Radiation Leakage from Cylinder with ANISN Flux Calculation

International Nuclear Information System (INIS)

Cramer, S.N.; Solomito, M.

1974-01-01

1 - Nature of physical problem solved: SPACETRAN is designed to calculate the energy-dependent total flux or some proportional quantity such as kerma, due to the radiation leakage from the surface of a right-circular cylinder at detector positions located at arbitrary distances from the surface. The assumptions are made that the radiation emerging from the finite cylinder has no spatial dependence and that a vacuum surrounds the cylinder. 2 - Method of solution: There are three versions of the program in the code package. SPACETRAN-I uses the surface angular fluxes calculated by the discrete ordinates SN code ANISN, as input. SPACETRAN-II assumes that the surface angular flux for all energies can be represented as a function (Cos(PHI))**N, where PHI is the angle between surface outward normal and radiation direction, and N is an integer specified by the user. For both versions the energy group structure and the number and location of detectors is arbitrary. The flux (or response function) for a given energy group at some detection point is computed by summing the contributions from each surface area element over the entire surface. The surface area elements are defined by input data. SPACETRAN-III uses surface angular fluxes from DOT-3. SPACETRAN-I handles contributions either from a cylinder 'end' or 'side', so the total contributions must be obtained by adding the results of separate end and side runs. ANISN angular fluxes are specified for discrete directions. In general, the direction between the detector and contributing area will not exactly coincide with one of these discrete directions. In this case, the ANISN angular flux for the 'closest' discrete direction is used to approximate the contribution to the detector. SPACETRAN-II handles contributions from both the side and end of a cylinder in a single run. Since the assumed angular distribution is specified by a continuous function, it is not necessary to perform the angle selection described above. For

Interpretation of the measurement of ions fluxes through a biological membrane with a cellular compartment: example of the movements of sodium through the skin of frogs; Interpretation de la mesure des flux d'ions a travers une membrane biologique comportant un ''compartiment'' cellulaire; exemple des mouvements de sodium a travers la peau de grenouille

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

Two-way ion fluxes which can be measured in vitro through a living epithelial membrane (such as frog skin) by the indicator method take place across the cells which behave like an intermediate ionic 'compartment'. Two membranes and four fluxes have thus to be considered. Measurements in vitro of the total sodium fluxes as a function of the sodium concentration in the medium in contact with the external face of the skin have been interpreted in this spirit. Making use of certain hypotheses, the permeability coefficients for sodium of the two cellular membranes, the four sodium fluxes, the intracellular sodium concentration and the membrane potentials have been calculated for each value of the sodium concentration in the external medium. (author) [French] Les flux ioniques bidirectionnels que l'on peut mesurer in vitro a travers une membrane epitheliale vivante (comme la peau de grenouille) a l'aide de la methode des indicateurs, s'effectuent a travers les cellules qui se comportent comme un 'compartiment' ionique intermediaire. On doit donc considerer deux membranes et quatre flux. Des mesures in vitro des flux totaux de sodium en fonction de la concentration du sodium dans le milieu baignant la face externe de la peau ont ete interpretees dans cette perspective. Moyennant certaines hypotheses, les coefficients de permeabilite pour le sodium des deux membranes cellulaires, les quatre flux de sodium ainsi que la concentration du sodium intracellulaire et les potentiels de membrane ont pu etre calcules pour chaque valeur de la concentration du sodium dans le milieu externe. (auteur)

Spatial flux instabilities, and their control in the graphite gas power reactors; Les instabilites spatiales du flux et leur controle dans les reacteurs de puissance graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

Radial-azimuthal and axial spatial flux instabilities in graphite-gas reactors are studied by means of an analytical approach. Results are checked with those which are given by two dimensional (r, z and r, {theta}) kinetic models programmed for an IBM 7094 computer. At least, conclusions on the control of instabilities obtained from these models are reported. (author) [French] Les instabilites spatiales du flux dans les reacteurs graphite-gaz, radiales et azimutales d'une part, axiales d'autre part, sont etudiees au moyen d'une formulation analytique. Les resultats sont confrontes avec ceux que fournissent des modeles cinetiques a deux dimensions (r, z et r, {theta}) programmes sur IBM 7094. On donne enfin les conclusions relatives au controle de ces instabilites que ces modeles ont permis de degager. (auteur)

Exploring the use of a deterministic adjoint flux calculation in criticality Monte Carlo simulations

International Nuclear Information System (INIS)

Jinaphanh, A.; Miss, J.; Richet, Y.; Martin, N.; Hebert, A.

2011-01-01

The paper presents a preliminary study on the use of a deterministic adjoint flux calculation to improve source convergence issues by reducing the number of iterations needed to reach the converged distribution in criticality Monte Carlo calculations. Slow source convergence in Monte Carlo eigenvalue calculations may lead to underestimate the effective multiplication factor or reaction rates. The convergence speed depends on the initial distribution and the dominance ratio. We propose using an adjoint flux estimation to modify the transition kernel according to the Importance Sampling technique. This adjoint flux is also used as the initial guess of the first generation distribution for the Monte Carlo simulation. Calculated Variance of a local estimator of current is being checked. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

International Nuclear Information System (INIS)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

Energy Technology Data Exchange (ETDEWEB)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

An analytical transport theory method for calculating flux distribution in slab cells

International Nuclear Information System (INIS)

Abdel Krim, M.S.

2001-01-01

A transport theory method for calculating flux distributions in slab fuel cell is described. Two coupled integral equations for flux in fuel and moderator are obtained; assuming partial reflection at moderator external boundaries. Galerkin technique is used to solve these equations. Numerical results for average fluxes in fuel and moderator and the disadvantage factor are given. Comparison with exact numerical methods, that is for total reflection moderator outer boundaries, show that the Galerkin technique gives accurate results for the disadvantage factor and average fluxes. (orig.)

Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

International Nuclear Information System (INIS)

Kwon, S.G.; Lee, S.Y.; Yook, C.C.

1981-01-01

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

Lifetime measurement in Proserpine by reactivity modulation (1960); Mesure du temps de vie dans proserpine par modulation de la reactivite (1960)

Energy Technology Data Exchange (ETDEWEB)

Clouet d' Orval, C; Tachon, J; Bertrand, J; Lecoustey, P [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

The measurement method consists in varying the neutron flux periodically by means of an oscillator with cadmium sectors. From the signal received on a detector the lifetime {tau} can be determined; for various velocities {omega}, we have: (n/{delta}n){sup 2} = ({beta}{sub eff}/{delta}k){sup 2} + ({tau}/{delta}k){sup 2}{omega}{sup 2}. Various corrections are involved, in particular the calculation of the rates of different harmonics in the oscillator signal. (author) [French] La methode de mesure consiste a faire varier periodiquement le flux de neutrons au moyen d'un oscillateur a secteurs de cadmium. Le signal recu sur un detecteur permet la determination du temps de vie {tau}; pour differentes vitesses {omega}, on a: (n/{delta}n){sup 2} = ({beta}{sub eff}/{delta}k){sup 2} + ({tau}/{delta}k){sup 2}{omega}{sup 2}. Differentes corrections interviennent, en particulier, le calcul du taux des differents harmoniques dans le signal de l'oscillateur. (auteur)

Comparison of calculated energy flux of internal tides with microstructure measurements

Directory of Open Access Journals (Sweden)

Saeed Falahat

2014-10-01

Full Text Available Vertical mixing caused by breaking of internal tides plays a major role in maintaining the deep-ocean stratification. This study compares observations of dissipation from microstructure measurements to calculations of the vertical energy flux from barotropic to internal tides, taking into account the temporal variation due to the spring-neap tidal cycle. The dissipation data originate from two surveys in the Brazil Basin Tracer Release Experiment (BBTRE, and one over the LArval Dispersal along the Deep East Pacific Rise (LADDER3, supplemented with a few stations above the North-Atlantic Ridge (GRAVILUCK and in the western Pacific (IZU. A good correlation is found between logarithmic values of energy flux and local dissipation in BBTRE, suggesting that the theory is able to predict energy fluxes. For the LADDER3, the local dissipation is much smaller than the calculated energy flux, which is very likely due to the different topographic features of BBTRE and LADDER3. The East Pacific Rise consists of a few isolated seamounts, so that most of the internal wave energy can radiate away from the generation site, whereas the Brazil Basin is characterised by extended rough bathymetry, leading to a more local dissipation. The results from all four field surveys support the general conclusion that the fraction of the internal-tide energy flux that is dissipated locally is very different in different regions.

Improvement of low energy atmospheric neutrino flux calculation using the JAM nuclear interaction model

International Nuclear Information System (INIS)

Honda, M.; Kajita, T.; Kasahara, K.; Midorikawa, S.

2011-01-01

We present the calculation of the atmospheric neutrino fluxes with an interaction model named JAM, which is used in PHITS (Particle and Heavy-Ion Transport code System) [K. Niita et al., Radiation Measurements 41, 1080 (2006).]. The JAM interaction model agrees with the HARP experiment [H. Collaboration, Astropart. Phys. 30, 124 (2008).] a little better than DPMJET-III[S. Roesler, R. Engel, and J. Ranft, arXiv:hep-ph/0012252.]. After some modifications, it reproduces the muon flux below 1 GeV/c at balloon altitudes better than the modified DPMJET-III, which we used for the calculation of atmospheric neutrino flux in previous works [T. Sanuki, M. Honda, T. Kajita, K. Kasahara, and S. Midorikawa, Phys. Rev. D 75, 043005 (2007).][M. Honda, T. Kajita, K. Kasahara, S. Midorikawa, and T. Sanuki, Phys. Rev. D 75, 043006 (2007).]. Some improvements in the calculation of atmospheric neutrino flux are also reported.

Transport calculations of. gamma. -ray flux density and dose rate about implantable californium-252 sources

Energy Technology Data Exchange (ETDEWEB)

Shapiro, A; Lin, B I [Cincinnati Univ., Ohio (USA). Dept. of Chemical and Nuclear Engineering; Windham, J P; Kereiakes, J G

1976-07-01

..gamma.. flux density and dose rate distributions have been calculated about implantable californium-252 sources for an infinite tissue medium. Point source flux densities as a function of energy and position were obtained from a discrete-ordinates calculation, and the flux densities were multiplied by their corresponding kerma factors and added to obtain point source dose rates. The point dose rates were integrated over the line source to obtain line dose rates. Container attenuation was accounted for by evaluating the point dose rate as a function of platinum thickness. Both primary and secondary flux densities and dose rates are presented. The agreement with an independent Monte Carlo calculation was excellent. The data presented should be useful for the design of new source configurations.

Fusion neutron yield and flux calculation on HT-7 superconducting tokamak

International Nuclear Information System (INIS)

Fu Yanzhang; Zhu Yubao; Chen Juequan

2006-01-01

Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)

Câbles électriques - Calcul du courant admissible - Partie 2: Résistance thermique - Section 2: Méthode de calcul des coefficients de réduction de l'intensité de courant admissible pour des groupes de câbles posés à l'air libre et protégés du rayonnement solaire direct

CERN Document Server

1995-01-01

Câbles électriques - Calcul du courant admissible - Partie 2: Résistance thermique - Section 2: Méthode de calcul des coefficients de réduction de l'intensité de courant admissible pour des groupes de câbles posés à l'air libre et protégés du rayonnement solaire direct

Study of cosmic ray interaction model based on atmospheric muons for the neutrino flux calculation

International Nuclear Information System (INIS)

Sanuki, T.; Honda, M.; Kajita, T.; Kasahara, K.; Midorikawa, S.

2007-01-01

We have studied the hadronic interaction for the calculation of the atmospheric neutrino flux by summarizing the accurately measured atmospheric muon flux data and comparing with simulations. We find the atmospheric muon and neutrino fluxes respond to errors in the π-production of the hadronic interaction similarly, and compare the atmospheric muon flux calculated using the HKKM04 [M. Honda, T. Kajita, K. Kasahara, and S. Midorikawa, Phys. Rev. D 70, 043008 (2004).] code with experimental measurements. The μ + +μ - data show good agreement in the 1∼30 GeV/c range, but a large disagreement above 30 GeV/c. The μ + /μ - ratio shows sizable differences at lower and higher momenta for opposite directions. As the disagreements are considered to be due to assumptions in the hadronic interaction model, we try to improve it phenomenologically based on the quark parton model. The improved interaction model reproduces the observed muon flux data well. The calculation of the atmospheric neutrino flux will be reported in the following paper [M. Honda et al., Phys. Rev. D 75, 043006 (2007).

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

TORT application in reactor pressure vessel neutron flux calculations

International Nuclear Information System (INIS)

Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

1994-01-01

The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

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)

Damage flux analysis. Solid state detector and Monte-Carlo calculation

International Nuclear Information System (INIS)

Genthon, J.P.; Nimal, J.C.; Vergnaud, T.

1975-09-01

The change of resistivity induced by radiation in materials is particularly suitable for the measurement of equivalent damage fluxes, when it is used at low fluence for calibration of more classical activation reactions used at high fluences. A graphite and a tungsten detector are briefly described and results obtained in a good number of European reactors are given. The polykinetic three dimensional Monte-Carlo code Tripoli is used for calculation of damage fluxes. Comparison with above measurements shows a good agreement and confirms the use of the EURATOM damaging function for graphite [fr

Quantitative calculations of helium ion escape fluxes from the polar ionospheres

International Nuclear Information System (INIS)

Raitt, W.J.; Schunk, R.W.; Banks, P.M.

1978-01-01

Recent experimental measurements of He + outward fluxes have been obtained for winter and summer hemispheres. The observed fluxes indicate an average He + escape flux of 2 x 10 7 cm -2 s -1 in the winter hemisphere and a factor of 10-20 lower in the summer hemisphere. Earlier theoretical calculations had yielded winter fluxes a factor of 4 lower than the measured values and summer fluxes a further factor of 20 below the winter fluxes. We have attempted to reduce this discrepancy between our earlier theoretical model and the experimental observations by improving our theoretical model in the following ways. The helium photoionization cross sections used are accurate to 10%, the latest solar EUV fluxes measured by the Atmosphere Explorer satellites have been incorporated, and the most recent MSIS model of the neutral atmosphere is contained in the model. A range of conditions covering solar cycle, seasonal, and geomagnetic conditions were studied. The results show a maximum He + escape flux of 1.4 x 10 7 cm -2 s -1 for solar maximum, winter, low magnetic activity conditions, which is within the scatter of the measured fluxes. The computed summer He + escape flux is a factor of 20 lower than the winter value, a result which is in reasonable agreement with the summer experimental observations. Possible reasons for the slight discrepancy between theory and experiment in summer are discussed

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

An implicit flux-split algorithm to calculate hypersonic flowfields in chemical equilibrium

Science.gov (United States)

Palmer, Grant

1987-01-01

An implicit, finite-difference, shock-capturing algorithm that calculates inviscid, hypersonic flows in chemical equilibrium is presented. The flux vectors and flux Jacobians are differenced using a first-order, flux-split technique. The equilibrium composition of the gas is determined by minimizing the Gibbs free energy at every node point. The code is validated by comparing results over an axisymmetric hemisphere against previously published results. The algorithm is also applied to more practical configurations. The accuracy, stability, and versatility of the algorithm have been promising.

Analytical 3-D force calculation of a transverse flux machine

NARCIS (Netherlands)

Kremers, M.F.J.; Paulides, J.J.H.; Janssen, J.L.G.; Lomonova, E.A.

2014-01-01

Transverse Flux Machine (TFM) designs are, in general, based on 3-D Finite Element Methods (FEM). Previous attempts to perform analytical designs have been limited to Magnetic Equivalent Circuits (MEC). In this paper, for the first time, propulsion force calculation of TFMs is performed using an

Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

1996-01-01

A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

The influence of xenon poisoning in high-flux reactors on the choice of control rod speeds (1961); Influence de l'empoisonnement xenon dans les piles a haut flux sur le choix de la vitesse des barres de controle (1961)

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

- The general laws are restated concerning the changes in xenon and iodine concentrations in thermal neutron reactors, assuming an uniform neutron flux distribution in the core. It is shown how the evolution in the xenon poisoning influences the selection of the control rod speed, at start-up. Certain simple methods of calculation are developed making it possible to resolve the problem of the choice of this speed in the case where the xenon poisoning is taken into account. (author) [French] - On rappelle les lois generales relatives aux evolutions de concentration xenon et iode dans les piles atomiques a neutrons thermiques lorsqu'on suppose une repartition uniforme du flux de neutrons dans le coeur. On montre comment l'evolution de l'empoisonnement xenon influe sur le choix de la vitesse des barres de controle en periode de demarrage. On developpe certaines methodes de calculs simples permettant de resoudre le probleme du choix de la vitesse des barres de controle, dans le cas ou l'on tient compte de l'empoisonnement xenon. (auteur)

A sensitivity study on neutron flux variation due to 10B concentration in dose calculation for BNCT

International Nuclear Information System (INIS)

Jung, Sang Hoon

2006-02-01

The effects of inclusion of 10 B concentration on neutron flux and dose in dose calculation were studied. In order to provide the quantitative effects of inclusion of 10 B concentrations on depressions of neutron and photon flux and dose, the fluxes and doses with voxel head phantoms for various 10 B concentrations homogeneously distributed were calculated by using MCNPX simulations. A lithium target system and beam shaping assembly, which have been developed at the Hanyang University, were used as epithermal neutron beam. The calculation results show that the neutron flux at the center of the head phantom decreases by approximately 5.4% per 10 ppm of 10 B concentration in comparison with the neutron flux in the case of boron-free. It was also observed that the tissue dose at the center of the head phantom is depressed by approximately 4.7% per 10 ppm of the 10 B concentration and the tumor dose by approximately 5.3% per 10 ppm. According to depth of tumors, it was observed that the depressions of the doses in the tumors are ranged in 3.7 ∼ 9.2%. The dose calculations in the case of boron-free show that it is overestimated in comparison with the dose calculations in the cases of the inclusion of 10 B concentrations for the normal tissue and the tumors. Therefore, in dose calculation for BNCT, the depressions of neutron flux and dose should be considered. The results in this study are available to setting up the depression ratios which can be used for converting neutron and gamma fluxes and doses in phantom with boron free into the fluxes and doses in phantom with inclusion of 10 B concentrations in treatment. It is expected that the depression ratios is practicable to dose evaluation for BNCT

Calculation of gamma-ray flux density above the Venus and Earth surfaces

International Nuclear Information System (INIS)

Surkov, Yu.A.; Manvelyan, O.S.

1987-01-01

Calculational results of dependence of flux density of nonscattered gamma-quanta on the height above the Venus and Earth planet surfaces are presented in the paper. Areas, where a certain part of gamma quanta is accumulated, are calaculted for each height. Spectra of scattered gamma quanta and their integral fluxes at different heights above the Venera planet surface are calculated. Effect of the atmosphere on gamma radiation recorded is considered. The results obtained allow to estimate optimal conditions for measuring gamma-fields above the Venus and Earth planet surfaces, to determine the area of the planet surface investigated. They are also necessary to determine the elementary composition of the rock according to the characteristic gamma radiation spectrum recorded

The Eddington approximation calculation of radiation flux in the atmosphere–ocean system

International Nuclear Information System (INIS)

Shi, Chong; Nakajima, Teruyuki

2015-01-01

An analytical approximation method is presented to calculate the radiation flux in the atmosphere–ocean system using the Eddington approximation when the upwelling radiation from the ocean body is negligibly small. Numerical experiments were carried out to investigate the feasibility of the method in two cases: flat and rough ocean surfaces. The results show good consistency for the reflectivity at the top of atmosphere and transmissivity just above the ocean surface, in comparison with the exact values calculated by radiative transfer models in each case. Moreover, an obvious error might be introduced for the calculation of radiation flux at larger solar zenith angles when the roughness of the ocean surface is neglected. - Highlights: • The Eddington approximation method is extended to the atmosphere–ocean system. • The roughness of ocean surface cannot be neglected at lager solar zenith angles. • Unidirectional reflectivity for rough ocean surface is proposed

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2001-01-01

The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

International Nuclear Information System (INIS)

Scharmer, K.

1969-01-01

The results of experiments in the light water cooled D 2 O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k eff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D 2 O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [fr

Atmospheric gamma-ray observation with the BETS detectorfor calibrating atmospheric neutrino flux calculations

CERN Document Server

Kasahara, K.; Torii, S.; Tamura, T.; Tateyama, N.; Yoshida, K.; Yamagami, T.; Saito, Y.; Nishimura, J.; Murakami, H.; Kobayashi, T.; Komori, Y.; Honda, M.; Ohuchi, T.; Midorikawa, S.; Yuda, T.

2002-01-01

We observed atmospheric gamma-rays around 10 GeV at balloon altitudes (15~25 km) and at a mountain (2770 m a.s.l). The observed results were compared with Monte Carlo calculations to find that an interaction model (Lund Fritiof1.6) used in an old neutrino flux calculation was not good enough for describing the observed values. In stead, we found that two other nuclear interaction models, Lund Fritiof7.02 and dpmjet3.03, gave much better agreement with the observations. Our data will serve for examining nuclear interaction models and for deriving a reliable absolute atmospheric neutrino flux in the GeV region.

Application of generalized perturbation theory to flux disadvantage factor calculations

International Nuclear Information System (INIS)

Sallam, O.H.; Akimov, I.S.; Naguib, K.; Hamouda, I.

1979-01-01

The possibility of using the generalized perturbation theory to calculate the perturbation of the flux disadvantage factors of reactor cell, resulting from the variation of the cell parameters, is studied. For simplicity the one-group diffusion approximation is considered. All necessary equations are derived for variations both of the cell dimensions. Numerical results are presented in the paper

Simultaneous global calculation of flux and importance with forward Monte Carlo

International Nuclear Information System (INIS)

Deutsch, O.L.; Carter, L.L.

1977-01-01

A procedure is described for obtaining flux and importance globally in one Monte Carlo calculation at small to moderate incremental cost in terms of the time required to process a fixed number of particle histories. The application of this procedure and analysis of results are illustrated for a prototypical controlled thermonuclear reactor (CTR) streaming problem with coolant pipe penetrations through a concrete magnet shield. Our experience indicates that the availability of global information about both flux and importance can help to generate intuition in multidimensional shielding problems and can be of significant value during the early phase of shield design

Measurement and calculation of fast neutron flux in a zero-energy reactor

International Nuclear Information System (INIS)

Day, D.H.; Fox, W.N.; Hyder, H.R.

1963-05-01

An activation technique for measuring relative fast neutron fluxes is described which has some advantages over the normal method using U238 fission. The technique is based on the formation of Rh 103 after inelastic scattering of neutrons above 100 keV in energy. This isomer decays with a 57.4 minute half-life giving an easily measurable γ-activity. The energy dependence of the inelastic scattering cross-section of Rh 103 is similar to that of the fission cross-section of U 238 thus making the results of direct relevance to reactor calculations. Using the Rh 103 activation technique, measurements have been made of the fast neutron flux distribution in a typical pressure tube heavy water lattice and are compared in this report with theoretical calculations using the MONTE CARLO method. (author)

Calculation of neutron flux and reactivity by perturbation theory at high order

International Nuclear Information System (INIS)

Silva, W.L.P. da; Silva, F.C. da; Thome Filho, Z.D.

1982-01-01

A high order pertubation theory is studied, independent of time, applied to integral parameter calculation of a nuclear reactor. A pertubative formulation, based on flux difference technique, which gives directy the reactivity and neutron flux up to the aproximation order required, is presented. As an application of the method, global pertubations represented by fuel temperature variations, are used. Tests were done aiming to verify the relevancy of the approximation order for several intensities of the pertubations considered. (E.G.) [pt

Improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1986-01-01

An improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell has been developed. Expanding the neutron flux and source into a series of even powers of the radius, one' gets a convenient method for integration of the one-energy group integral transport equation. It is shown that it is possible to perform an analytical integration in the x-y plane in one variable and to use the effective Gaussian integration over another one. Choosing a convenient distribution of space points in fuel and moderator the transport matrix calculation and cell reaction rate integration were condensed. On the basis of the proposed method, the computer program DISKRET for the ZUSE-Z 23 K computer has been written. The suitability of the proposed method for the calculation of the thermal-neutron-flux distribution in a reactor cell can be seen from the test results obtained. Compared with the other collision probability methods, the proposed treatment excels with a mathematical simplicity and a faster convergence. (author)

Calculation of the anti-trap factor in heavy water lattices; Calcul du facteur antitrappe dans les reseaux a eau lourde

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

The calculation of the anti-trap factor of a lattice is complex when a large fraction of captures occurs in a range of energies where the spectrum in the fuel is considerably different from the simple dE/E law. This is particularly true for heavy water lattices in which the distances. between the bars are generally fairly large with respect to the slowing-down length. In order to take into account this effect it is necessary both to know the constitution of the effective resonance integral as a function of the energy, and to be able to calculate the distribution in the fuel. This report is devoted to these two problems. An improved method of treating the statistical domain makes it possible to plot the curves of the cross-sections per unit lethargy for various shapes of the fuel. Furthermore, the slowing-down of the neutrons is studied using a Monte-Carlo method which makes it possible in particular to take into account the perturbations caused by the non-moderating rods. A study is also made of the problem of shielding effects due to the captures themselves. (authors) [French] Le calcul du facteur antitrappe dans un reseau est complique lorsqu'une fraction importante des captures a lieu dans un domaine d'energie ou le spectre dans le combustible s'ecarte sensiblement de la loi simple en dE/E. Ceci est particulierement vrai pour les reseaux a eau lourde dans lesquels les distances entre barres sont en general assez grandes vis-a-vis de la longueur de ralentissement. Pour tenir compte de cet effet il faut connaitre d'une part la decomposition de l'integrale de resonance effective en fonction de l'energie, d'autre part savoir calculer le spectre dans le combustible. Le rapport est consacre a ces deux problemes. Un traitement ameliore du domaine statistique permet de tracer des courbes de sections de capture par unite de lethargie pour differentes geometries de barreaux. D'autre part le ralentissement des neutrons est etudie par une methode de Monte Carlo, qui permet

Calculation of the Flux in a Square Lattice Cell and a Comparison with Measurements

Energy Technology Data Exchange (ETDEWEB)

Apelqvist, G [State Power Board, Stockholm (Sweden)

1961-05-15

A calculation has been made of the thermal neutron flux in a square lattice cell using methods devised by Galanin. The f and L lattice parameters have been expressed in measurable quantities and a comparison made between measured and calculated values.

Eddy current loss calculation and thermal analysis of axial-flux permanent magnet couplers

Directory of Open Access Journals (Sweden)

Di Zheng

2017-02-01

Full Text Available A three-dimensional magnetic field analytical model of axial-flux permanent magnet couplers is presented to calculate the eddy current loss, and the prediction of the copper plate temperature under various loads is analyzed. The magnetic field distribution is calculated, and then the eddy current loss is obtained, with the magnetic field analytical model established in cylindrical coordinate. The influence of various loads on eddy current loss is analyzed. Furthermore, a thermal model of axial-flux permanent magnet couplers is established by taking the eddy current loss as the heat source, using the electromagnetic-thermal coupled method. With the help of the thermal model, the influence of various loads on copper plate temperature rise is also analyzed. The calculated results are compared with the results of finite element method and measurement. The comparison results confirm the validity of the magnetic field analytical model and thermal model.

Time-dependent magnetization of a type-II superconductor numerically calculated by using the flux-creep equation

International Nuclear Information System (INIS)

Lee, J. H.; Park, I. S.; Ahmad, D.; Kim, D.; Kim, Y. C.; Ko, R. K.; Jeong, D. Y.

2012-01-01

The macroscopic magnetic behaviors of a type-II superconductor, such as the field- or the temperature-dependent magnetization, have been described by using critical state models. However, because the models are time-independent, the magnetic relaxation in a type-II superconductor cannot be described by them, and the time dependence of the magnetization can affect the field or the temperature-dependent magnetization curve described by the models. In order to avoid the time independence of critical state models, we try the numerical calculation used by Qin et al., who mainly calculated the temperature dependence of the ac susceptibility χ(T). Their calculation showed that the frequency-dependent χ(T) could be obtained by using the flux-creep equation. We calculated the field-dependent magnetization and magnetic relaxation by using a numerical method. The calculated field-dependent magnetization M(H) curves shows the shapes of a typical type-II superconductor. The calculated magnetic relaxation do not show a logarithmic decay of the magnetization, but the addition of a surface barrier to the relaxation calculation caused a clear logarithmic decay of the magnetization, producing a crossover at a mid-time. This means that the logarithmic magnetic relaxation is caused by not only flux creep but also a combination of flux creep and a surface barrier.

Localisation of a neutron source using measurements and calculation of the neutron flux and its gradient

CERN Document Server

Linden, P; Dahl, B; Pázsit, I; Por, G

1999-01-01

We have performed laboratory measurements of the neutron flux and its gradient in a static model experiment, similar to a model problem proposed in Pazsit (Ann. Nucl. Energy 24 (1997) 1257). The experimental system consists of a radioactive neutron source located in a water tank. The measurements are performed using a recently developed very small optical fibre detector. The measured values of the flux and its gradient are then used to test the possibility of localising the source. The results show that it is possible to measure the flux on the circumference of a circle and from this calculate the flux gradient vector. Then, by comparison of the measured quantities with corresponding MCNP calculations, both the direction and the distance to the source are found and thus the position of the source can be determined.

Use of CITATION code for flux calculation in neutron activation analysis with voluminous sample using an Am-Be source

International Nuclear Information System (INIS)

Khelifi, R.; Idiri, Z.; Bode, P.

2002-01-01

The CITATION code based on neutron diffusion theory was used for flux calculations inside voluminous samples in prompt gamma activation analysis with an isotopic neutron source (Am-Be). The code uses specific parameters related to the energy spectrum source and irradiation system materials (shielding, reflector). The flux distribution (thermal and fast) was calculated in the three-dimensional geometry for the system: air, polyethylene and water cuboidal sample (50x50x50 cm). Thermal flux was calculated in a series of points inside the sample. The results agreed reasonably well with observed values. The maximum thermal flux was observed at a distance of 3.2 cm while CITATION gave 3.7 cm. Beyond a depth of 7.2 cm, the thermal flux to fast flux ratio increases up to twice and allows us to optimise the detection system position in the scope of in-situ PGAA

Identification, Calculation Of The Three Dimensional Orbit, And Flux Of Asteroid 2007 TD14

Science.gov (United States)

Pereira, Vincent; Martin, E.; Millan, J.

2012-01-01

In recent years the rate of discovery of asteroids has improved dramatically and has far outstripped efforts to physically characterize them. In this work, we took part in the International Astronomical Search Campaign and confirmed the discovery of asteroid 2007 TD14. We then calculated the two and three dimensional orbit of the asteroid around the sun, given its six elements of orbit. Once the heliocentric and geocentric distances are known, and the visual magnitude of the asteroid obtained through photometry, its diameter can be calculated assuming a suitable value for the albedo. The diameter was 0.718 km and the albedo was 0.039. Using the Standard Thermal Model we calculated the temperature distribution on the surface of the asteroid and the flux of the asteroid in the thermal infrared (1.095 mJy at 22 microns on March 19, 2010). To the best of our knowledge there have been no previous reports of the diameter and flux of the asteroid. Our ultimate goal is to compare our flux values with newly released data from NASA Wide-field Infrared Survey Explorer Mission and thus obtain better estimates of the asteroid diameter and albedo.

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

The calculation of neutron flux using Monte Carlo method

Science.gov (United States)

Günay, Mehtap; Bardakçı, Hilal

2017-09-01

In this study, a hybrid reactor system was designed by using 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII.0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

A method for prompt calculation of neutron flux from measured SPND [self-powered neutron detectors] currents

International Nuclear Information System (INIS)

Kulacsy, K.; Lux, I.

1997-01-01

A new, approximate method is given to calculate the in-core flux from the current of SPNDs, with a delay of only a few seconds. The stability of this stepwise algorithm is proven to be satisfactory, and the results of tests performed both on synthetic and on real data are presented. The reconstructed flux is found to follow both steady state and transient fluxes well. (author)

Tritium enrichment by thermal diffusion. I. Calculation of an installation for measuring natural tritium; Enrichissement du tritium par diffusion thermique. - I. Calcul d'une installation destinee a la mesure du tritium naturel

Energy Technology Data Exchange (ETDEWEB)

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

1966-07-01

The natural content of tritium is so low that its measurement generally requires a preliminary enrichment. The thermal diffusion on hydrogen is studied as an enrichment method. The installation studied comprises two stages of columns of the hot-wire type, together with a device for transferring the tritium from the water sample into the hydrogen in the columns using catalytic exchange. A complete mathematical treatment for the operation of such a unit has been made and programmed for the IBM 7094 computer. An optimization has been effected by means of this program. It is shown that for similar performances, less hydrogen is retained in the case of hot-wire type columns than in the case of columns composed of concentric tubes. (authors) [French] La teneur naturelle du tritium est si faible que sa mesure demande generalement un enrichissement prealable. On etudie la diffusion thermique sur l'hydrogene comme moyen d'enrichissement. L'installation que l'on etudie comprend deux etages de colonnes du type fil chaud, et un dispositif de transfert du tritium de l'echantillon d'eau dans l'hydrogene des colonnes par echange catalytique. Un traitement mathematique complet du fonctionnement d'un tel ensemble a ete etabli et programme sur machine IBM 7094. Une optimisation a ete faite a l'aide du programme. On montre egalement qu'a performances egales, la retenue d'hydrogene est plus faible dans le cas des colonnes du type fil chaud que dans le cas des colonnes du type tubes concentriques. (auteurs)

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

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

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

One group neutron flux at a point in a cylindrical reactor cell calculated by Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Kocic, A [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1974-01-15

Mean values of the neutron flux over material regions and the neutron flux at space points in a cylindrical annular cell (one group model) have been calculated by Monte Carlo. The results are compared with those obtained by an improved collision probability method (author)

Influences of various calculation options on heat, water and carbon fluxes determined by open- and closed-path eddy covariance methods

Directory of Open Access Journals (Sweden)

Masahito Ueyama

2012-07-01

Full Text Available Synthesis studies using multiple-site datasets for eddy covariance potentially contain uncertainties originating from the use of different flux calculation options, because the choice of the process for calculating half-hourly fluxes from raw time series data is left to individual researchers. In this study, we quantified the uncertainties associated with different flux calculation methods at seven sites. The differences in the half-hourly fluxes were small, generally of the order less than a few percentiles, but they were substantial for the annual fluxes. After the standardisation under current recommendations in the FLUXNET communities, we estimated the uncertainties in the annual fluxes associated with the flux calculations to be 2.6±2.7 W m−2 (the mean 90% ± confidence interval for the sensible heat flux, 72±37 g C m−2 yr−1 for net ecosystem exchange (NEE, 12±6% for evapotranspiration, 12±6% for gross primary productivity and 16±10% for ecosystem respiration. The self-heating correction strongly influenced the annual carbon balance (143±93 g C m−2 yr−1, not only for cold sites but also for warm sites, but did not fully account for differences between the open- and closed-path systems (413±189 g C m−2 yr−1.

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Effects of tropospheric aerosols on radiative flux calculations at UV and visible wavelengths

International Nuclear Information System (INIS)

Grossman, A.S.; Grant, K.E.

1994-08-01

The surface fluxes in the wavelength range 175 to 735nm have been calculated for an atmosphere which contains a uniformly mixed aerosol layer of thickness 1km at the earth's surface. Two different aerosol types were considered, a rural aerosol, and an urban aerosol. The visibility range for the aerosol layers was 95 to 15 km. Surface flux ratios (15km/95km) were in agreement with previously published results for the rural aerosol layer to within about 2%. The surface flux ratios vary from 7 to 14% for the rural aerosol layer and from 13 to 23% for the urban aerosol layer over the wavelength range. A tropospheric radiative forcing of about 1.3% of the total tropospheric flux was determined for the 95km to 15km visibility change in the rural aerosol layer, indicating the potential of tropospheric feedback effects on the surface flux changes. This effect was found to be negligible for the urban aerosol layer. Stratospheric layer heating rate changes due to visibility changes in either the rural or urban aerosol layer were found to be negligible

An Analysis on the Calculation Efficiency of the Responses Caused by the Biased Adjoint Fluxes in Hybrid Monte Carlo Simulation

International Nuclear Information System (INIS)

Khuat, Quang Huy; Kim, Song Hyun; Kim, Do Hyun; Shin, Chang Ho

2015-01-01

This technique is known as Consistent Adjoint Driven Importance Sampling (CADIS) method and it is implemented in SCALE code system. In the CADIS method, adjoint transport equation has to be solved to determine deterministic importance functions. Using the CADIS method, a problem was noted that the biased adjoint flux estimated by deterministic methods can affect the calculation efficiency and error. The biases of adjoint function are caused by the methodology, calculation strategy, tolerance of result calculated by the deterministic method and inaccurate multi-group cross section libraries. In this paper, a study to analyze the influence of the biased adjoint functions into Monte Carlo computational efficiency is pursued. In this study, a method to estimate the calculation efficiency was proposed for applying the biased adjoint fluxes in the CADIS approach. For a benchmark problem, the responses and FOMs using SCALE code system were evaluated as applying the adjoint fluxes. The results show that the biased adjoint fluxes significantly affects the calculation efficiencies

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

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

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

The neutrons flux density calculations by Monte Carlo code for the double heterogeneity fuel

International Nuclear Information System (INIS)

Gurevich, M.I.; Brizgalov, V.I.

1994-01-01

This document provides the calculation technique for the fuel elements which consists of the one substance as a matrix and the other substance as the corn embedded in it. This technique can be used in the neutron flux density calculation by the universal Monte Carlo code. The estimation of accuracy is presented too. (authors). 6 refs., 1 fig

Tritium enrichment by thermal diffusion. I. Calculation of an installation for measuring natural tritium; Enrichissement du tritium par diffusion thermique. - I. Calcul d'une installation destinee a la mesure du tritium naturel

Energy Technology Data Exchange (ETDEWEB)

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

1966-07-01

The natural content of tritium is so low that its measurement generally requires a preliminary enrichment. The thermal diffusion on hydrogen is studied as an enrichment method. The installation studied comprises two stages of columns of the hot-wire type, together with a device for transferring the tritium from the water sample into the hydrogen in the columns using catalytic exchange. A complete mathematical treatment for the operation of such a unit has been made and programmed for the IBM 7094 computer. An optimization has been effected by means of this program. It is shown that for similar performances, less hydrogen is retained in the case of hot-wire type columns than in the case of columns composed of concentric tubes. (authors) [French] La teneur naturelle du tritium est si faible que sa mesure demande generalement un enrichissement prealable. On etudie la diffusion thermique sur l'hydrogene comme moyen d'enrichissement. L'installation que l'on etudie comprend deux etages de colonnes du type fil chaud, et un dispositif de transfert du tritium de l'echantillon d'eau dans l'hydrogene des colonnes par echange catalytique. Un traitement mathematique complet du fonctionnement d'un tel ensemble a ete etabli et programme sur machine IBM 7094. Une optimisation a ete faite a l'aide du programme. On montre egalement qu'a performances egales, la retenue d'hydrogene est plus faible dans le cas des colonnes du type fil chaud que dans le cas des colonnes du type tubes concentriques. (auteurs)

The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method

International Nuclear Information System (INIS)

Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang

2011-01-01

Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)

Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2005-03-01

The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2006-01-01

The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

La diffusion du tourisme dans l'archipel comorien

Directory of Open Access Journals (Sweden)

Jean-Christophe Gay

2001-12-01

Full Text Available Des considérations politiques ont largement influencé la diffusion du tourisme dans l'archipel comorien. Sa balkanisation explique que deux îles sont concernées par le tourisme, en dépit d'un flux réduit.

Inferring CO2 Fluxes from OCO-2 for Assimilation into Land Surface Models to Calculate Net Ecosystem Exchange

Science.gov (United States)

Prouty, R.; Radov, A.; Halem, M.; Nearing, G. S.

2016-12-01

Investigations of mid to high latitude atmospheric CO2 show a growing seasonal amplitude. Land surface models poorly predict net ecosystem exchange (NEE) and are unable to substantiate these sporadic observations. An investigation of how the biosphere has reacted to changes in atmospheric CO2 is essential to our understanding of potential climate-vegetation feedbacks. A global, seasonal investigation of CO2-flux is then necessary in order to assimilate into land surface models for improving the prediction of annual NEE. The Atmospheric Radiation Measurement program (ARM) of DOE collects CO2-flux measurements (in addition to CO2 concentration and various other meteorological quantities) at several towers located around the globe at half hour temporal frequencies. CO2-fluxes are calculated via the eddy covariance technique, which utilizes CO2-densities and wind velocities to calculate CO2-fluxes. The global coverage of CO2 concentrations as provided by the Orbiting Carbon Observatory (OCO-2) provide satellite-derived CO2 concentrations all over the globe. A framework relating the satellite-inferred CO2 concentrations collocated with the ground-based ARM as well as Ameriflux stations would enable calculations of CO2-fluxes far from the station sites around the entire globe. Regression techniques utilizing deep-learning neural networks may provide such a framework. Additionally, meteorological reanalysis allows for the replacement of the ARM multivariable meteorological variables needed to infer the CO2-fluxes. We present the results of inferring CO2-fluxes from OCO-2 CO2 concentrations for a two year period, Sept. 2014- Sept. 2016 at the ARM station located near Oklahoma City. A feed-forward neural network (FFNN) is used to infer relationships between the following data sets: F([ARM CO2-density], [ARM Meteorological Data]) = [ARM CO2-Flux] F([OCO-2 CO2-density],[ARM Meteorological Data]) = [ARM CO2-Flux] F([ARM CO2-density],[Meteorological Reanalysis]) = [ARM CO2-Flux

Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

Directory of Open Access Journals (Sweden)

Avramović Ivana

2007-01-01

Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

Flux and brightness calculations for various synchrotron radiation sources

International Nuclear Information System (INIS)

Weber, J.M.; Hulbert, S.L.

1991-11-01

Synchrotron radiation (SR) storage rings are powerful scientific and technological tools. The first generation of storage rings in the US., e.g., SURF (Washington, D.C.), Tantalus (Wisconsin), SSRL (Stanford), and CHESS (Cornell), revolutionized VUV, soft X-ray, and hard X-ray science. The second (present) generation of storage rings, e.g. the NSLS VUV and XRAY rings and Aladdin (Wisconsin), have sustained the revolution by providing higher stored currents and up to a factor of ten smaller electron beam sizes than the first generation sources. This has made possible a large number of experiments that could not performed using first generation sources. In addition, the NSLS XRAY ring design optimizes the performance of wigglers (high field periodic magnetic insertion devices). The third generation storage rings, e.g. ALS (Berkeley) and APS (Argonne), are being designed to optimize the performance of undulators (low field periodic magnetic insertion devices). These extremely high brightness sources will further revolutionize x-ray science by providing diffraction-limited x-ray beams. The output of undulators and wigglers is distinct from that of bending magnets in magnitude, spectral shape, and in spatial and angular size. Using published equations, we have developed computer programs to calculate the flux, central intensity, and brightness output bending magnets and selected wigglers and undulators of the NSLS VUV and XRAY rings, the Advanced Light Source (ALS), and the Advanced Photon Source (APS). Following is a summary of the equations used, the graphs and data produced, and the computer codes written. These codes, written in the C programming language, can be used to calculate the flux, central intensity, and brightness curves for bending magnets and insertion devices on any storage ring

A simple calculation algorithm to separate high-resolution CH4 flux measurements into ebullition and diffusion-derived components

Science.gov (United States)

Hoffmann, Mathias; Schulz-Hanke, Maximilian; Garcia Alba, Joana; Jurisch, Nicole; Hagemann, Ulrike; Sachs, Torsten; Sommer, Michael; Augustin, Jürgen

2016-04-01

Processes driving methane (CH4) emissions in wetland ecosystems are highly complex. Especially, the separation of CH4 emissions into ebullition and diffusion derived flux components, a perquisite for the mechanistic process understanding and identification of potential environmental driver is rather challenging. We present a simple calculation algorithm, based on an adaptive R-script, which separates open-water, closed chamber CH4 flux measurements into diffusion- and ebullition-derived components. Hence, flux component specific dynamics are revealed and potential environmental driver identified. Flux separation is based on a statistical approach, using ebullition related sudden concentration changes obtained during high resolution CH4 concentration measurements. By applying the lower and upper quartile ± the interquartile range (IQR) as a variable threshold, diffusion dominated periods of the flux measurement are filtered. Subsequently, flux calculation and separation is performed. The algorithm was verified in a laboratory experiment and tested under field conditions, using flux measurement data (July to September 2013) from a flooded, former fen grassland site. Erratic ebullition events contributed 46% to total CH4 emissions, which is comparable to values reported by literature. Additionally, a shift in the diurnal trend of diffusive fluxes throughout the measurement period, driven by the water temperature gradient, was revealed.

Experimental study and technique for calculation of critical heat fluxes in helium boiling in tubes

International Nuclear Information System (INIS)

Arkhipov, V.V.; Kvasnyuk, S.V.; Deev, V.I.; Andreev, V.K.

1979-01-01

Studied is the effect of regime parameters on critical heat loads in helium boiling in a vertical tube in the range of mass rates of 80 2 xc) and pressures of 100<=p<=200 kPa for the vapor content range corresponding to the heat exchange crisis of the first kind. The method for calculating critical heat fluxes describing experimental data with the error less than +-15% is proposed. The critical heat loads in helium boiling in tubes reduce with the growth of pressure and vapor content in the regime parameter ranges under investigation. Both positive and negative effects of the mass rate on the critical heat flux are observed. The calculation method proposed satisfactorily describes the experimental data

Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Williams, L.R.

1994-07-01

This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

Neutron flux calculation by means of Monte Carlo methods

International Nuclear Information System (INIS)

Barz, H.U.; Eichhorn, M.

1988-01-01

In this report a survey of modern neutron flux calculation procedures by means of Monte Carlo methods is given. Due to the progress in the development of variance reduction techniques and the improvements of computational techniques this method is of increasing importance. The basic ideas in application of Monte Carlo methods are briefly outlined. In more detail various possibilities of non-analog games and estimation procedures are presented, problems in the field of optimizing the variance reduction techniques are discussed. In the last part some important international Monte Carlo codes and own codes of the authors are listed and special applications are described. (author)

Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

Energy Technology Data Exchange (ETDEWEB)

Slater, C.O.

1990-07-01

Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1967-05-15

Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.

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

1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E{sub M} between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E{sup 1/2} e{sup -{beta}}{sup E} from of the spectrum above 3 MeV. The values of {beta} are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E {sigma}{sub S}(E)) form ({sigma}{sub S}(E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [French] 1. Neutrons thermiques: La temperature du spectre des neutrons thermiques a ete determinee par la methode (or-lutecium). Cette temperature varie de 300 deg. K (temperature du moderateur) a 30 cm du coeur, a 350 deg. K dans une encoche du coeur. 2. Neutrons en ralentissement: 4 detecteurs resonnants ont ete employes (Au, In, Co, Mn). Le spectre peut etre mis sous la forme 1/E quelle que soit la distance a la limite coeur-reflecteur. L'energie de raccordement E{sub M} entre spectre des neutrons thermiques et spectre en ralentissement est environ 0,23 eV et independante de la temperature du spectre de Maxwell. La diminution relative du flux en ralentissement par rapport au flux thermique quand la distance au coeur

Spatially resolved charge exchange flux calculations on the Toroidal Pumped Limiter of Tore Supra

International Nuclear Information System (INIS)

Marandet, Y.; Tsitrone, E.; Boerner, P.; Reiter, D.; Beaute, A.; Delchambre, E.; Escarguel, A.; Brezinsek, S.; Genesio, P.; Gunn, J.; Monier-Garbet, P.; Mitteau, R.; Pegourie, B.

2009-01-01

A spatially resolved calculation of the charge exchange particle and energy fluxes on the Toroidal Pumped Limiter (TPL) of Tore Supra is presented, as a first step towards a better understanding and modelling of carbon erosion, migration, as well as deuterium codeposition and bulk diffusion of deuterium in Tore Supra. The results are obtained with the EIRENE code run in a 3D geometry. Physical and chemical erosion maps on the TPL are calculated, and the contribution of neutrals to erosion, especially in the self-shadowed area, is calculated.

A transmission probability method for calculation of neutron flux distributions in hexagonal geometry

International Nuclear Information System (INIS)

Wasastjerna, F.; Lux, I.

1980-03-01

A transmission probability method implemented in the program TPHEX is described. This program was developed for the calculation of neutron flux distributions in hexagonal light water reactor fuel assemblies. The accuracy appears to be superior to diffusion theory, and the computation time is shorter than that of the collision probability method. (author)

Formulation of coarse mesh finite difference to calculate mathematical adjoint flux; Formulacao de diferencas finitas de malha grossa para calculo do fluxo adjunto matematico

Energy Technology Data Exchange (ETDEWEB)

Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

2002-07-01

The objective of this work is the obtention of the mathematical adjoint flux, having as its support the nodal expansion method (NEM) for coarse mesh problems. Since there are difficulties to evaluate this flux by using NEM. directly, a coarse mesh finite difference program was developed to obtain this adjoint flux. The coarse mesh finite difference formulation (DFMG) adopted uses results of the direct calculation (node average flux and node face averaged currents) obtained by NEM. These quantities (flux and currents) are used to obtain the correction factors which modify the classical finite differences formulation . Since the DFMG formulation is also capable of calculating the direct flux it was also tested to obtain this flux and it was verified that it was able to reproduce with good accuracy both the flux and the currents obtained via NEM. In this way, only matrix transposition is needed to calculate the mathematical adjoint flux. (author)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1969-07-01

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

Systematic assembly homogenization and local flux reconstruction for nodal method calculations of fast reactor power distributions

International Nuclear Information System (INIS)

Dorning, J.J.

1991-01-01

A simultaneous pin lattice cell and fuel bundle homogenization theory has been developed for use with nodal diffusion calculations of practical reactors. The theoretical development of the homogenization theory, which is based on multiple-scales asymptotic expansion methods carried out through fourth order in a small parameter, starts from the transport equation and systematically yields: a cell-homogenized bundled diffusion equation with self-consistent expressions for the cell-homogenized cross sections and diffusion tensor elements; and a bundle-homogenized global reactor diffusion equation with self-consistent expressions for the bundle-homogenized cross sections and diffusion tensor elements. The continuity of the angular flux at cell and bundle interfaces also systematically yields jump conditions for the scaler flux or so-called flux discontinuity factors on the cell and bundle interfaces in terms of the two adjacent cell or bundle eigenfunctions. The expressions required for the reconstruction of the angular flux or the 'de-homogenization' theory were obtained as an integral part of the development; hence the leading order transport theory angular flux is easily reconstructed throughout the reactor including the regions in the interior of the fuel bundles or computational nodes and in the interiors of the pin lattice cells. The theoretical development shows that the exact transport theory angular flux is obtained to first order from the whole-reactor nodal diffusion calculations, done using the homogenized nuclear data and discontinuity factors, is a product of three computed quantities: a ''cell shape function''; a ''bundle shape function''; and a ''global shape function''. 10 refs

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

International Nuclear Information System (INIS)

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

Fast modeling of flux trapping cascaded explosively driven magnetic flux compression generators.

Science.gov (United States)

Wang, Yuwei; Zhang, Jiande; Chen, Dongqun; Cao, Shengguang; Li, Da; Liu, Chebo

2013-01-01

To predict the performance of flux trapping cascaded flux compression generators, a calculation model based on an equivalent circuit is investigated. The system circuit is analyzed according to its operation characteristics in different steps. Flux conservation coefficients are added to the driving terms of circuit differential equations to account for intrinsic flux losses. To calculate the currents in the circuit by solving the circuit equations, a simple zero-dimensional model is used to calculate the time-varying inductance and dc resistance of the generator. Then a fast computer code is programmed based on this calculation model. As an example, a two-staged flux trapping generator is simulated by using this computer code. Good agreements are achieved by comparing the simulation results with the measurements. Furthermore, it is obvious that this fast calculation model can be easily applied to predict performances of other flux trapping cascaded flux compression generators with complex structures such as conical stator or conical armature sections and so on for design purpose.

Calculation of the flux density of gamma rays above the surface of Venus and the Earth

International Nuclear Information System (INIS)

Surkov, Yu.A.; Manvelyan, O.S.

1987-01-01

In this article the authors present the results of calculating the flux density of unscattered gamma rays as a function of height above the surfaces of Venus and the Earth. At each height they calculate the areas which will collect a certain fraction of the gamma rays. The authors calculate the spectra of scattered gamma rays, as well as their integrated fluxes at various heights above the surface of Venus. They consider how the atmosphere will affect the recording of gamma rays. Their results enable them to evaluate the optimal conditions for measuring the gamma-ray fields above the surfaces of Venus and the Earth and to determine the area of the planet which can be investigated in this way. These results are also necessary if they are to determine the elemental composition of the rock from the characteristic recorded spectrum of gamma radiation

Energy balance of the cathode connection and calculation of the yield of caesium diodes; Bilan des energies dans la connexion de cathode et calcul du rendement dans les diodes a cesium

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

The cathode connection is one of the critical elements in the calculation of the yield of caesium diodes. In particular a study is made of the balance of the losses due to the Joule effect, to radiation and to thermal conduction as a function of the shape and the nature of the cathode connection. The internal voltage drop of the diode is obtained from the expression for the total Joule losses of the metallic conductors. The length of the cathode connection is calculated as a function of the losses and the yield is optimized as a function of the relationship existing between the cathodic losses due to thermal conduction and the Joule losses in the connection. The calculations lead to hyper-elliptic integrals which can only be manipulated by series development. The results obtained are approximate. (author) [French] Dans le calcul du rendement des diodes a cesium, la connexion de cathode est l'un des elements critiques. On etudie specialement le bilan des pertes par effet joule, rayonnement et conduction thermique, en fonction de la forme et de la nature de la connexion de cathode. On obtient la chute de tension interne de la diode a partir de l'expression des pertes joules totales dans les conducteurs metalliques. La longueur de la connexion de cathode est calculee en fonction des pertes et l'on optimise le rendement en fonction du rapport qui existe entre les pertes par conduction thermique de la cathode et les pertes joules dans la connexion. Les calculs conduisent a des integrales hyperelliptiques dont la manipulation n'est possible que par des developpements en serie. Les resultats cherches sont approximatifs. (auteur)

A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method; Mise au point d'un estimateur ponctuel du flux et des taux de reactions dans les calculs de transport de particules neutres par la methode de monte carlo

Energy Technology Data Exchange (ETDEWEB)

Authier, N

1998-12-01

One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidencelevel may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method; Mise au point d'un estimateur ponctuel du flux et des taux de reactions dans les calculs de transport de particules neutres par la methode de monte carlo

Energy Technology Data Exchange (ETDEWEB)

Authier, N

1998-12-01

One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidencelevel may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

Optimisation of flux calculation in rivers from discrete water quality surveys, a step towards an expert system

Science.gov (United States)

Raymond, S.; Moatar, F.; Meybeck, M.; Bustillo, V.

2009-04-01

Good estimates of fluxes of suspended particulate matter (SPM), total dissolved solids (TDS) and nutrients and contaminants are required for both Earth System science and river basin management. However, in most cases discrete sampling (weekly to monthly) is the rule. Few flux calculation methods are commonly used, yet their performances, i.e. uncertainties for given frequencies, at given stations and for each water quality variables, remain unknown. Based on a rare set of 1085 station-year of daily flux record for SPM, TDS and nutrients (dissolved and total), the performance of 9 calculations methods is explored. Discrete surveys at various frequencies (3days to 30 days) are simulated by Monte-Carlo sorting (100 runs) on which the 9 fluxes are calculated (annual and interannual). At this stage, the sub-daily variations of fluxes for the medium and large basins are not considered. The dataset for SPM corresponds to 55 stations (600 to 600 000 km2 basin area), 34 stations (700 to 1000000 km2) for TDS and for nutrients we consider 9 stations for NO3-, NH4+, PO43- and Ptot (600 to 30 000 km2). About 80% of the dataset originates from US records (USGS and Lake Erie tributaries survey) and 20% from French stations, this covering a wide range of hydrological and geochemical conditions in the temperate zone. Each sorted flux is compared to known fluxes established on daily records: percentiles of their relative errors (e10, e50 and e90) are used to determine the biases (e50) and the imprecisions (e90-e10) (Walling and Webb, 1981) which are then compared for each of the 6 water quality variables, for each flux methods and for various simulated survey frequencies. The calculation methods include 5 rating-curve approaches (linear"M1", "M2", Phillipps et al, 1999) with and without Ferguson correction (Ferguson, 1987), polynomial, truncated at discharges exceeding median annual or long-term water discharge), 2 methods based on hydrograph separation (Phillips et al, 1999

Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2014-01-01

The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

Calculation of the radial and axial flux and power distribution for a CANDU 6 reactor with both the MCNP6 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada)

2014-07-01

The most recent versions of the Monte Carlo-based probabilistic transport code MCNP6 and the continuous energy reactor physics burnup calculation code Serpent allow for a 3-D geometry calculation accounting for the detailed geometry without unit-cell homogenization. These two codes are used to calculate the axial and radial flux and power distributions for a CANDU6 GENTILLY-2 nuclear reactor core with 37-element fuel bundles. The multiplication factor, actual flux distribution and power density distribution were calculated by using a tally combination for MCNP6 and detector analysis for Serpent. Excellent agreement was found in the calculated flux and power distribution. The Serpent code is most efficient in terms of the computational time. (author)

Measurements and Calculations of the Slowing-Down and Migration Time; Mesures et Calcul du Temps de Ralentissement et de Migration; Izmereniya i raschety vremeni zamedleniya i migratsii; Medicion y Calculo del Tiempo de Moderacion y de Migracion

Energy Technology Data Exchange (ETDEWEB)

Profio, A. E.; Koppel, J. U. [General Atomic Division of General Dynamics Corporation, John Jay Hopkins Laboratory for Pure and Applied Science, San Diego, CA (United States); Adamantiades, A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1965-08-15

'experiences. Le temps moyen est une correction pour les mesures par la methode du temps de vol des spectres de neutrons dans des milieux de grande diffusion, et la variance a pour effet d'imposer une limite a la resolution des experiences. La grandeur de ces parametres est egalement importante pour les detecteurs qui dependent du ralentissement, dans les experiences par | la methode du temps de vol des spectres de neutrons de faible energie sont fournis par un ralentisseur place pres de la source puisee, et en spectrometrie du temps de ralentissement. Diverses methodes analytiques et numeriques ont ete mises au point pour calculer la fonction espaceenergie- angle-temps ou des integrales de cette fonction. Les auteurs montrent que les moments Empty-Set {sup (n)} (r, {Omega}, v, t) = {integral}{sub 0}{sup {infinity}}t{sup n} Empty-Set (r, {Omega}, v, t)dt peuvent etre calcules par application repetee d'un code de transport pour un etat stationnaire. Le terme de la source pour le calcul du nieme moment est egal a nv{sup -1} Empty-Set {sup (n-1)}. Les auteurs presentent des resultats pour des modeles multiplicateurs et non multiplicateurs du reacteur Triga. Une autre methode de calcul tres utile est le code de Monte-Carlo variable dans le temps. Les auteurs presentent les resultats d'un calcul de flux de fuite a partir d'une mince plaque de plomb. Les auteurs ont mesure le temps de ralentissement jusqu 'au seuil cadmium et jusqu 'a la resonance 1,46 eV de l'indium dans l'eau et dans le toluene. Ils ont deceleles rayonsgamma de capture avec un compteura scintillation. La methode necessite l'utilisation d'une source assez intense et d'un detecteur efficace a cause de la brievete du cycle de fonctionnement (petite largeur de bouffee pour la resolution du temps de ralentissement, grand intervalle entre les bouffees pour l'evanouissement des neutrons thermiques) et de la faible probabilite de capture. (author) [Spanish] En muchos experimentos, reviste gran importancia el tiempo medio que

Method of Relative Magnitudes for Calculating Magnetic Fluxes in Electrical Machine

Directory of Open Access Journals (Sweden)

Oleg A.

2018-03-01

Full Text Available Introduction: The article presents the study results of the model of an asynchronous electric motor carried out by the author within the framework of the Priorities Research Program “Research and development in the priority areas of development of Russia’s scientific and technical complex for 2014–2020”. Materials and Methods: A model of an idealized asynchronous machine (with sinusoidal distribution of magnetic induction in air gap is used in vector control systems. It is impossible to create windings for this machine. The basis of the new calculation approach was the Conductivity of Teeth Contours Method, developed at the Electrical Machines Chair of the Moscow Power Engineering Institute (MPEI. Unlike this method, the author used not absolute values, but relative magnitudes of magnetic fluxes. This solution fundamentally improved the method’s capabilities. The relative magnitudes of the magnetic fluxes of the teeth contours do not required the additional consideration for exact structure of magnetic field of tooth and adjacent slots. These structures are identical for all the teeth of the machine and differ only in magnitude. The purpose of the calculations was not traditional harmonic analysis of magnetic induction distribution in air gap of machine, but a refinement of the equations of electric machine model. The vector control researchers used only the cos(θ function as a value of mutual magnetic coupling coefficient between the windings. Results: The author has developed a way to take into account the design of the windings of a real machine by using imaginary measuring winding with the same winding design as a real phase winding. The imaginary winding can be placed in the position of any machine windings. The calculation of the relative magnetic fluxes of this winding helped to estimate the real values of the magnetic coupling coefficients between the windings, and find the correction functions for the model of an idealized

Development of a locally mass flux conservative computer code for calculating 3-D viscous flow in turbomachines

Science.gov (United States)

Walitt, L.

1982-01-01

The VANS successive approximation numerical method was extended to the computation of three dimensional, viscous, transonic flows in turbomachines. A cross-sectional computer code, which conserves mass flux at each point of the cross-sectional surface of computation was developed. In the VANS numerical method, the cross-sectional computation follows a blade-to-blade calculation. Numerical calculations were made for an axial annular turbine cascade and a transonic, centrifugal impeller with splitter vanes. The subsonic turbine cascade computation was generated in blade-to-blade surface to evaluate the accuracy of the blade-to-blade mode of marching. Calculated blade pressures at the hub, mid, and tip radii of the cascade agreed with corresponding measurements. The transonic impeller computation was conducted to test the newly developed locally mass flux conservative cross-sectional computer code. Both blade-to-blade and cross sectional modes of calculation were implemented for this problem. A triplet point shock structure was computed in the inducer region of the impeller. In addition, time-averaged shroud static pressures generally agreed with measured shroud pressures. It is concluded that the blade-to-blade computation produces a useful engineering flow field in regions of subsonic relative flow; and cross-sectional computation, with a locally mass flux conservative continuity equation, is required to compute the shock waves in regions of supersonic relative flow.

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

Semi-analytic flux formulas for shielding calculations

International Nuclear Information System (INIS)

Wallace, O.J.

1976-06-01

A special coordinate system based on the work of H. Ono and A. Tsuro has been used to derive exact semi-analytic formulas for the flux from cylindrical, spherical, toroidal, rectangular, annular and truncated cone volume sources; from cylindrical, spherical, truncated cone, disk and rectangular surface sources; and from curved and tilted line sources. In most of the cases where the source is curved, shields of the same curvature are allowed in addition to the standard slab shields; cylindrical shields are also allowed in the rectangular volume source flux formula. An especially complete treatment of a cylindrical volume source is given, in which dose points may be arbitrarily located both within and outside the source, and a finite cylindrical shield may be considered. Detector points may also be specified as lying within spherical and annular source volumes. The integral functions encountered in these formulas require at most two-dimensional numeric integration in order to evaluate the flux values. The classic flux formulas involving only slab shields and slab, disk, line, sphere and truncated cone sources become some of the many special cases which are given in addition to the more general formulas mentioned above

Calculation of neutron flux distribution of thermal neutrons from microtron converter in a graphite moderator with water reflector

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

The calculation is made of the thermal neutron flux in the moderator and reflector by solving the neutron diffusion equation using the four-group theory. The correction for neutron absorption in the moderator was carried out using the perturbation theory. The calculation was carried out for four groups with the following energy ranges: the first group 2 MeV to 3 keV, the second group 3 keV to 5 eV, the third group 5 eV to 0.025 eV and the fourth group 0.025 eV. The values of the macroscopic cross section of capture and scattering, of the diffusion coefficient, the macroscopic cross section of the moderator, of the neutron age and the extrapolation length for the water-graphite moderator used in the calculations are given. The spatial distribution of the thermal neutron flux is graphically represented for graphite of a 30, 40, and 50 cm radius and for graphite of a 30 and 40 cm radius with a 10 cm water reflector; a graphic comparison is made of the distribution of the thermal neutron flux in water and in graphite, both 40 cm in radius. The system of graphite with reflector proved to be the best and most efficient system for raising the flux density of thermal neutrons. (J.P.)

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)

Continuous-energy adjoint flux and perturbation calculation using the iterated fission probability method in Monte-Carlo code TRIPOLI-4 and underlying applications

International Nuclear Information System (INIS)

Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.

2013-01-01

The first goal of this paper is to present an exact method able to precisely evaluate very small reactivity effects with a Monte Carlo code (<10 pcm). it has been decided to implement the exact perturbation theory in TRIPOLI-4 and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4 is described. To illustrate the efficiency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the 'direct' estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the 'direct' method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. It offers the possibility to split reactivity contributions on both isotopes and reactions. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters

Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

Directory of Open Access Journals (Sweden)

Mohammadnia Meysam

2013-01-01

Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

Spray Evaporation in Turbulent Flow: Numerical Calculations and Detailed Experiments by Phase-Doppler Anemometry Évaporation de brouillard en flux turbulent : calculs numériques et expériences détaillées par anémometrie de phase-Doppler

Directory of Open Access Journals (Sweden)

Sommerfeld M.

2006-11-01

Full Text Available The present paper concerns experiments and numerical calculations of an isopropyl-alcohol spray evaporating in a co-flowing turbulent heated air flow. The measurements provided detailed inlet and boundary conditions for the numerical calculations and allowed the validation of the numerical method and models. Phase-Doppler anemometry was used in order to obtain the spatial change of the droplet size distribution and the correlation between droplet size and velocity throughout the flow field. Additionally, a reliable method based on the detection of the signal amplitudes was applied to determine the droplet mass flux. By integration of the droplet mass flux profiles, the global evaporation rates could be determined for different flow conditions. Numerical calculations of the evaporating spray were performed by the Eulerian / Lagrangian approach. The modelling of droplet evaporation is briefly reviewed prior to the description of the applied numerical models and methods. Calculations for a single phase flow showed good agreement with the experiments. Also for all of the droplet phase properties reasonable agreement with the experiments could be achieved and the global evaporation rates agreed well with the measurements. Cet article expose en détail les expériences et les calculs concernant l'évaporation d'isopropanol pulvérisé dans un flux d'air chaud turbulent. Les mesures ont fourni le détail des conditions initiales et des conditions limites pour les calculs numériques ; elles ont également permis de valider la méthode et le modèle. L'anémométrie de phase-Doppler a permis de définir la modification spatiale de la distribution des dimensions de gouttelettes ainsi que la corrélation entre dimension et vitesse des gouttelettes, dans l'ensemble du champ d'écoulement. De plus, une méthode fiable fondée sur la détection des amplitudes de signal a été appliquée afin de déterminer le débit massique des gouttelettes. L

Elaboration d’Indice composite de Développement du secteur bovin Laitier (IDL

Directory of Open Access Journals (Sweden)

K. KESSAB BELKHAYAT

2014-03-01

Full Text Available Plusieurs travaux de recherche ont été publiés sur la mesure du niveau de développement du secteur bovin laitier. Toutefois, aucun de ces travaux ne traite du secteur dans sa globalité. L’objectif du travail est de construire un tableau de bord du secteur bovin laitier à travers le développement d’un indice composite.Pour la construction de l’indice composite, 39 indicateurs du secteur bovin laitier sont identifiés dans le cadre conceptuel couvrant 8 dimensions. La collecte des données a concerné 41 variables composant les indicateurs, 37 pays et sur une période de 11 années (2000-2010. Après le traitement des données manquantes, la base de données complète est constituée de 21 indicateurs, de 23 pays sur 9 années. Un modèle a été développé pour la normalisation, la pondération des indicateurs puis pour le calcul de l’indice composite. Le test de robustesse est déroulé par le calcul du coefficient de corrélation de Pearson. Ce test a montré que de l’indice composite calculé selon 3 méthodes différentes de normalisation et de pondération est robuste. Les pays ont été classés selon leur l’IDL. Plusieurs axes d’analyse sont possibles à travers l’IDL notamment son évolution dans le temps, les points forts et les points faibles pour chaque pays et les leviers de développement du secteur.

Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

International Nuclear Information System (INIS)

Martinez C, E.

2011-01-01

One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-θ and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-θ, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm 2 s, at a height H 4 (239.07 cm) and angle 32.236 o in the core shroud and 4.00 E + 09 n/cm 2 s at a height H 4 and angle 35.27 o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

Experimentally guided Monte Carlo calculations of the atmospheric muon flux for interdisciplinary applications

International Nuclear Information System (INIS)

Mitrica, B.; Brancus, I.M.; Toma, G.; Bercuci, A.; Aiftimiei, C.; Wentz, J.; Rebel, H.

2004-01-01

Atmospheric muons are produced in the interactions of primary cosmic rays particle with Earth's atmosphere, mainly by the decay of pions and kaons generated in hadronic interactions. They decay further in electrons and positrons and electron and muon neutrinos. Being the penetrating cosmic rays component, the muons manage to pass entirely through the atmosphere and can pass even larger absorbers before they interact with the material at the Earth's surface, and due to cosmogenic production of isotopes by atmospheric muons, information of astrophysical, environmental and material research interest can be obtained. Up to now, mainly semi-analytical approximations have been used to calculate the muon flux for estimating the cosmogenic isotope production, necessary for different applications. Our estimation of the atmospheric muon flux is based on a Monte-Carlo simulation program CORSIKA, in which we simulate the development in the atmosphere of the extensive air showers, using different models for the description of the hadronic interaction. Atmospheric muons are produced in the interactions of primary cosmic rays particle with Earth's atmosphere, mainly by the decay of pions and kaons generated in hadronic interactions. They decay further in electrons and positrons and electron and muon neutrinos. Being the penetrating cosmic rays component, the muons manage to pass entirely through the atmosphere and can pass even larger absorbers before they interact with the material at the Earth's surface, and due to cosmogenic production of isotopes by atmospheric muons, information of astrophysical, environmental and material research interest can be obtained. Up to now, mainly semi-analytical approximations have been used to calculate the muon flux for estimating the cosmogenic isotope production, necessary for different applications. Our estimation of the atmospheric muon flux is based on a Monte-Carlo simulation program CORSIKA, in which we simulates the development in the

CERN : Nouveaux records lors d'une période d'exploitation au PS/SPS ; Un nouveau rôle pour les ISR ; Deuxième région d'intersection pour le système pp du SPS ; Détermination de la durée de vie du charme ; Calcul en Sicile et au CERN

CERN Multimedia

1979-01-01

CERN : Nouveaux records lors d'une période d'exploitation au PS/SPS ; Un nouveau rôle pour les ISR ; Deuxième région d'intersection pour le système pp du SPS ; Détermination de la durée de vie du charme ; Calcul en Sicile et au CERN

Effect of micrometric hot spots on surface temperature measurement and flux calculation in the middle and long infrared

International Nuclear Information System (INIS)

Delchambre, E; Counsell, G; Kirk, A

2009-01-01

The non-uniformity of the target temperature due to micrometric hot spots (Hermann et al 2004 Phys. Scr. T 111 98) is an explanation for the experimental fact that near-infrared measurements yield higher temperature values than mid-infrared measurements (Hildebrandt et al 2003 InfraMation 2003 Proc. (Las Vegas, USA, October 2003), Delchambre et al 2005 J. Nucl. Mater. 337-339 1069). The issue of micrometric hot spot disturbance in the surface temperature (T surf ) measurement and heat load calculation is addressed in this paper. The theoretical investigation at 3, 5 and 12 μm and experiments in the range 3.5-5 μm indicate that the surface state can play an important role in the non-uniform heating surface and consequently in the overestimation of the bulk temperature. The contribution of the hot spots to temperature measurements and flux calculations has been simulated at different wavelengths. Calculations show that (1) the overestimation of the bulk temperature decreases with the wavelength and (2) the overestimation depends on the temperature difference, ΔT, between the bulk and the micrometric hot spots. In addition, experiments have been carried out in order to compare the flux calculations at different wavelengths on different graphite (polished, dusty). The results obtained are very sensitive to the surface state pointing out the difficulties in improving the heat flux calculation model, since the surface state can change during the plasma discharges. This paper shows that the problem of non-homogenous surface temperature can be significantly diminished on working at longer wavelengths.

Influence of the flux axial form on the conversion rate and duration of cycle between recharging for ThPu and U{sub nat} fuels in CANDU reactors; Influence de la forme axiale du flux sur le taux de conversion et la duree du cycle entre rechargements pour du combustible ThPu et U{sub nat} dans les reacteurs CANDU

Energy Technology Data Exchange (ETDEWEB)

Chambon, Richard [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier / CNRS-IN2P3, 53 Avenue des Martyrs, F-38026 Grenoble (France)

2007-01-15

To face the increasing world power demand the world nuclear sector must be continuously updated and developed as well. Thus reactors of new types are introduced and advanced fuel cycles are proposed. The technological and economic feasibility and the transition of the present power park to a renewed park require thorough studies and scenarios, which are highly dependent on the reactor performances. The conversion rate and cycle span between recharging are important parameters in the scenarios studies. In this frame, we have studied the utilisation of thorium in the CANDU type reactors and particularly the influence of axial form of the flux, i.e. of the recharging mode, on the conversion rate and duration of the cycle between recharging. The results show that up to a first approximation the axial form of the flux resulting from the neutron transport calculations for assessing the conversion rate is not necessary to be taken into account. However the time span between recharging differs up to several percents if the axial form of the flux is taken into consideration in transport calculations. Thus if the burnup or the recharging frequency are parameters which influence significantly the deployment scenarios of a nuclear park an approach more refined than a simple transport evolution in a typical cell/assembly is recommended. Finally, the results of this study are not more general than for the assumed conditions but they give a thorough calculation method valid for any recharging/fuel combination in a CANDU type reactor.

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

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

DETERMINATION EXPERIMENTALE DU COEFFICIENT DE DISPERSION D’UNE MATRICE DE SOL

Directory of Open Access Journals (Sweden)

R MAOUI

2005-06-01

Full Text Available La caractérisation des interactions entre une espèce chimique et une matrice de sol est généralement considérée dans les études des phénomènes de transport. Ce type de réaction est assujetti à l’estimation du coefficient de dispersion D. Ce coefficient représente le gradient de concentration de l’élément et le flux de transport ; il dépend de la distribution de la vitesse, de la porosité du milieu et des courants du liquide. Le présent travail décrit le procédé expérimental et la technique associée utilisés en vue de quantifier le coefficient D de deux traceurs radioactifs. Les résultats obtenus à l’échelle du laboratoire sont très satisfaisants et restent en conformité avec les résultats publiés

Neutron flux monitor

International Nuclear Information System (INIS)

Oda, Naotaka.

1993-01-01

The device of the present invention greatly saves an analog processing section such as an analog filter and an analog processing circuit. That is, the device of the present invention comprises (1) a neutron flux detection means for detecting neutron fluxed in the reactor, (2) a digital filter means for dividing signals corresponding to the detected neutron fluxes into predetermined frequency band regions, (3) a calculation processing means for applying a calculation processing corresponding to the frequency band regions to the neutron flux detection signals divided by the digital filter means. With such a constitution, since the neutron detection signals are processed by the digital filter means, the accuracy is improved and the change for the property of the filter is facilitated. Further, when a neutron flux level is obtained, a calculation processing corresponding to the frequency band region can be conducted without the analog processing circuit. Accordingly, maintenance and accuracy are improved by greatly decreasing the number of parts. Further, since problems inherent to the analog circuit are solved, neutron fluxes are monitored at high reliability. (I.S.)

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

Science.gov (United States)

Khattab, K; Sulieman, I

2009-04-01

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

Study and comparison of some methods for calculating the transmission factor of a potential barrier in quantum mechanics (1963); Etude et comparaison de quelques methodes de calcul du facteur de transmission d'une barriere de potentiel en mecanique quantique (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

The author formulates two accurate methods for the calculation of the transmission coefficient of a one-dimensional potential barrier. The principles of these methods are. expressed in a symmetrical form with respect to the two sides of the potential barrier; this constitutes a proof of the fact that the transmission coefficient is path direction independent. The numerical application is carried out on several examples and the results are compared to those provided by the WKB method. (author) [French] L'auteur formule deux methodes exactes pour calculer le coefficient de transmission d'une barriere de potentiel unidimensionnelle. Les principes de ces methodes s'enoncent sous forme symetrique par rapport aux deux cotes de la barriere de potentiel, ce qui constitue une demonstration du fait que le coefficient de transmission est independant du sens de parcours. L'application numerique est faite sur quelques exemples et les resultats sont compares a ceux fournis par la methode B KW. (auteur)

Neutron flux monitoring device

International Nuclear Information System (INIS)

Shimazu, Yoichiro.

1995-01-01

In a neutron flux monitoring device, there are disposed a neutron flux measuring means for outputting signals in accordance with the intensity of neutron fluxes, a calculation means for calculating a self power density spectrum at a frequency band suitable to an object to be measured based on the output of the neutron flux measuring means, an alarm set value generation means for outputting an alarm set value as a comparative reference, and an alarm judging means for comparing the alarm set value with the outputted value of the calculation means to judge requirement of generating an alarm and generate an alarm in accordance with the result of the judgement. Namely, the time-series of neutron flux signals is put to fourier transformation for a predetermined period of time by the calculation means, and from each of square sums for real number component and imaginary number component for each of the frequencies, a self power density spectrum in the frequency band suitable to the object to be measured is calculated. Then, when the set reference value is exceeded, an alarm is generated. This can reliably prevent generation of erroneous alarm due to neutron flux noises and can accurately generate an alarm at an appropriate time. (N.H.)

Formulation of the energetic spectral distribution of in pile neutron flux (energies greater than a few hundred electron volts) (1963); Formulation des repartitions spectrales energetiques de flux neutroniques en pile (energies superieures a quelques centaines d'electrons-volts) (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

In the present report an expression is deduced for the spectral distribution of flux of over a few hundred electron volts; it is valid for most cases of thermal neutron reactors. This expression is: {psi}(E) = K [{psi}{sub o}(E) + h {psi}{sub e}(E)] {psi}{sub o}(E) is the so-called 'homogeneous' constituent of the flux; it corresponds approximately to the case of an infinite homogeneous medium; it is of the type: Y(V - E) e{sup (b{radical}}{sup E)}/E + Y(E-V) F E{sup {upsilon}} e{sup -{beta}}{sup E} The parameters V and F are such that {psi}{sub o}(E) and its derivative are continuous at the junction energy V. {psi}{sub e}(E) is the 'heterogeneous' constituent of the flux; it is of the type: E{sup {upsilon}} e{sup -{beta}}{sup E}. The various parameters of {psi}(E) are on the one hand characteristic of the nature of the reactor moderator, and on the other hand determined by a resonant flux measurement and one, or possibly two, measurements using a fast neutron threshold detector. We have been led furthermore to define an expression for the threshold reaction section which is more exact than the conventional transition function. (author) [French] Il est etabli, dans le present rapport, une formulation {psi}(E) des repartitions spectrales de flux au-dessus de quelques centaines d'electron-s volts, valable dans la majeure partie des cas de piles dites a neutrons thermiques. Cette formulation s'exprime: {psi}(E) = K [{psi}{sub o}(E) + h {psi}{sub e}(E)] {psi}{sub o}(E) est la composante dite 'homogene' du flux; elle correspond a peu pres au cas d'un milieu infini homogene; elle est du type: Y(V - E) e{sup (b{radical}}{sup E)}/E + Y(E-V) F E{sup {upsilon}} e{sup -{beta}}{sup E} les parametres V et F sont tels que {psi}{sub o}(E) et sa derivee soient continues a l'energie de jonction V. {psi}{sub e} est la composante dite 'heterogene' du flux ; elle est du type: E{sup {upsilon}} e{sup -{beta}}{sup E}. Les differents parametres de {psi}(E) sont, d'une part

Calculation of the flux attenuation and multiple scattering correction factors in time of flight technique for double differential cross section measurements

International Nuclear Information System (INIS)

Martin, G.; Coca, M.; Capote, R.

1996-01-01

Using Monte Carlo method technique , a computer code which simulates the time of flight experiment to measure double differential cross section was developed. The correction factor for flux attenuation and multiple scattering, that make a deformation to the measured spectrum, were calculated. The energy dependence of the correction factor was determined and a comparison with other works is shown. Calculations for Fe 56 at two different scattering angles were made. We also reproduce the experiment performed at the Nuclear Analysis Laboratory for C 12 at 25 celsius degree and the calculated correction factor for the is measured is shown. We found a linear relation between the scatter size and the correction factor for flux attenuation

A one-dimensional, one-group absorption-production nodal method for neutron flux and power distributions calculations

International Nuclear Information System (INIS)

Ferreira, C.R.

1984-01-01

It is presented the absorption-production nodal method for steady and dynamical calculations in one-dimension and one group energy. It was elaborated the NOD1D computer code (in FORTRAN-IV language). Calculations of neutron flux and power distributions, burnup, effective multiplication factors and critical boron concentration were made with the NOD1D code and compared with results obtained through the CITATION code, which uses the finite difference method. The nuclear constants were produced by the LEOPARD code. (M.C.K.) [pt

Verification of Monte Carlo calculations of the neutron flux in the carousel channels of the TRIGA Mark II reactor, Ljubljana

International Nuclear Information System (INIS)

Jacimovic, R.; Maucec, M.; Trkov, A.

2002-01-01

In this work experimental verification of Monte Carlo neutron flux calculations in the carousel facility (CF) of the 250 kW TRIGA Mark II reactor at the Jozef Stefan Institute is presented. Simulations were carried out using the Monte Carlo radiation-transport code, MCNP4B. The objective of the work was to model and verify experimentally the azimuthal variation of neutron flux in the CF for core No. 176, set up in April 2002. '1'9'8Au activities of Al-Au(0.1%) disks irradiated in 11 channels of the CF covering 180'0 around the perimeter of the core were measured. The comparison between MCNP calculation and measurement shows relatively good agreement and demonstrates the overall accuracy with which the detailed spectral characteristics can be predicted by calculations.(author)

REMTWO - WRS system module number 299 for calculating the removal flux in a slab shield

International Nuclear Information System (INIS)

Grimstone, M.J.

1978-06-01

The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the uncollided flux due to a distributed source in a one-dimensional plane geometry system. A method is used in which the flux is approximated by the sum of diffusion equation solutions. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)

Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

Response of actinides to flux changes in high-flux systems

International Nuclear Information System (INIS)

Sailor, W.C.

1993-01-01

When discussing the transmutation of actinides in accelerator-based transmutation of waste (ATW) systems, there has been some concern about the dynamics of the actinides under high transient fluxes. For a pure neptunium feed, it has been estimated that the 238 Np/ 237 Np ratio increase due to an increasing flux may lead to an unstable, positive reactivity growth. In this analysis, a perturbation method is used to calculate the response of the entire set of actinides in a general way that allows for more species than just neptunium. The time response of the system can be calculated; i.e., a plot of fuel composition and reactivity versus time after a change in flux can be made. The effects of fission products can also be included. The procedure is extremely accurate on short time scales (∼ 1000 s) for the flux levels we contemplate. Calculational results indicate that the reactivity insertions are always smaller than previously estimated

Comparisons of Measured and Calculated Neutron Fluxes in Laminated iron and Heavy Water

Energy Technology Data Exchange (ETDEWEB)

Aalto, E

1964-10-15

Measurements of neutron fluxes have been performed in configurations depicting the regions extending radially and axially outwards from the core of a PHWR reactor in order to test the accuracy of the available methods in shield design on thin alternating laminae of Fe and D{sub 2}O. A 'dry' experimental set-up was constructed, i.e. the D{sub 2}O was contained in flat tanks made of Al. The first set of measurements was performed through solid Fe and D{sub 2}O layers, and only the results of these experiments are described in this report. The set-up allowed measurements also in a mock-up of a reactor top penetrated by D{sub 2}O or air-filled channels (to be reported later). The results are compared to fluxes calculated by the British 18-group removal-diffusion method and by the NRN method developed at AE. The results show that the values predicted may be expected to be within a factor of 2 from the true values in most cases. The predicted relative flux distributions follow the observed ones with a very good accuracy in spite of the apparent misuse of diffusion theory for the thin regions in question. Finally, it is shown that the predicted change in the fast spectrum while penetrating these set-ups should be confirmable with certain threshold detectors.

Les flux de matières dissoutes et particulaires des Andes de Bolivie vers le rio Madeira en Amazonie brésilienne

Directory of Open Access Journals (Sweden)

1995-01-01

Full Text Available Le Rio Madeira est le principal affluent méridional de l’Amazone et le deuxième tributaire andin du bassin. L’utilisation des données obtenues en Bolivie par le PHICAB (1, et au Brésil sur le réseau du DNAEE (2, a permis de suivre l’évolution des flux de matières dissoutes et particulaires, depuis les Andes jusqu’à l’Amazone. Les flux de matières en solution [36 106 t an-1 à Villabella (3] augmentent progressivement d’amont en aval avec les débits. Les flux de matières en suspension diminuent du piedmont andin jusqu’à Villabella (258 106 t an-1 du fait d’une forte sédimentation. Ensuite, les différences notables observées au Brésil posent clairement le problème des modes d’échantillonnage et de calcul des flux sédimentaires. LOS FLUJOS DE MATERIA DISUELTA Y PARTICULAR DESDE LOS ANDES DE BOLIVIA HASTA EL RÍO MADEIRA EN EL AMAZONAS DE BRASIL. El río Madeira es el principal afluente meridional del Amazonas, y el segundo tributario andino de la cuenca. La utilización de los datos obtenidos en Bolivia por el PHICAB (1, y en el Brasil sobre la red del DNAEE (2 permite conocer la evolución de los flujos de materia disuelta y particular, desde los Andes hasta el Amazonas. Los aportes en materia disuelta [36 106 t año-1 en Villabella (3] aumentan progresivamente de aguas arriba para aguas abajo, con el caudal. Los flujos de materias en suspensión disminuyen desde el Piedemonte hasta Villabella (258 106 t año-1 por causa de una fuerte sedimentación. Después, las diferencias notables observadas en el Brasil colocan claramente el problema de los tipos de muestreo y de cálculo de los flujos de sedimentos. THE DISSOLVED AND SEDIMENT YIELDS FROM THE ANDES OF BOLIVIA TO THE MADEIRA RIVER IN THE BRAZILIAN AMAZON. The Madeira river is the main Southern tributary of the Amazon, and the second Andean tributary of the drainage basin. The use of Bolivian data from the PHICAB (1, and Brazilian data from the DNAEE (2, allow to

A PC-program for the calculation of neutron flux and element contents using the ki-method of neutron activation analysis

International Nuclear Information System (INIS)

Boulyga, E.G.; Boulyga, S.F.

2000-01-01

A computer program is described, which calculates the induced activities of isotopes after irradiation in a known neutron field, thermal and epithermal neutron fluxes from the measured induced activities and from nuclear data of 2-4 monitor nuclides as well as the element concentrations in samples irradiated together with the monitors. The program was developed for operation in Windows 3.1 (or higher). The application of the program for neutron activation analysis allows to simplify the experimental procedure and to reduce the time. The program was tested by measuring different types of standard reference materials at the FRJ-2 (Research Centre, Juelich, Germany) and Triga (University Mainz, Germany) reactors. Comparison of neutron flux parameters calculated by this program with those calculated by a VAX program developed at the Research Centre, Juelich was done. The results of testing seem to be satisfactory. (author)

Variation du coefficient de collage de l'argent et de l'or durant leur condensation sur film de MgO

Science.gov (United States)

Desrousseaux, G.; Carlan, A.; Jiang, Z.

1993-10-01

The incidence rate dependence of the sticking coefficient during the growth of gold (or silver) condensate on MgO film is experimentally investigated for a substrate kept at room temperature. This MgO substrate is deposited on quartz monitor by evaporation under UHV from a Knudsen cell. Then, from another cell, the flux of metal atoms impinges on the MgO film which covers the quartz. Under this flux of incident metal atoms, the changes (respectively θ and f) of the temperature and of the frequency of the quartz oscillator are then simultaneously recorded. Both continuous measurements enable us to calculate, at different times t, the sticking coefficient η (t) defined as the ration between the condensed metal rate q and the incident rate R. We use the results to plot η versus t and examine the slope of η-rise until the time t at which η (t) reaches unity. The progression of η (t) reaches unity. The progression of η (t) from zero to unity during the metal deposition is explained by assuming that the nucleation on preferred sites with capture of adatoms at the edge of stable germs is the prevailing condensation mechanism at the initial stage of the condensation. Our experimental results could be fitted better by assuming an exponential growth of the site occupation than an increase of the number density n_s(t) with (Rt)^{1/3}. On étudie expérimentalement la dépendance que peut présenter, par rapport au flux métallique incident, la progression vers l'unité du coefficient de collage. Pour cela, on suit cette progression pour l'or et l'argent condensés sur film de MgO à température proche de l'ambiante. Le substrat de MgO est déposé sur un quartz de pesée par évaporation sous ultra-vide à partir d'une cellule de Knudsen. Une autre cellule est utilsée pour déposer l'un de ces métaux sur le film de MgO couvrant le quartz. Les variations θ et f de la température et de la fréquence d'oscillations du quartz de pesée sous le flux d'atomes incidents

Calculation of heat fluxes induced by radio frequency heating on the actively cooled protections of ion cyclotron resonant heating (ICRH) and lower hybrid (LH) antennas in Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Ritz, G., E-mail: Guillaume.ritz@gmail.com [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Corre, Y., E-mail: Yann.corre@cea.fr [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Rault, M.; Missirlian, M. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint Paul-lez-Durance (France); Martinez, A.; Ekedahl, A.; Colas, L.; Guilhem, D.; Salami, M.; Loarer, T. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France)

2013-10-15

Highlights: ► The heat flux generated by radiofrequency (RF) heating was calculated using Tore Supra's heating antennas. ► The highest heat flux value, generated by ions accelerated in RF-rectified sheath potentials, was 5 MW/m{sup 2}. ► The heat flux on the limiters of antennas was in the same order of magnitude as that on the toroidal pumping limiter. -- Abstract: Lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) are recognized as important auxiliary heating and current drive methods for present and next step fusion devices. However, these radio frequency (RF) systems generate a heat flux up to several MW/m{sup 2} on the RF antennas during plasma operation. This paper focuses on the determination of the heat flux deposited on the lateral protections of the RF antennas in Tore Supra. The heat flux was calculated by finite element method (FEM) using a model of the lateral protection. The FEM calculation was based on surface temperature measurements using infrared cameras monitoring the RF antennas. The heat flux related to the acceleration of electrons in front of the LHCD grills (LHCD active) and to the acceleration of ions in RF-rectified sheath potentials (ICRH active) were calculated. Complementary results on the heat flux related to fast ions (ICRH active with a relatively low magnetic field) are also reported in this paper.

Core calculations of JMTR

Energy Technology Data Exchange (ETDEWEB)

Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-03-01

In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

Actes des 5èmes Journées Scientifiques du GDR3544 Sciences du Bois. Journées Annuelles du GDR 3544 Sciences du Bois

OpenAIRE

CHAPLAIN, Myriam; CARE, Sabine; GRIL, Joseph

2016-01-01

Le Groupement de Recherche en Sciences du bois (GDR3544 Sciences du Bois) a été créé en 2012 par le CNRS et renouvelé en 2016 pour 5 ans. La mission de ce groupement est : (1) de structurer la recherche sur le bois en France pour lui donner une visibilité nationale, (2) de contribuer au développement de la formation en sciences du bois et (3) de servir de relai aux réseaux internationaux de sciences du bois. Les 5èmes journées annuelles du GDR Bois ont été organisées à Bordeaux, au domaine du...

FluxPyt: a Python-based free and open-source software for 13C-metabolic flux analyses.

Science.gov (United States)

Desai, Trunil S; Srivastava, Shireesh

2018-01-01

13 C-Metabolic flux analysis (MFA) is a powerful approach to estimate intracellular reaction rates which could be used in strain analysis and design. Processing and analysis of labeling data for calculation of fluxes and associated statistics is an essential part of MFA. However, various software currently available for data analysis employ proprietary platforms and thus limit accessibility. We developed FluxPyt, a Python-based truly open-source software package for conducting stationary 13 C-MFA data analysis. The software is based on the efficient elementary metabolite unit framework. The standard deviations in the calculated fluxes are estimated using the Monte-Carlo analysis. FluxPyt also automatically creates flux maps based on a template for visualization of the MFA results. The flux distributions calculated by FluxPyt for two separate models: a small tricarboxylic acid cycle model and a larger Corynebacterium glutamicum model, were found to be in good agreement with those calculated by a previously published software. FluxPyt was tested in Microsoft™ Windows 7 and 10, as well as in Linux Mint 18.2. The availability of a free and open 13 C-MFA software that works in various operating systems will enable more researchers to perform 13 C-MFA and to further modify and develop the package.

Influence of the heat flux and of the gas on heat transfer and friction coefficients in a smooth cylindrical tube; Influence du flux de chaleur et de la nature du gaz sur les coefficients d'echange et le frottement dans un tube cylindrique lisse

Energy Technology Data Exchange (ETDEWEB)

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

1962-12-15

paroi peut atteindre 430 deg C, le flux de chaleur atteignant 16 watts/cm{sup 2}. Les nombres de Reynolds Re{sub m}, de Margoulis M{sub m} et le coefficient de frottement f sont calcules en evaluant les proprietes physiques du gaz a la temperature moyenne T{sub m}. Pour un nombre de Reynolds Re{sub m} donne, on observe une diminution de M{sub m} et de f lorsque le flux de chaleur augmente: elle peut atteindre 10 pour cent dans les essais presentes. Une formulation est proposee qui permet de traduire cet effet au moyen d'un parametre de flux (T{sub m} - T{sub m}) / T{sub p} utilise comme terme correctif (T{sub p} = temperature de la paroi). Les formules de correlation proposees sont: M{sub m} = 0,0168 Re{sub m}{sup -0,18} P{sub m}{sup -0,6} (1 - 0,4 [(T{sub p} - T{sub m}) / T{sub p}]) pour l'air f = f{sub 0} (1 - 0,25 [(T{sub p} - T{sub m}) / T{sub p}]) pour l'air M{sub m} 0,0171 Re{sub m}{sup -0,18} P{sub m}{sup -0,6} (1 - 0,2 [(T{sub p} - T{sub m}) / T{sub p}]) pour le CO{sub 2} f = f{sub 0} (1 - 0,20 [(T{sub p} - T{sub m}) / T{sub p}]) pour le gaz carbonique avec f{sub 0} coefficient de frottement en ecoulement isotherme. (auteur)

Anomalous high-frequency wave activity flux preceding anomalous changes in the Northern polar jet

Science.gov (United States)

Nakamura, Mototaka; Kadota, Minoru; Yamane, Shozo

2010-05-01

Anomalous forcing by quasi-geostrophic (QG) waves has been reported as an important forcing factor in the Northern Annular Mode (NAM) in recent literatures. In order to shed a light on the dynamics of the NAM from a different angle, we have examined anomalous behavior of the winter jets in the upper troposphere and stratosphere by focusing our diagnosis on not the anomalous geopotential height (Z) itself, but on the anomalous change in the Z (dZ) between two successive months and preceding transient QG wave activity flux during the cold season. We calculated EOFs of dZ between two successive months at 150hPa for a 46-year period, from 1958 to 2003, using the monthly mean NCEP reanalysis data. We then formed anomaly composites of changes in Z and the zonal velocity (U), as well as the preceding and following wave activity flux, Z, U, and temperature at various heights, for both positive and negative phases of the first EOF. For the wave forcing fields, we adopted the diagnostic system for the three-dimensional QG transient wave activity flux in the zonally-varying three-dimensional mean flow developed by Plumb (1986) with a slight modification in its application to the data. Our choice of the Plumb86 is based on the fact that the winter mean flow in the Northern Hemisphere is characterized by noticeable zonal asymmetry, and has a symbiotic relationship with waves in the extra-tropics. The Plumb86 flux was calculated for high-frequency (period of 2 to 7 days) and low-frequency (period of 10 to 20 days) waves with the ultra-low-frequency (period of 30 days or longer) flow as the reference state for each time frame of the 6 hourly NCEP reanalysis data from 1958 to 2003. By replacing the mean flow with the ultra-low-frequency flow in the application of the Plumb86 formula, the flux fields were calculated as time series at 6 hour intervals. The time series of the wave activity flux was then averaged for each month. The patterns of composited anomalous dZ and dU clearly

Effets de la microdose sur la production du niébé, du mil et du ...

African Journals Online (AJOL)

Effets de la microdose sur la production du niébé, du mil et du sorgho en fonction la toposéquence. Fatimata Saba, Sibiri Jean Baptiste Taonda, Idriss Serme, Alimata A. Bandaogo, Augustin P. Sourwema, Adama Kabre ...

Compilation of the calculation elements of the electronic equipments reliability; Recueil des elements de calcul de la fiabilite des equipements electroniques

Energy Technology Data Exchange (ETDEWEB)

Lefevre, R.; D' Harcourt, A.; Dupuy, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The aim of this work is to allow the practical execution of the estimated calculation of the electronic devices reliability and to standardize the source and the approach of the calculations while giving a systematic character to their execution mode. The systematic character of the calculations allows a comparison of the reliability of different materials and a rapid control of the calculations validity; at last, it brings out the use conditions of all the components of a set. A reliability calculation made according to the method described here reveals: -components badly used -the relative influence on the reliability of the set, of a component or a components group taking into account of the number, of the characteristics and of the use of these ones. At last, the results of the calculation allows to organize the exploitation (availability) and the maintenance (staff, stock of components) of the materials. The failure rates given in this book are only relating to the components service-life and do not correspond to precocious failures. (authors) [French] L'objet du present recueil est d'une part de permettre l'execution pratique du calcul previsionnel de la fiabilite des equipements electroniques et d'autre part d'uniformiser l'origine et la presentation de calculs en donnant un caractere systematique a leur mode d'execution. Le caractere systematique des calculs permet une comparaison de la fiabilite de differents materiels et un controle rapide de la validite des calculs; enfin il met en relief les conditions d'utilisation de tous les composants d'un ensemble. Un calcul de fiabilite realise selon la methode preconise par ce document permet la mise en evidence: -des composants mal utilises -de l'influence relative sur la fiabilite de l'ensemble, d'un composant ou d'un groupe de composants compte tenu du nombre, des caracteristiques et de l'utilisation de ceux-ci. Enfin, les resultats du

OpenFLUX: efficient modelling software for 13C-based metabolic flux analysis

Directory of Open Access Journals (Sweden)

Nielsen Lars K

2009-05-01

Full Text Available Abstract Background The quantitative analysis of metabolic fluxes, i.e., in vivo activities of intracellular enzymes and pathways, provides key information on biological systems in systems biology and metabolic engineering. It is based on a comprehensive approach combining (i tracer cultivation on 13C substrates, (ii 13C labelling analysis by mass spectrometry and (iii mathematical modelling for experimental design, data processing, flux calculation and statistics. Whereas the cultivation and the analytical part is fairly advanced, a lack of appropriate modelling software solutions for all modelling aspects in flux studies is limiting the application of metabolic flux analysis. Results We have developed OpenFLUX as a user friendly, yet flexible software application for small and large scale 13C metabolic flux analysis. The application is based on the new Elementary Metabolite Unit (EMU framework, significantly enhancing computation speed for flux calculation. From simple notation of metabolic reaction networks defined in a spreadsheet, the OpenFLUX parser automatically generates MATLAB-readable metabolite and isotopomer balances, thus strongly facilitating model creation. The model can be used to perform experimental design, parameter estimation and sensitivity analysis either using the built-in gradient-based search or Monte Carlo algorithms or in user-defined algorithms. Exemplified for a microbial flux study with 71 reactions, 8 free flux parameters and mass isotopomer distribution of 10 metabolites, OpenFLUX allowed to automatically compile the EMU-based model from an Excel file containing metabolic reactions and carbon transfer mechanisms, showing it's user-friendliness. It reliably reproduced the published data and optimum flux distributions for the network under study were found quickly ( Conclusion We have developed a fast, accurate application to perform steady-state 13C metabolic flux analysis. OpenFLUX will strongly facilitate and

Calculation of the magnetic flux density distribution in type-II superconductors with finite thickness and well-defined geometry

International Nuclear Information System (INIS)

Forkl, A.; Kronmueller, H.

1995-01-01

The distribution of the critical current density j c (r) in hard type-II superconductors depends strongly on their sample geometry. Rules are given for the construction of j c (r). Samples with homogeneous thickness are divided into cakelike regions with a unique current direction. The spatial magnetic flux density distribution and the magnetic polarization of such a cakelike unit cell with homogeneous current density are calculated analytically. The magnetic polarization and magnetic flux density distribution of a superconductor in the mixed state is then given by an adequate superposition of the unit cell solutions. The theoretical results show good agreement with magneto-optically determined magnetic flux density distributions of a quadratic thin superconducting YBa 2 Cu 3 O 7-x film. The current density distribution is discussed for several sample geometries

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

Science.gov (United States)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

Directory of Open Access Journals (Sweden)

Brovchenko Mariya

2017-01-01

Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

L'administration du travail et la production du droit du travail (1906-1960). : Note de synthèse du rapport de recherche

OpenAIRE

Le Crom , Jean-Pierre

2007-01-01

Synthèse d'un rapport de recherche dont l'objectif est d'appréhender le rôle de l'administration centrale du travail dans la production du droit du travail. Deux dimensions sont explorées : le profil des rédacteurs et l'organisation des structures.

Calculation of electromagnetic torque for synchronous motor with modulated magnetic flux and smooth harmonic rotor

Science.gov (United States)

Shevchenko, A. F.; Shevchenko, L. G.

2017-10-01

Results of the electromagnetic torque calculation for the synchronous motor with modulated magnetic flux and a smooth harmonic rotor are presented in this paper. The value of the torque is determined from the electromagnetic forces, which appear due to interaction of magnetic field in the gap with the rotor surface elements. The obtained analytical expression makes it possible to determine easily the electromagnetic torque for the considered motor in the MathCAD environment.

Improvement of the neutron flux calculations in thick shield by conditional Monte Carlo and deterministic methods

International Nuclear Information System (INIS)

Ghassoun, Jillali; Jehoauni, Abdellatif

2000-01-01

In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)

Effet du Pediococcus acidilactici sur le bilan lipidique sanguin du ...

African Journals Online (AJOL)

Les résultats relatifs aux performances zootechniques ont montré que l'addition du probiotique a amélioré significativement le gain de poids pendant la phase de croissance se traduisant par un indice de consommation meilleur. Les dosages du cholestérol total, des triglycérides, du HDL et du LDL ont été déterminés à la ...

RIFIFI: Analytical calculation method of the critical condition and flux in a varied regions reactor by two-group theory and one dimension developed for the Mercury-Ferranti computer; Rififi: methode de calcul analytique de la condition critique et des flux d'une pile a regions variees en theorie a deux groupes et a une dimension programmee pour le calculateur electronique Mercury (Ferranti)

Energy Technology Data Exchange (ETDEWEB)

Amouyal, A; Bacher, P; Lago, B; Mengin, F L; Parker, E [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

The calculation method presented in this report has been developed for the Mercury-Ferranti computer of the C.E.N.S. This calculation method allows to resolve the diffusion equations and continuity equations of flux and flow with two groups of neutrons and one dimension in spherical, cylindrical and linear geometry. In the cylindrical and linear configurations, we can take the height and extrapolated radius into account. The critical condition can be realised by varying linearly one or more parameters: k{sub {infinity}}, medium frontier, height or extrapolated radius. The calculation method enables also to calculate the flux, adjoint flux and various integrals. In the first part, it explains what is needed to know before using the method: data presentation, method possibilities, results presentation with some information about restrictions, accuracy and calculation time. The complete formulation of the calculation method is given in the second part. (M.P.)

Alimentation du nouveau-ne et du nourrisson dans la region ...

African Journals Online (AJOL)

Alimentation du nouveau-ne et du nourrisson dans la region centrale du togo : pratiques familiales et communautaires avant la mise en oeuvre de la strategie « prise en charge integree des maladies de l'enfant »

Aerosol-Induced Radiative Flux Changes Off the United States Mid-Atlantic Coast: Comparison of Values Calculated from Sunphotometer and In Situ Data with Those Measured by Airborne Pyranometer

Science.gov (United States)

Russell, P. B.; Livingston, J. M.; Hignett, P.; Kinne, S.; Wong, J.; Chien, A.; Bergstrom, R.; Durkee, P.; Hobbs, P. V.

2000-01-01

The Tropospheric Aerosol Radiative Forcing Observational Experiment (TARFOX) measured a variety of aerosol radiative effects (including flux changes) while simultaneously measuring the chemical, physical, and optical properties of the responsible aerosol particles. Here we use TARFOX-determined aerosol and surface properties to compute shortwave radiative flux changes for a variety of aerosol situations, with midvisible optical depths ranging from 0.06 to 0.55. We calculate flux changes by several techniques with varying degrees of sophistication, in part to investigate the sensitivity of results to computational approach. We then compare computed flux changes to those determined from aircraft measurements. Calculations using several approaches yield downward and upward flux changes that agree with measurements. The agreement demonstrates closure (i.e. consistency) among the TARFOX-derived aerosol properties, modeling techniques, and radiative flux measurements. Agreement between calculated and measured downward flux changes is best when the aerosols are modeled as moderately absorbing (midvisible single-scattering albedos between about 0.89 and 0.93), in accord with independent measurements of the TARPOX aerosol. The calculated values for instantaneous daytime upwelling flux changes are in the range +14 to +48 W/sq m for midvisible optical depths between 0.2 and 0.55. These values are about 30 to 100 times the global-average direct forcing expected for the global-average sulfate aerosol optical depth of 0.04. The reasons for the larger flux changes in TARFOX include the relatively large optical depths and the focus on cloud-free, daytime conditions over the dark ocean surface. These are the conditions that produce major aerosol radiative forcing events and contribute to any global-average climate effect.

Calculation of the dependence on the Moon and Mars γ-quantum flux on the relief and distance to the surface

International Nuclear Information System (INIS)

Surkov, Yu.A.; Noskaleva, L.P.; Manvelyan, O.S.

1978-01-01

The dependence of the gamma quantum flux on height over a planet, area over which the gamma radiation is ''collected'', and surface relief is calculated. The effect of the planet atmosphere on detected gamma radiation is considered. If the specific power of gamma-quantum sources is known, the results obtained allow to determine for any height over a planet the gamma-quantum flux due to the planet rock and its atmosphere radiations, as well as the detector spatial resolution

Portage vaginal du streptocoque du groupe B chez la femme ...

African Journals Online (AJOL)

Introduction: le streptocoque du groupe B est le principal agent impliqué dans les infections materno-fœtales, les septicémies et les méningites du nouveau-né à terme. L'objectif est de déterminer le taux de portage maternel du streptocoque du groupe B (SGB) à terme. Méthodes: un prélèvement vaginal a été réalisé de ...

Chaînes logistiques et consommation d'énergie : cas du yaourt et du jean

OpenAIRE

Rizet , C.; Keita , Basile

2005-01-01

Contrat INRETS/ADEME n°: 0203034; Rapport de recherche; Pour analyser l'efficacité énergétique et les émissions de GES des chaînes logistiques, on compare différentes chaînes aboutissant aux mêmes consommateurs. Deux produits sont étudiés : le yaourt, caractéristique de la chaîne du froid et des flux tendus et le blue jean, intégré au marché mondial, tant pour sa matière première principale (le coton) que pour les différentes étapes de sa fabrication. Les chaînes étudiées pour le jean différe...

Study of the influence of the fast neutron spectrum on the production of defects in solids and liquids; Etude de l'influence du spectre des neutrons rapides sur la creation de defauts dans les solides et les liquides

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

In the first part of this work a calculation has been made of the number of defects formed in graphite by a given neutron flux having various spectral distributions. The defect formation function is that of KINCHIN and PEASE; its formulation is briefly given. An efficiency function is then defined for a fast neutron spectrum. This defects produced in a light water reactor and those produced in a graphite reactor. Finally an application of this method is given for comparing the defect forming tendency in graphite in the case of the reactor Melusine and of the reactor G-2 and G-3. In the second part are calculated the integrals for the energy release brought about by fast neutrons in carbon oxygen and hydrogen. In a region of 25 cm around the core of a swimming-pool type reactor these energy release integrals are approximately proportional to the neutron flux above 1 MeV. The determination of the energy released as a result of the passage of neutrons in organic liquids can therefore be reduced to the measurement of the flux above 1 MeV for the real spectral distribution. A calorimetric verification has been carried out in the case of water. (authors) [French] Dans la premiere partie de cette etude, on a calcule le nombre de defauts crees dans le graphite par un flux de neutrons donne, pris sous differents spectres. On a ainsi fait les calculs pour 4 spectres de pile. piscine et 2 spectres de pile au graphite. La fonction creation de defauts est celle de KINCHIN et PEASE. On rappelle brievement sa formulation. Puis on definit une fonction efficacite d'un spectre de neutrons rapides. Cette fonction permet alors d'etablir la relation entre defauts crees dans une pile a eau legere et defauts crees dans une pile au graphite. Enfin, une application de cette methode est donnee pour comparer l'aptitude a creer les defauts dans le graphite par la pile MELUSINE et les piles G-2 ou G-3. Dans la deuxieme partie, on a calcule les integrales de degagement d'energie provoque par

VIRGIN2007, Calculates Un-collided Neutron Flux and Neutron Reactions from Transmission in ENDF Format

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: VIRGIN calculates un-collided flux and reactions due to transmission of a mono-directional beam of neutrons through any thickness of material. In order to simulate an experimental measurement the results are given as integrals over energy tally groups (as opposed to point-wise in energy). IAEA0932/10: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: Virgin VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased in-core page size from 60,000 to 240,000. 2 - Method of solution: By taking the ratio of reactions to flux in each group an equivalent spatially dependent group averaged cross section is calculated. 3 - Restrictions on the complexity of the problem: The evaluated data must be in the ENDF/B format. However it must be linear-linear interpolable in energy-cross section between tabulated points. Since only cross sections (file 3) are used, this program will work on any version of ENDF/B

Dosimetry work and calculations in connection with the irradiation of large devices in the high flux materials testing reactor BR2

International Nuclear Information System (INIS)

De Raedt, C.; Leenders, L.; Tourwe, H.; Farrar, H. IV.

1982-01-01

For about fifteen years the high flux reactor BR2 has been involved in the testing of fast reactor fuel pins. In order to simulate the fast reactor neutron environment most devices are irradiated under cadmium screen, cutting off the thermal flux component. Extensive neutronic calculations are performed to help the optimization of the fuel bundle design. The actual experiments are preceded by irradiations of their mock-ups in BR02, the zero power model of BR2. The mock-up irradiations, supported by supplementary calculations, are performed for the determination of the main neutronic characteristics of the irradiation proper in BR2 and for the determination of the corresponding operation data. At the end of the BR2 irradiation, the experimental results, such as burn-ups, neutron fluences, helium production in the fuel pin claddings, etc. are correlated by neutronic calculations in order to examine the consistency of the post-irradiation results and to validate the routine calculation procedure and cross-section data employed. A comparison is made in this paper between neutronic calculation results and some post-irradiation data for MOL 7D, a cadmium screened sodium cooled loop containing a nineteen fuel pin bundle

Soil heat flux calculation for sunlit and shaded surfaces under row crops: 1 - Model Development and sensitivity analysis

Science.gov (United States)

Soil heat flux at the surface (G0) is strongly influenced by whether the soil is shaded or sunlit, and therefore can have large spatial variability for incomplete vegetation cover, such as across the interrows of row crops. Most practical soil-plant-atmosphere energy balance models calculate G0 as a...

Le ministre du Commerce international du Canada rencontre des ...

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

17 juil. 2017 ... La promotion de l'entrepreneuriat, la façon dont le commerce peut profiter aux femmes et à leur famille, et la création d'emplois pour les plus vulnérables étaient au coeur de la discussion en table ronde du ministre du Commerce international du Canada, l'honorable François-Philippe Champagne, et des ...

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

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

Neutron point-flux calculation by Monte Carlo

International Nuclear Information System (INIS)

Eichhorn, M.

1986-04-01

A survey of the usual methods for estimating flux at a point is given. The associated variance-reducing techniques in direct Monte Carlo games are explained. The multigroup Monte Carlo codes MC for critical systems and PUNKT for point source-point detector-systems are represented, and problems in applying the codes to practical tasks are discussed. (author)

Picholine marocaine clones X Picholine du Languedoc

African Journals Online (AJOL)

SERVER

2007-12-17

Dec 17, 2007 ... marocaine" cultivar and the "Picholine du Languedoc" cultivar, the descendants and their parents were analysed using 35 ... marocaine" is the most dominant cultivar; more than 98% ... this proportion, a χ2 test was calculated to test the null hypothesis. ... a lethal or semi-lethal locus (Bradshaw and Stettler,.

PODESY program for flux mapping of CNA II reactor:

International Nuclear Information System (INIS)

Ribeiro Guevara, Sergio

1988-01-01

The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

A novel approach to calculate inductance and analyze magnetic flux density of helical toroidal coil applicable to Superconducting Magnetic Energy Storage systems (SMES)

International Nuclear Information System (INIS)

Alizadeh Pahlavani, M.R.; Shoulaie, A.

2010-01-01

In this paper, formulas are proposed for the self and mutual inductance calculations of the helical toroidal coil (HTC) by the direct and indirect methods at superconductivity conditions. The direct method is based on the Neumann's equation and the indirect approach is based on the toroidal and the poloidal components of the magnetic flux density. Numerical calculations show that the direct method is more accurate than the indirect approach at the expense of its longer computational time. Implementation of some engineering assumptions in the indirect method is shown to reduce the computational time without loss of accuracy. Comparison between the experimental measurements and simulated results for inductance, using the direct and the indirect methods indicates that the proposed formulas have high reliability. It is also shown that the self inductance and the mutual inductance could be calculated in the same way, provided that the radius of curvature is >0.4 of the minor radius, and that the definition of the geometric mean radius in the superconductivity conditions is used. Plotting contours for the magnetic flux density and the inductance show that the inductance formulas of helical toroidal coil could be used as the basis for coil optimal design. Optimization target functions such as maximization of the ratio of stored magnetic energy with respect to the volume of the toroid or the conductor's mass, the elimination or the balance of stress in some coordinate directions, and the attenuation of leakage flux could be considered. The finite element (FE) approach is employed to present an algorithm to study the three-dimensional leakage flux distribution pattern of the coil and to draw the magnetic flux density lines of the HTC. The presented algorithm, due to its simplicity in analysis and ease of implementation of the non-symmetrical and three-dimensional objects, is advantageous to the commercial software such as ANSYS, MAXWELL, and FLUX. Finally, using the

Comparaison du filtre adaptatif RIF et du filtre a base de reseau de ...

African Journals Online (AJOL)

Comparaison du filtre adaptatif RIF et du filtre a base de reseau de neurones pour le filtrage du courant de reference pour la commande du filtre actif parallele. C Benachaiba, A Bassou, B Mazari ...

Atmospheric neutrino fluxes

International Nuclear Information System (INIS)

Honda, M.; Kasahara, K.; Hidaka, K.; Midorikawa, S.

1990-02-01

A detailed Monte Carlo simulation of neutrino fluxes of atmospheric origin is made taking into account the muon polarization effect on neutrinos from muon decay. We calculate the fluxes with energies above 3 MeV for future experiments. There still remains a significant discrepancy between the calculated (ν e +antiν e )/(ν μ +antiν μ ) ratio and that observed by the Kamiokande group. However, the ratio evaluated at the Frejus site shows a good agreement with the data. (author)

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)

Limits of agricultural greenhouse gas calculators to predict soil N₂O and CH₄ fluxes in tropical agriculture

DEFF Research Database (Denmark)

Richards, Meryl; Metzel, Ruth; Chirinda, Ngonidzashe

2016-01-01

measurements from Africa, Asia, and Latin America. Estimates based on GHG calculators were greater than measurements in 70% of the cases, exceeding twice the measured flux nearly half the time. For 41% of the comparisons, calculators incorrectly predicted whether emissions would increase or decrease...

Developpement d'une methode de Monte Carlo dependante du temps et application au reacteur de type CANDU-6

Science.gov (United States)

Mahjoub, Mehdi

La resolution de l'equation de Boltzmann demeure une etape importante dans la prediction du comportement d'un reacteur nucleaire. Malheureusement, la resolution de cette equation presente toujours un defi pour une geometrie complexe (reacteur) tout comme pour une geometrie simple (cellule). Ainsi, pour predire le comportement d'un reacteur nucleaire,un schema de calcul a deux etapes est necessaire. La premiere etape consiste a obtenir les parametres nucleaires d'une cellule du reacteur apres une etape d'homogeneisation et condensation. La deuxieme etape consiste en un calcul de diffusion pour tout le reacteur en utilisant les resultats de la premiere etape tout en simplifiant la geometrie du reacteur a un ensemble de cellules homogenes le tout entoure de reflecteur. Lors des transitoires (accident), ces deux etapes sont insuffisantes pour pouvoir predire le comportement du reacteur. Comme la resolution de l'equation de Boltzmann dans sa forme dependante du temps presente toujours un defi de taille pour tous types de geometries,un autre schema de calcul est necessaire. Afin de contourner cette difficulte, l'hypothese adiabatique est utilisee. Elle se concretise en un schema de calcul a quatre etapes. La premiere et deuxieme etapes demeurent les memes pour des conditions nominales du reacteur. La troisieme etape se resume a obtenir les nouvelles proprietes nucleaires de la cellule a la suite de la perturbation pour les utiliser, au niveau de la quatrieme etape, dans un nouveau calcul de reacteur et obtenir l'effet de la perturbation sur le reacteur. Ce projet vise a verifier cette hypothese. Ainsi, un nouveau schema de calcul a ete defini. La premiere etape de ce projet a ete de creer un nouveau logiciel capable de resoudre l'equation de Boltzmann dependante du temps par la methode stochastique Monte Carlo dans le but d'obtenir des sections efficaces qui evoluent dans le temps. Ce code a ete utilise pour simuler un accident LOCA dans un reacteur nucleaire de type

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

Science.gov (United States)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Un aspect du calcul d'erreur sur les réserves en place d'un gisement : L'influence du nombre et de la disposition spatiale des puits One Aspect of Error Computing for Reserves in a Reservoir. Influence of Well Number and Spacing

Directory of Open Access Journals (Sweden)

Haas A.

2006-11-01

Full Text Available L'erreur sur l'évaluation des réserves en place d'un gisement d'hydrocarbures dépend de trois facteurs : - erreur aux puits sur la détermination des porosités et saturations; - erreur géostatistique d'extension au gisement des mesures effectuées aux puits ; - erreur géométrique d'évaluation de la surface ou du volume du gisement. Dans ce texte nous avons étudié l'influence du nombre et de la distribution des puits sur l'erreur géostatistique dans le cas d'un gisement fictif de forme elliptique. Nous nous sommes placés à différents niveaux de reconnaissance, depuis l'implantation d'un seul puits en position variable jusqu'à la couverture complète du gisement par une grille régulière de 48 puits... La méthode utilisée est le « krigeage » élaboré par G. MATHERON de l'École des Mines de Paris. Les calculs ont été réalisés à l'aide du programme KRIGEPACK développé par une association CFP-SNPA. L'erreur d'estimation dépend de la position des puits dans le gisement, de la plus ou moins grande continuité spatiale de la variable et des erreurs aux puits. L'erreur que l'on peut calculer par la statistique classique ne dépend que du nombre de puits et suivant le cas peut être trop importante (si les puits sont implantés de manière optimale ou au contraire trop faible (si les puits sont mal disposés. The error in evaluating reserves in place in a hydrocarbon reservoir depends on The following three factors I - an erorr in wells when determining porosities and saturations ; 2 - a geosfatistical error in extending well measurements ta the entire reservoir; 3 - a geometric error in evaluating the area or volume of the reservoir. This article studies the influence of well number and distribution on the geostatistical error in the case of an elliptically-shaped imaginary reservoir. Various levels of exploration are considered, from the existence of a single well in various positions to full coverage of the reservoir by a

Neutron flux distribution forecasting device of reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1991-01-01

A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

Specification of ROP flux shape

Energy Technology Data Exchange (ETDEWEB)

Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Gray, A [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1997-06-01

The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutron flux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutron flux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs.

Specification of ROP flux shape

International Nuclear Information System (INIS)

Min, Byung Joo; Gray, A.

1997-06-01

The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutron flux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutron flux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

1) Introduction: The difficulties of the formulation of the equations of phenomena occurring during the operation of a fusion reactor are underlined. 2) The possibilities presented by analog computation of the solution of nonlinear differential equations are enumerated. The accuracy and limitations of this method are discussed. 3) The analog solution in the stationary problem of the measurement of the discharge confinement is given and comparison with experimental results. 4) The analog solution of the dynamic problem of the evolution of the discharge current in a simple case is given and it is compared with experimental data. 5) The analog solution of the motion of an isolated ion in the electromagnetic field is given. A spatial field simulator used for this problem (bidimensional problem) is described. 6) The analog solution of the preceding problem for a tridimensional case for particular geometrical configurations using simultaneously 2 field simulators is given. 7) A method of computation derived from Monte Carlo method for the study of dynamic of plasma is described. 8) Conclusion: the essential differences between the analog computation of fission reactors and fusion reactors are analysed. In particular the theory of control of a fusion reactor as described by SCHULTZ is discussed and the results of linearized formulations are compared with those of nonlinear simulation. (author)Fren. [French] 1) Introduction. On souligne les difficultes que presente la mise en equation des phenomenes mis en jeu lors du fonctionnement d'un reacteur a fusion. On selectionne un certain nombre d'equations generalement utilisees et on montre les impossibilites analytiques auxquelles on se heurte alors. 2) On rappelle les possibilites du calcul analogique pour la resolution des systemes differentiels non lineaires et on indique la precision de la methode ainsi que ses limitations. 3) On decrit esolution analogique du probleme statique de la mesure du confinement de la decharge

Réflexion sur l’origine du processus de segmentation du marche du travail

Directory of Open Access Journals (Sweden)

Attia Nicole

2006-01-01

Full Text Available (francuski Ce travail propose une réflexion sur l'origine du processus de segmentation du marché du travail par rapport à l'entreprise. Se situe-t-elle au sein même de l'entreprise ou en amont, c'est à dire entre les entreprises? Cela revient à se demander si on peut avoir une approche microéconomique ou macroéconomique de la segmentation et, à s'interroger sur le rôle réel tenu par les firmes dans le processus. Déterminant pour la théorie, ce rôle est à repenser selon la réponse apportée à notre question.

SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations

Energy Technology Data Exchange (ETDEWEB)

Adams, C. H.

1976-07-01

This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.

SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations

International Nuclear Information System (INIS)

Adams, C.H.

1976-07-01

This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center

Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

International Nuclear Information System (INIS)

Bobkov, V.P.

2000-01-01

The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

Empirical Calibration for Dolomite Stoichiometry Calculation: Application on Triassic Muschelkalk- Lettenkohle Carbonates (French Jura Calibration empirique pour le calcul de la stoechiométrie de la dolomite : application aux carbonates triasiques du Muschelkalk-Lettenkohle (Jura français

Directory of Open Access Journals (Sweden)

Turpin M.

2012-02-01

��équation de Lumsden et de méthodes antérieures. L’approche proposée est vérifiée grâce à l’analyse d’un échantillon référence de dolomite (Eugui et appliquée à des roches carbonatées du Trias (Muschelkalk supérieur-Lettenkohle du Jura français. Elle est combinée à une étude pétrographique et isotopique et peut être appliquée tant aux roches qu’aux ciments dolomitiques. Les dolomies du Muschelkalk supérieur se sont formées au cours d’une dolomitisation d’enfouissement associée à des fluides dont la température augmente et ayant une composition isotopique variable au cours de l’enfouissement. Outre la pétrographie, ceci est également mis en évidence par le pourcentage de Ca calculé dans les dolomites, qui atteint progressivement une stoechiométrie idéale (de 53,16 % à 51,19 % parallèlement au développement de la dolomitisation. Les dolomites du Lettenkohle montrent une stoechiométrie proche d’une stoechiométrie idéale (51,06 % Ca, sont ordonnées et associées à de l’anhydrite. Leur formation est liée à deux étapes de dolomitisation : de type sabkha et d’enfouissement. Cette étude permet ainsi une meilleure caractérisation des différents types de dolomites dans les roches sédimentaires et par conséquent une meilleure détermination de leur potentiel en tant que réservoir.

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

Science.gov (United States)

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Original Paper Performances comparées du HDL-cholestérol et du ...

African Journals Online (AJOL)

CT/HDL-C) et du HDL-Cholestérol est le meilleur prédicteur du SMet chez les adultes béninois. .... (Canada) et du. Ministère de la Santé du Bénin. Le consentement éclairé écrit a été obtenu de chaque participant avant leur recrutement dans.

Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

International Nuclear Information System (INIS)

Abit, K.

1984-01-01

The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

Régionalisation du recrutement du personnel de santé au Burkina ...

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

Régionalisation du recrutement du personnel de santé au Burkina Faso ... le ministère a adopté une politique de recrutement régionalisé de certaines catégories du ... de comprendre le contexte dans lequel une telle stratégie a été formulée, ...

Limits of agricultural greenhouse gas calculators to predict soil N2O and CH4 fluxes in tropical agriculture

Science.gov (United States)

Richards, Meryl; Metzel, Ruth; Chirinda, Ngonidzashe; Ly, Proyuth; Nyamadzawo, George; Duong Vu, Quynh; de Neergaard, Andreas; Oelofse, Myles; Wollenberg, Eva; Keller, Emma; Malin, Daniella; Olesen, Jørgen E.; Hillier, Jonathan; Rosenstock, Todd S.

2016-05-01

Demand for tools to rapidly assess greenhouse gas impacts from policy and technological change in the agricultural sector has catalyzed the development of ‘GHG calculators’— simple accounting approaches that use a mix of emission factors and empirical models to calculate GHG emissions with minimal input data. GHG calculators, however, rely on models calibrated from measurements conducted overwhelmingly under temperate, developed country conditions. Here we show that GHG calculators may poorly estimate emissions in tropical developing countries by comparing calculator predictions against measurements from Africa, Asia, and Latin America. Estimates based on GHG calculators were greater than measurements in 70% of the cases, exceeding twice the measured flux nearly half the time. For 41% of the comparisons, calculators incorrectly predicted whether emissions would increase or decrease with a change in management. These results raise concerns about applying GHG calculators to tropical farming systems and emphasize the need to broaden the scope of the underlying data.

35 Typologie des eaux de surface du bassin du Sebou par multi ...

African Journals Online (AJOL)

PR BOKO

2Service de protection de la qualité de l'eau, Agence du Bassin Hydraulique du ... pour montrer l'évolution de la qualité biologique des eaux de surface du ..... Biological Indicators of Freshwater Pollution and Environmental Management,.

Pollutant transport over complex terrain: Flux and budget calculations for the pollumet field campaign

Science.gov (United States)

Lehning, Michael; Richner, Hans; Kok, Gregory L.

Especially over complex terrain, transport processes dominate the local pollutant concentrations observed. The data gathered during the POLLUMET measuring campaign in 1993 allow a quantitative analysis of the pollutant fluxes and the pollutant budgets. The data include airborne measurements by NCAR's King Air, radio soundings, radar wind profiles, and data from meteorological ground stations. The regions of interest were the rather densely populated Swiss Plateau, which is embedded between the Alps and the Jura Mountains, and a box south of the Alps covering the south Ticino region and parts of northern Italy. An interpolation scheme was developed to reconstruct the wind field from all available measurements. From the wind field and the reconstruction of the concentration field the fluxes into and out of a box with fixed boundaries are calculated. The pollutant budgets are obtained from the sum of the fluxes and considering a mean vertical velocity. To assess the uncertainties introduced through the interpolation of the measurements, an extensive sensitivity analysis is included. The Swiss Plateau exports ozone and nitrogen oxides. The export rates can be interpreted as an ozone accumulation or fraction of 'homemade pollution' between 3 and 10% and require a net production rate of 1-2 ppb h -1. Accumulation of nitrogen oxides amounts to 20-60%. The box south of the Alps imports polluted air from northern Italy. Thus, oxidized nitrogen is not exported but a net production of ozone still occurs at a rate of 1-2 ppb h -1. The interpolated flow and concentration fields are decomposed into the mean over a box-boundary and the deviation from that mean. This allows isolation of the contribution of local circulations and large-scale turbulence to the total flux. It is shown how the local thermotopographic circulations increasingly dominate the transport as typical Alpine topography is approached. Even over the Swiss Plateau, approximately 20 km away from Alpine topography

Caractérisation des sables et morphologie du fond du lac du ...

African Journals Online (AJOL)

Une analyse sédimentologique et minéralogique réalisée sur un cycle hydrologique entre octobre 2004 et août 2005 a permis d\\'évaluer les charges solides en suspension et de caractériser les sédiments du lac du barrage de Taabo. La concentration moyenne en matières en suspension (12 mg.L-1) et la turbidité ...

Caractérisation des sables et morphologie du fond du lac du ...

African Journals Online (AJOL)

Administrateur

Une analyse sédimentologique et minéralogique réalisée sur un cycle hydrologique entre octobre 2004 et août 2005 a permis d'évaluer les charges solides en suspension et de caractériser les sédiments du lac du barrage de Taabo. La concentration moyenne en matières en suspension (12 mg.L-1) et la turbidité ...

Engagez-vous, devenez délégué(e) du personnel du CERN

CERN Multimedia

Staff Association

2017-01-01

Dans notre ECHO N° 275, nous avons annoncé les élections à venir au Conseil du personnel du CERN. Dans le présent ECHO, nous vous informons du lancement du processus des élections qui débute par le dépôt des candidatures. Tous les titulaires, boursiers et associés, qui sont aussi membres de l’Association du personnel, peuvent s’engager et déposer leur candidature entre le 11 septembre à 08 h 00 et le 13 octobre 2017 à 17 h 00. N’hésitez plus, remplissez le formulaire de candidature, présentez-vous aux élections au Conseil du personnel afin de pouvoir représenter et défendre vos collègues du personnel du CERN. ÊTRE DÉLÉGUÉ(E), C’EST QUOI ? Poser la question à plusieurs d...

APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN

International Nuclear Information System (INIS)

Kawasaki, Hiromitsu; Seki, Yasushi

1983-01-01

A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron

Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method

International Nuclear Information System (INIS)

Shafii, M. Ali; Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny; Ariani, Menik; Yulianti, Yanti

2010-01-01

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k eff and other parameters are investigated.

D-[U-11C]glucose uptake and metabolism in the brain of insulin-dependent diabetic subjects

International Nuclear Information System (INIS)

Gutniak, M.; Blomqvist, G.; Widen, L.; Stone-Elander, S.; Hamberger, B.; Grill, V.

1990-01-01

We used D-[U-11C]glucose to evaluate transport and metabolism of glucose in the brain in eight nondiabetic and six insulin-dependent diabetes mellitus (IDDM) subjects. IDDM subjects were treated by continuous subcutaneous insulin infusion. Blood glucose was regulated by a Biostator-controlled glucose infusion during a constant insulin infusion. D-[U-11C]-glucose was injected for positron emission tomography studies during normoglycemia as well as during moderate hypoglycemia [arterial plasma glucose 2.74 +/- 0.14 in nondiabetic and 2.80 +/- 0.26 mmol/l (means +/- SE) in IDDM subjects]. Levels of free insulin were constant and similar in both groups. The tracer data were analyzed using a three-compartment model with a fixed correction for 11CO2 egression. During normoglycemia the influx rate constant (k1) and blood-brain glucose flux did not differ between the two groups. During hypoglycemia k1 increased significantly and similarly in both groups (from 0.061 +/- 0.007 to 0.090 +/- 0.006 in nondiabetic and from 0.061 +/- 0.006 to 0.093 +/- 0.013 ml.g-1.min-1 in IDDM subjects). During normoglycemia the tracer-calculated metabolism of glucose was higher in the whole brain in the nondiabetic than in the diabetic subjects (22.0 +/- 1.9 vs. 15.6 +/- 1.1 mumol.100 g-1.min-1, P less than 0.01). During hypoglycemia tracer-calculated metabolism was decreased by 40% in nondiabetic subjects and by 28% in diabetic subjects. The results indicate that uptake of glucose is normal, but some aspect of glucose metabolism is abnormal in a group of well-controlled IDDM subjects

Ecologie du phytoplancton du lac Kivu

Directory of Open Access Journals (Sweden)

Sarmento, H.

2008-01-01

Full Text Available Speciation within the African Coffee Pathogen. Cet article analyse s'il est avantageux d'utiliser le compost au lieu de l'engrais minéral pour produire la laitue dans la zone urbaine et péri-urbaine de Yaoundé. Les résultats de terrain montrent l'obtention de rendements et profits plus élevés lorsqu'on utilise le compost. Les résultats de la fonction de production Cobb-Douglas prouvent que l'utilisation du compost est statistiquement significative pour expliquer la variation de rendement de la laitue et que le compost est l'intrant le plus productif. D'autres résultats montrent que le compost fournit la matière organique utile au sol et que les besoins d'irrigation en eau de la culture sont réduits grâce à l'utilisation du compost. Par conséquent, malgré le fait que l'application du compost demande une main-d'oeuvre beaucoup plus élevée, son utilisation est généralement bénéfique pour les agriculteurs vivant aux alentours de Yaoundé. Les programmes de vulgarisation de cet intrant pour encourager son adoption devraient donc figurer parmi les points prioritaires dans la politique agricole du gouvernement camerounais.

Analytical Calculation of the Magnetic Field distribution in a Flux-Modulated Permanent-Magnet Brushless Motor

DEFF Research Database (Denmark)

Zhang, Xiaoxu; Liu, Xiao; Chen, Zhe

2015-01-01

This paper presents a rapid approach to compute the magnetic field distribution in a flux-modulated permanent-magnet brushless motor. Partial differential equations are used to describe the magnet field behavior in terms of magnetic vector potentials. The whole computational domain is divided...... into several regions, i.e., magnet, air-gaps, slot-openings, and slots. The numerical solution could be obtained by applying the boundary constraints on the interfaces between these regions. The accuracy of the proposed analytical model is verified by comparing the no-load magnetic field and armature reaction...... magnetic field with those calculated by finite element method....

TRANSHEX, 2-D Thermal Neutron Flux Distribution from Epithermal Flux in Hexagonal Geometry

International Nuclear Information System (INIS)

Patrakka, E.

1994-01-01

1 - Description of program or function: TRANSHEX is a multigroup integral transport program that determines the thermal scalar flux distribution arising from a known epithermal flux in two- dimensional hexagonal geometry. 2 - Method of solution: The program solves the isotropic collision probability equations for a region-averaged scalar flux by an iterative method. Either a successive over-relaxation or an inner-outer iteration technique is applied. Flat flux collision probabilities between trigonal space regions with white boundary condition are utilized. The effect of epithermal flux is taken into consideration as a slowing-down source that is calculated for a given spatial distribution and 1/E energy dependence of the epithermal flux

Neutron flux calculations for criticality safety analysis using the narrow resonance approximations. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.

Fiche technique du spermogramme et du spermocytogramme ...

African Journals Online (AJOL)

En Afrique la stérilité du couple constitue un drame social. Selon l'OMS, environ 8 à 12 % des couples africains sont touchés par une infertilité. La responsabilité masculine dans la stérilité est comprise entre 30 à 40%. Les causes de l'infertilité masculine peuvent être l'impuissance et/ ou l'altération du sperme. L'étude de ...

Experimental apparatus for in-pile studies of: creep of nuclear fuels, and Young's-modulus of structural materials; Dispositifs experimentaux pour etudes en pile du fluage des materiaux combustibles, et du module d'elasticite des materiaux de structure

Energy Technology Data Exchange (ETDEWEB)

Gautier, A; Le Bret, P; Alfille, L; Pesenti, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

Creep test under compression: the apparatus described allows to study, in an horizontal beam hole of a research reactor such as EL2, the creep behaviour of nuclear fuel samples under neutron flux. The maximum stress applied on the specimens is a constant compression chooses between 0.200 and 0.400 kg/mm{sup 2} (285 psi and 570 psi). - Young's Modulus measurement: in another horizontal beam hole of such a reactor, an apparatus allows to study the irradiation effect on Young's Modulus of a structural material specimen. (author)Fren. [French] Essai de fluage en compression: l'appareillage decrit permet d'etudier, dans un canal horizontal d'une pile experimentale type EL2, le fluage en compression d'eprouvettes de materiaux fissiles sous le flux de neutrons, sous une contrainte maximum de 500 g/mm{sup 2}. - Mesure du module d'Young: dans un canal identique au precedent, un appareillage permet de suivre l'influence du rayonnement sur le module d'elasticite d'une eprouvette d'un materiau de structure. (auteur)

Calculation of the thermal neutron flux depression in the loop VISA-1; Izracunavanje depresije fluksa termalnih neutrona u 'petlji' VISA-1

Energy Technology Data Exchange (ETDEWEB)

Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Among other applications, the VISA-1 loop is to be used for thermal load testing of materials. For this type of testing one should know the maximum power generated in the loop. This power is determined from the maximum thermal neutron flux in the VK-5 channel and mean flux depression in the fissile component of the loop. Thermal neutron flux depression is caused by neutron absorption in the components of the loop, shape of the components and neutron leaking through gaps as well as properties of the surrounding medium of the core. All these parameters were taken into account for calculating the depression of thermal neutron flux in the VISA-1 loop. Two group diffusion theory was used. Fast neutron from the fission in the loop and slowed down were taken into account. Depression of the thermal neutron flux is expressed by depression factor which represents the ratio of the mean thermal neutron flux in the fissile loop component and the thermal neutron flux in the VK-5 without the loop. Calculation error was estimated and it was recommended to determine the depression factor experimentally as well. [Serbo-Croat] Petlja VISA-1 namenjena je izmedju ostalog ispitivanju materiajala na termicka naprezanja. Za ova ispitivanja potrebno je poznavati maksimalnu snagu koja se razvija u petlji, a ona se odredjuje na osnovu maksimalnog fluksa termalnih neutrona u kanalu VK-5 i srednje depresije fluksa u fisibilnoj komponenti petlje. Depresija fluksa termalnih neutrona uzrokovana je apsorpcijom neutrona u komponentama petlje, geometrijom komponeni i isticanjem neutrona preko supljina u petlji kao i osobinama reaktorske sredine koja okruzuje petlju. Svi ovi faktori uzeti su u obzir pri proracunu depresije fluksa termalnih neutrona u petlji VISA-1. Primenjena je difuziona dvo grupna teorija. Uzeti su u obzir brzi neutroni nastali fisijom u petlji i usporeni u aktivnoj zoni RA. Depresija neutronskog fluksa izrazena je depresionim faktorom, koji predstavlja odnos srednjeg fluksa

Burnout calculation

International Nuclear Information System (INIS)

Li, D.

1980-01-01

Reviewed is the effect of heat flux of different system parameters on critical density in order to give an initial view on the value of several parameters. A thorough analysis of different equations is carried out to calculate burnout is steam-water flows in uniformly heated tubes, annular, and rectangular channels and rod bundles. Effect of heat flux density distribution and flux twisting on burnout and storage determination according to burnout are commended [ru

Study of {gamma} radiation from uranium rods during deactivation; Etude du rayonnement {gamma} des barres d'uranium en court de desactivation

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

The classical formulae giving the {gamma} activities of the fission products contained in a uranium rod after unloading from the pile are reviewed without being proved. The knowledge of these activities makes it possible, by means of the method proposed here, to determine the intensities of ionisation at a point outside the rod, and thus to establish {gamma} radiation diagrams. The different parameters introduced in the calculation are geometric (dimensions of the bars and coordinates of the point considered), energetic (power at which the bar has been irradiated) and temporal (duration of the irradiation and deactivation). A numerical example follows the demonstration of the general formulae, {gamma} flux measurements carried out in the deactivation well of P2 (Saclay pile) define the accuracy of the method. In conclusion, it is suggested that radiation diagrams be used in (planning the use of) industrial irradiators for radiochemical polymerisation or the preservation of food products. (author) [French] On rappelle sans demonstration les formules classiques donnant les activites {gamma} des produits de fission contenus dans une barre d'uranium apres defournement. La connaissance de ces activites permet par la methode proposee de passer aux intensites d'ionisation en un point exterieur a la barre et d'etablir ainsi des diagrammes de rayonnement {gamma}. Les differents parametres introduits dans le calcul sont d'ordre geometrique (dimensions des barres et coordonnees du point considere), d'ordre energetique (puissance a laquelle la barre a ete irradiee) et fonction du temps (duree d'irradiation et de desactivation). Un exemple numerique fait suite a la demonstration des formules generales. Des mesures de flux {gamma} effectuees au puits de desactivation de P2 (pile de Saclay) fixent le degre d'approximation de la methode. En conclusion, on suggere l'utilisation des diagrammes de rayonnement dans l'etablissement de projets d'irradiateurs industriels pour les

Calculating the dermal flux of chemicals with OELs based on their molecular structure: An attempt to assign the skin notation.

Science.gov (United States)

Kupczewska-Dobecka, Małgorzata; Jakubowski, Marek; Czerczak, Sławomir

2010-09-01

Our objectives included calculating the permeability coefficient and dermal penetration rates (flux value) for 112 chemicals with occupational exposure limits (OELs) according to the LFER (linear free-energy relationship) model developed using published methods. We also attempted to assign skin notations based on each chemical's molecular structure. There are many studies available where formulae for coefficients of permeability from saturated aqueous solutions (K(p)) have been related to physicochemical characteristics of chemicals. The LFER model is based on the solvation equation, which contains five main descriptors predicted from chemical structure: solute excess molar refractivity, dipolarity/polarisability, summation hydrogen bond acidity and basicity, and the McGowan characteristic volume. Descriptor values, available for about 5000 compounds in the Pharma Algorithms Database were used to calculate permeability coefficients. Dermal penetration rate was estimated as a ratio of permeability coefficient and concentration of chemical in saturated aqueous solution. Finally, estimated dermal penetration rates were used to assign the skin notation to chemicals. Defined critical fluxes defined from the literature were recommended as reference values for skin notation. The application of Abraham descriptors predicted from chemical structure and LFER analysis in calculation of permeability coefficients and flux values for chemicals with OELs was successful. Comparison of calculated K(p) values with data obtained earlier from other models showed that LFER predictions were comparable to those obtained by some previously published models, but the differences were much more significant for others. It seems reasonable to conclude that skin should not be characterised as a simple lipophilic barrier alone. Both lipophilic and polar pathways of permeation exist across the stratum corneum. It is feasible to predict skin notation on the basis of the LFER and other published

Measurement and simulation of thermal neutron flux distribution in the RTP core

Science.gov (United States)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Analyse des systèmes de gestion des déchets par l’étude du mécanisme de répartition des déchets

OpenAIRE

Bergeron, Francis

2015-01-01

Le renforcement de l’usage concurrentiel des déchets nous amène à étudier les processus d’approvisionnement des filières de valorisation et d’élimination sous l’angle de la circulation des flux de déchets au sein d’un système de gestion des déchets. L’objectif du présent article de recherche est de concevoir un cadre théorique autour du concept de mécanisme de répartition des déchets qui désigne le processus de distribution des flux de déchets entre les filières d’élimination et de valorisati...

Calculating flux to predict future cave radon concentrations

Czech Academy of Sciences Publication Activity Database

Rowberry, Matthew David; Martí, Xavier; Frontera, C.; Van De Wiel, M.J.; Briestenský, Miloš

2016-01-01

Roč. 157, JUN (2016), 16-26 ISSN 0265-931X R&D Projects: GA MŠk LM2010008 Institutional support: RVO:67985891 ; RVO:68378271 Keywords : cave radon concentration * cave radon flux * cave ventilation * radioactive decay * fault slip * numerical modelling Subject RIV: DC - Siesmology, Volcanology, Earth Structure; BG - Nuclear, Atomic and Molecular Physics, Colliders (FZU-D) Impact factor: 2.310, year: 2016

Uncertainty calculations made easier

International Nuclear Information System (INIS)

Hogenbirk, A.

1994-07-01

The results are presented of a neutron cross section sensitivity/uncertainty analysis performed in a complicated 2D model of the NET shielding blanket design inside the ITER torus design, surrounded by the cryostat/biological shield as planned for ITER. The calculations were performed with a code system developed at ECN Petten, with which sensitivity/uncertainty calculations become relatively simple. In order to check the deterministic neutron transport calculations (performed with DORT), calculations were also performed with the Monte Carlo code MCNP. Care was taken to model the 2.0 cm wide gaps between two blanket segments, as the neutron flux behind the vacuum vessel is largely determined by neutrons streaming through these gaps. The resulting neutron flux spectra are in excellent agreement up to the end of the cryostat. It is noted, that at this position the attenuation of the neutron flux is about 1 l orders of magnitude. The uncertainty in the energy integrated flux at the beginning of the vacuum vessel and at the beginning of the cryostat was determined in the calculations. The uncertainty appears to be strongly dependent on the exact geometry: if the gaps are filled with stainless steel, the neutron spectrum changes strongly, which results in an uncertainty of 70% in the energy integrated flux at the beginning of the cryostat in the no-gap-geometry, compared to an uncertainty of only 5% in the gap-geometry. Therefore, it is essential to take into account the exact geometry in sensitivity/uncertainty calculations. Furthermore, this study shows that an improvement of the covariance data is urgently needed in order to obtain reliable estimates of the uncertainties in response parameters in neutron transport calculations. (orig./GL)

Effect of chamber enclosure time on soil respiration flux: A comparison of linear and non-linear flux calculation methods

DEFF Research Database (Denmark)

Kandel, Tanka P; Lærke, Poul Erik; Elsgaard, Lars

2016-01-01

One of the shortcomings of closed chamber methods for soil respiration (SR) measurements is the decreased CO2 diffusion rate from soil to chamber headspace that may occur due to increased chamber CO2 concentrations. This feedback on diffusion rate may lead to underestimation of pre-deployment flu......One of the shortcomings of closed chamber methods for soil respiration (SR) measurements is the decreased CO2 diffusion rate from soil to chamber headspace that may occur due to increased chamber CO2 concentrations. This feedback on diffusion rate may lead to underestimation of pre...... was placed on fixed collars, and CO2 concentration in the chamber headspace were recorded at 1-s intervals for 45 min. Fluxes were measured in different soil types (sandy, sandy loam and organic soils), and for various manipulations (tillage, rain and drought) and soil conditions (temperature and moisture......) to obtain a range of fluxes with different shapes of flux curves. The linear method provided more stable flux results during short enclosure times (few min) but underestimated initial fluxes by 15–300% after 45 min deployment time. Non-linear models reduced the underestimation as average underestimation...

On the use of flux limiters in the discrete ordinates method for 3D radiation calculations in absorbing and scattering media

International Nuclear Information System (INIS)

Godoy, William F.; DesJardin, Paul E.

2010-01-01

The application of flux limiters to the discrete ordinates method (DOM), S N , for radiative transfer calculations is discussed and analyzed for 3D enclosures for cases in which the intensities are strongly coupled to each other such as: radiative equilibrium and scattering media. A Newton-Krylov iterative method (GMRES) solves the final systems of linear equations along with a domain decomposition strategy for parallel computation using message passing libraries in a distributed memory system. Ray effects due to angular discretization and errors due to domain decomposition are minimized until small variations are introduced by these effects in order to focus on the influence of flux limiters on errors due to spatial discretization, known as numerical diffusion, smearing or false scattering. Results are presented for the DOM-integrated quantities such as heat flux, irradiation and emission. A variety of flux limiters are compared to 'exact' solutions available in the literature, such as the integral solution of the RTE for pure absorbing-emitting media and isotropic scattering cases and a Monte Carlo solution for a forward scattering case. Additionally, a non-homogeneous 3D enclosure is included to extend the use of flux limiters to more practical cases. The overall balance of convergence, accuracy, speed and stability using flux limiters is shown to be superior compared to step schemes for any test case.

Sensitivity of upper atmospheric emissions calculations to solar/stellar UV flux

Directory of Open Access Journals (Sweden)

Barthelemy Mathieu

2014-01-01

Full Text Available The solar UV (UltraViolet flux, especially the EUV (Extreme UltraViolet and FUV (Far UltraViolet components, is one of the main energetic inputs for planetary upper atmospheres. It drives various processes such as ionization, or dissociation which give rise to upper atmospheric emissions, especially in the UV and visible. These emissions are one of the main ways to investigate the upper atmospheres of planets. However, the uncertainties in the flux measurement or modeling can lead to biased estimates of fundamental atmospheric parameters, such as concentrations or temperatures in the atmospheres. We explore the various problems that can be identified regarding the uncertainties in solar/stellar UV flux by considering three examples. The worst case appears when the solar reflection component is dominant in the recorded spectrum as is seen for outer solar system measurements from HST (Hubble Space Telescope. We also show that the estimation of some particular line parameters (intensity and shape, especially Lyman α, is crucial, and that both total intensity and line profile are useful. In the case of exoplanets, the problem is quite critical since the UV flux of their parent stars is often very poorly known.

La population du Moyen-Orient et de l'Afrique du Nord contribue ...

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

29 avr. 2016 ... Le degré d'apport au contenu varie grandement d'une population à l'autre. Si les habitants du Moyen-Orient et de l'Afrique du Nord (région MENA) utilisent Wikipédia, ils y contribuent cependant moins que les populations d'autres régions du monde. Le contenu au sujet de la région MENA est, le plus ...

Albedo analytical method for multi-scattered neutron flux calculation in cavity

International Nuclear Information System (INIS)

Shin, Kazuo; Selvi, S.; Hyodo, Tomonori

1986-01-01

A simple formula which describes multi-scattered neutron flux in a spherical cavity was derived based on the albedo concept. The formura treats a neutron source which has an arbitrary energy-angle distribution and is placed at any point in the cavity. The derived formula was applied to the estimation of neutron fluxes in two cavities, i.e. a spherical concrete cell with a 14-MeV neutron source at the center and the ''YAYOI'' reactor cavity with a pencil beam of reactor neutrons. The results of the analytical formula agreed very well with the reference data in the both problems. It was concluded that the formula is applicable to estimate the neutron fluxes in a spherical cell except for special cases that tangential source neutrons are incident to the cavity wall. (author)

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

Science.gov (United States)

Töre, Candan; Ortego, Pedro

2005-01-01

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

Monte Carlo surface flux tallies

International Nuclear Information System (INIS)

Favorite, Jeffrey A.

2010-01-01

Particle fluxes on surfaces are difficult to calculate with Monte Carlo codes because the score requires a division by the surface-crossing angle cosine, and grazing angles lead to inaccuracies. We revisit the standard practice of dividing by half of a cosine 'cutoff' for particles whose surface-crossing cosines are below the cutoff. The theory behind this approximation is sound, but the application of the theory to all possible situations does not account for two implicit assumptions: (1) the grazing band must be symmetric about 0, and (2) a single linear expansion for the angular flux must be applied in the entire grazing band. These assumptions are violated in common circumstances; for example, for separate in-going and out-going flux tallies on internal surfaces, and for out-going flux tallies on external surfaces. In some situations, dividing by two-thirds of the cosine cutoff is more appropriate. If users were able to control both the cosine cutoff and the substitute value, they could use these parameters to make accurate surface flux tallies. The procedure is demonstrated in a test problem in which Monte Carlo surface fluxes in cosine bins are converted to angular fluxes and compared with the results of a discrete ordinates calculation.

User's guide for SLWDN9, a code for calculating flux-surfaced-averaging of alpha densities, currents, and heating in non-circular tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Miley, G.M.

1980-03-01

The code calculates flux-surfaced-averaged values of alpha density, current, and electron/ion heating profiles in realistic, non-circular tokamak plasmas. The code is written in FORTRAN and execute on the CRAY-1 machine at the Magnetic Fusion Energy Computer Center

Use of the Le Chatelier principle in calculating the uniform flux in channels formed by packed rods

International Nuclear Information System (INIS)

Skrebkov, G.P.; Lozhkin, S.N.

1986-01-01

A method is proposed for calculating the hydrodynamics of a uniform flow in channels with a cross section of complex form. The method takes into account the anisotropy of the momentum transfer. The anisotropy coefficient of the momentum transfer is determined by using the Le Chatelier principle in a virtual process of transition to the kinematic structure of a uniform flux in equilibrium with a specified set of external conditions which include the channel geometry, wall roughness, and the value of the piezometric gradient

Néotoponymie contestée à Potchefstroom / Tlokwe (Province du Nord-Ouest, Afrique du Sud

Directory of Open Access Journals (Sweden)

Myriam Houssay-Holzschuch

2010-09-01

Full Text Available Illustration 1 - Extrait du site de la municipalité de Potchefstroom / TlokweSource : http://www.potch.co.za/newsarchive/streetnames.htmlIllustration 2- Potchefstroom (Afrique du SudAuteur : Béatrice Obry-Guyot, début décembre 2007.Illustration 3 - Extrait du plan Google Map de PotchefstroomSource : Google Map.La question toponymique en Afrique du Sud sur la longue durée porte sur deux thèmes essentiels porteurs de controverses : le plurilinguisme et le marquage symbolique et mémoriel du te...

Ksenia Pimenova, Les sources de savoirs. Le renouveau du bouddhisme et du chamanisme chez les Touvas de la Sibérie du Sud

OpenAIRE

Pimenova, Ksenia

2013-01-01

Cette thèse présente une étude comparative du « renouveau » post-soviétique du chamanisme et du bouddhisme tibétain (école Guélougpa) chez les Touvas, un des peuples autochtones de la Sibérie du Sud (Russie). Ce phénomène a lieu après des décennies de politique antireligieuse (dès la fin des années 1920 jusqu’aux années 1980) ayant abouti à la destruction de la communauté bouddhique et à la marginalisation du chamanisme. Nous analysons le chamanisme et le bouddhisme post-soviétiques comme deu...

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

Energy Technology Data Exchange (ETDEWEB)

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

1952-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1952-07-01

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

Une méthode pour estimer l’interception du rayonnement par un couvert bas : application au colza avant montaison

Directory of Open Access Journals (Sweden)

Denoroy Pascal

2002-01-01

introduire dans l’équation 2 un paramètre représentant le taux de couverture du sol. Si l’estimation par mesure directe de epsiloni et donc l’estimation de k sont assez simples, au moins dans le principe pour un couvert développé en hauteur (colza après montaison [3], une telle mesure devient beaucoup plus complexe dans le cas d’un couvert discontinu et bas. Dans ce milieu hétérogène (et souvent dans des conditions hivernales sévères, une distribution représentative des capteurs de rayonnement est difficile, les risques de perturbation de la mesure (salissure des capteurs, déréglages sont élevés. Plusieurs études ont déjà traité de epsiloni et k dans le cas du colza [3-7]. Elles sont fondées sur des mesures d’interception du rayonnement et concernent des couverts fermés, après montaison. Une autre approche consiste à effectuer un calcul précis de PARi avec des modèles détaillés utilisant une description complète des paramètres (géométriques et radiatifs du feuillage, et du rayonnement incident [8]. Mais les données nécessaires sont alors nombreuses, complexes à obtenir et donc généralement non disponibles en expérimentation. Pour répondre aux besoins de calcul de l’interception du rayonnement dans les phases d’installation du colza, cet article présente une méthode permettant de calculer epsiloni en tenant compte des principales caractéristiques de la morphologie du couvert et du régime radiatif auquel ce dernier est soumis. Les données nécessaires sont limitées : l’indice foliaire, les taux de couverture du sol (deux définitions, la latitude du lieu considéré, la date et le rayonnement incident quotidien. Ce calcul permet une estimation du coefficient moyen d’extinction k (équation 2. On verra que k peut être corrélé à l’IF, la période de l’année et la latitude du lieu, ce qui permet de proposer une estimation simple de k et donc un calcul simplifié de epsiloni, adapté au cas des faibles IF

Spectrographic measurement of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

Energy Technology Data Exchange (ETDEWEB)

Artaud, J; Cittanova, J [Commissariat a l' Energie Atomique, Service d' Analyses et Recherches Chimiques Appliquees, Saclay (France). Centre d' Etudes Nucleaires; Crehange, G; Frequelin, S [Commissariat a l' Energie Atomique, Dir. des Applications Militaires, Service Chimie, Saclay (France). Centre d' Etudes Nucleaires; Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1961-07-01

We describe here a method for the spectrographic determination of beryllium on filters which is valid for amounts varying between 0,01 and 30 {mu}g of beryllium and which is independent of the nature of the beryllium compound involved. This is a flux method (graphite-lithium carbonate mixture), the excitation being by a direct current arc. (author) [French] Nous decrivons ici, une methode de dosage spectrographique de beryllium sur filtre, valable pour des teneurs comprises entre 0,01 et 30 {mu}g de beryllium et independante de la nature du compose de beryllium a doser. C'est une methode de 'flux' (melange graphite-carbonate de lithium) l'excitation etant un arc a courant continu. (auteur)

Tests on dynamic corrosion by water. Influence of the passage of a heat flux on the corrosion kinetics. pH measurement in water at high temperature; Essais de corrosion dynamique par l'eau. Influence du passage d'un flux thermique sur la cinetique de corrosion. Mesure du pH dans l'eau a haute temperature

Energy Technology Data Exchange (ETDEWEB)

Coriou, H; Grall, L; Hure, J; Saint-James, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod, [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France); peintre, Le [Centre National de la Recherche Scientifique (CNRS), 38 - Grenoble (France)

1958-07-01

The passage of a heat flux through the surface of a metal placed in a corrosive medium influences the rate of corrosion, these being higher than under adiabatic conditions. The apparatus developed for corrosion tests is described, it is possible to obtain with this equipment: 1) Heat fluxes greater than 200 W/cm{sup 2}, across aluminium canning, which is cooled by water (temperature 50 deg. C), circulating with flow rates of the order of 5 to 6 m/s. 2) Heat fluxes which can go up to 150 W/cm{sup 2}, across canning of zircaloy or stainless steel. The cooling fluid is pressurized water at a temperature around 280 deg. C, the flow-rate of circulation reaching 6 m/s. The results obtained on aluminium canning are studied from the viewpoint of corrosion, paying particular attention to cavitation phenomena which can cause serious damage in certain special circumstances. After developing a glass electrode system capable of supporting high pressures, the authors have investigated materials capable of functioning as a hydrogen electrode and of resisting satisfactorily corrosion by water at 200 deg. C. Various possibilities have been examined: electrodes of special glasses, quartz, metals, with a membrane etc... The results of the various tests and the practical limits of utilisation are given. (author)Fren. [French] Le passage d'un flux thermique a travers la surface d'un metal place dans un milieu corrosif influence les vitesses de corrosion, celles-ci etant plus elevees que dans des conditions adiabatiques. On decrit les appareils mis au point, pour essais de corrosion. Ils permettent d'obtenir: 1) A travers des gaine aluminium des flux thermiques depassant 200 W /cm{sup 2}. Les gaines sont refroidies par l'eau (temperature 50 deg. C), circulant a des vitesses de l'ordre de 5 a 6 m/s. 2) A travers des gaines en zircaloy ou acier inoxydable des flux thermiques pouvant s'elever a 150 W/cm{sup 2}. Le fluide de refroidissement est de l'eau sous pression a une temperature

Bulletin du CRDI #124

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

Les femmes jouent un rôle important dans les exploitations minières artisanales et à petite échelle en Afrique subsaharienne. De concert ... Couverture du livre: Une vie saine pour les femmes et les enfants vulnérables · Couverture du livre: Entre el activismo y la intervención · Couverture du livre: Revitalizing Health for All.

Bulletin du CRDI #125

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

L'IOSRS remporte le prix de la diplomatie scientifique · GrowInclusive : la plateforme tant attendue est en construction · Toutes les nouvelles. Activités à venir. Semaine du développement international 2018. Le CRDI célébrera la Semaine du développement international du 4 au 10 février 2018. Suivez-nous sur Twitter et ...

Calculation of the muon contamination in a {pi}- meson beam; Calcul de la contamination en muons d'un faisceau de mesons {pi}

Energy Technology Data Exchange (ETDEWEB)

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

1965-07-01

incidentes. Les deux cas suivants sont traites: 1 - le faisceau est defini par un compteur S{sub 1}: on calcule la contamination en {mu} au passage a travers ce compteur. 2 - le faisceau est defini par 2 compteurs S{sub 1} et S{sub 2}: on calcule la contamination en {mu} au passage a travers les 2 compteurs successivement. Apres avoir pose le probleme dans le premier paragraphe d'introduction, nous avons traite dans le paragraphe II le detail du calcul, en suivant l'ordre du programme. A la fin du paragraphe II, on trouvera les definitions d'un certain nombre de quantites que le programme calcule; ce sont des valeurs de la contamination dans l'un des deux cas precedents, integrees dans certains volumes de desintegration bien definis. Dans le paragraphe III, nous donnons a titre d'exemple un tableau de resultats pour quelques valeurs de parametres. Nous reproduisons en annexe le 'listing' du programme 'CONTAMU'. Ce programme a ete ecrit en 1963 pour effectuer les corrections aux valeurs experimentales des sections efficaces obtenues dans une experience effectuee autour du Synchrotron Saturne par le groupe Falk-Vairant. (auteur)

Profils des porteurs du VIH/SIDA au début du traitement ...

African Journals Online (AJOL)

But : Décrire les profils des porteurs de VIH/sida au début du traitement antirétroviral. Matériels et méthode: Les dossiers des porteurs du VIH/sida de la région maritime ont été analysés de mai 2008 à avril 2009 par le comité thérapeutique. Résultats: Parmi les 641 dossiers analysés, 67,40% venaient du district de Yoto.

Nitrous oxide fluxes from grassland in the Netherlands. 1. Statistical analysis of flux-chamber measurements

NARCIS (Netherlands)

Velthof, G.L.; Oenema, O.

1995-01-01

Accurate estimates of total nitrous oxide (N2O) losses from grasslands derived from flux-chamber measurements are hampered by the large spatial and temporal variability of N2O fluxes from these sites. In this study, four methods for the calculation o

Calculation of the thermal utilisation factor in a cell made up of a given number of concentric media; Calcul du facteur d'utilisation thermique dans une cellule formee d'un nombre quelconque de milieux concentriques

Energy Technology Data Exchange (ETDEWEB)

Amouyal, A; Benoist, P; Guionnet, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

The method of calculating the thermal utilisation factor, described in a previous report, is extended to the case of a cylindrical cell containing a given number of concentric media, certain of which may be empty. A collision by collision method is used in all but the peripheral medium, which may be treated by a theory of controlled diffusion. A programme for the IBM 650 calculator has been based on this method. Some numerical results are presented. An equivalent matrix formulation, due to C. Guionnet, is given as an appendix. (author) [French] La methode de calcul du facteur d'utilisation thermique, exposee dans un rapport precedent, est etendue au cas d'une cellule cylindrique comportant un nombre quelconque de milieux concentriques, certains de ces milieux pouvant etre vides. On utilise une methode choc par choc dans tous les milieux excepte le milieu peripherique qui peut etre traite par une theorie de diffusion ajustee. La methode a fait l'objet d'un programme pour le calculateur IBM 650. Quelques resultats numeriques sont presentes. Une formulation matricielle equivalente, due a C. Guionnet, est exposee en annexe. (auteur)

Crystallographic study of the tempering by irradiation of cold-worked uranium (1960); Etude cristallographique du revenu de l'uranium ecroui par irradiation (1960)

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

We have studied the phenomenon of the tempering of cold-worked uranium under the action of irradiation by observing the narrowing of the (114) and (133) X-ray diffraction lines as a function of the irradiation level. Simultaneously we have studied the broadening of the 114 line of a recrystallised uranium as a function of the irradiation level. The irradiation temperature was always less than 60 deg. C. Of these two processes, the first is the fastest. We have observed a saturation of the irradiation tempering for a flux of 10{sup 18} n/cm{sup 2}; we deduce from this the dimensions of the volume perturbed by one fission atom to be 10{sup -17} cm{sup 3}. (author) [French] Nous avons etudie le phenomene de revenu par irradiation d'echantillons d'uranium ecroui, en observant l'affinement des raies de diffraction de rayons X (114) et (133) en fonction du taux d'irradiation. Parallelement nous avons etudie l'elargissement de la raie (114 ) d'un uranium recristallise en fonction du taux d'irradiation. La temperature d'irradiation est toujours restee inferieure a 60 deg. C. De ces deux processus le premier est le plus rapide. Nous avons observe une saturation du revenu par irradiation pour un flux de 10{sup 18} n/cm{sup 2}; on en deduit une valeur du volume de la perturbation creee par un atome de fission egale a 10{sup -17} cm{sup 3}. (auteur)

Les effets du changement climatique dans le bassin du Congo : la ...

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

22 avr. 2016 ... Dans le bassin du fleuve Congo, plus de 80 % des habitants vivent exclusivement de l'agriculture, de la pêche, de l'élevage et de la cueillette, qui sont des activités largement tributaires du climat.

Dissecting Reactor Antineutrino Flux Calculations

Science.gov (United States)

Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

2017-09-01

Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

Impact du comportement vibratoire et acoustique des joints de portes et de vitrages sur le bruit d'origine aérodynamique

OpenAIRE

Oliver , Clara; Jézéquel , Louis; Besset , Sébastien; VAN HERPE , François; Abbadi , Zouhir

2015-01-01

International audience; Dans la modélisation des joints de portes et de vitrages pour la prédiction du bruit aérodynamique il est nécessaire de sortir du domaine classique du calcul vibro-acoustique linéaire, en raison du caractère fortement non-linéaire des joints. La déformation lors de la fermeture des portes ou vitres provoque des modifications des raideurs statique et dynamique, de l'amortissement et de la transparence acoustique des joints. Ensuite les efforts aérodynamiques modifient c...

Two-dimensional sensitivity calculation code: SENSETWO

International Nuclear Information System (INIS)

Yamauchi, Michinori; Nakayama, Mitsuo; Minami, Kazuyoshi; Seki, Yasushi; Iida, Hiromasa.

1979-05-01

A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)

(Sorghum bicolor (L.) Moench) du Nord du Burkina Faso

African Journals Online (AJOL)

SARAH

29 déc. 2014 ... sorghos à grains sucrés ont un cycle court et arrivent donc à maturité avant les autres sorghos et le mil d'où leur exploitation comme aliment de soudure par les paysans. L'organisation de la diversité morphologique des accessions de sorghos à grains sucrés du Nord du. Burkina autour principalement des ...

Etude de la transition ferroelectrique-ferroelastique du KD2PO4 forme du front de phase en fonction du gradient thermique

OpenAIRE

Kvítek, Zdeněk

2010-01-01

Etude de la transition ferroelectrique-ferroelastique du KD2PO4 forme du front de phase en fonction du gradient thermique The thesis explores complex process of first order transition of KD2PO4 crystal from tetragonal phase to ferroelectric - ferroelastic orthorhombic phase and back at temperature 209 K. The experimental set up of nitrogeneous cryostat allowes temperature and temperature gradient variations during simultaneous three axes optical sample observations, dielectric measurements. T...

Les Cahiers du CREAD

African Journals Online (AJOL)

Admin

politique de bas prix exercée par la Russie et le Qatar vient confirmer ce constat ; s'ajoute à cela l'entrée éventuelle du gaz non conven- tionnel, dont son prix actuel de 3/4 $US, offre aux USA l'opportunité d'être exportateur de ..... les compagnies à produire en matière du gaz naturel, tels le prix du gaz naturel, le prix des ...

Méthode analytique généralisée pour le calcul du coning. Nouvelle solution pour calculer le coning de gaz, d'eau et double coning dans les puits verticaux et horizontaux Generalized Analytical Method for Coning Calculation. New Solution to Calculation Both the Gas Coning, Water Coning and Dual Coning for Vertical and Horizontal Wells

Directory of Open Access Journals (Sweden)

Pietraru V.

2006-11-01

Full Text Available Une nouvelle méthode analytique d'évaluation du coning d'eau par bottom water drive et/ou de gaz par gas-cap drive dans les puits horizontaux et verticaux a été développée pour les réservoirs infinis [1]. Dans cet article, une généralisation de cette méthode est présentée pour les réservoirs confinés d'extension limitée dont le toit est horizontal. La généralisation proposée est basée sur la résolution des équations différentielles de la diffusivité avec prise en compte des effets de drainage par gravité et des conditions aux limites pour un réservoir confiné. La méthode est applicable aux réservoirs isotropes ou anisotropes. L'hypothèse de pression constante à la limite de l'aire de drainage dans l'eau et/ou dans le gaz a été adoptée. Les pertes de charge dans l'aquifère et dans le gas-cap sont donc négligées. Les principales contributions de cet article sont : - L'introduction de la notion de rayon de cône, différent du rayon de puits. La hauteur du cône et le débit critique dépendent du rayon de cône alors qu'ils sont indépendants du rayon du puits. - Une nouvelle corrélation pour le calcul du débit critique sous forme adimensionnelle en fonction de trois paramètres : le temps, la longueur du drain horizontal (nulle pour un puits vertical et le rayon de drainage. - Des corrélations pour le calcul du rapport des débits gaz/huile (GOR ou de la fraction en eau (fw, pendant les périodes critique et postcritique, qui tiennent compte de la pression capillaire et des perméabilités relatives. - Des corrélations pour le calcul des rapports de débits gaz/huile et eau/huile pendant les périodes pré, post et supercritique en double coning. - Des critères pour le calcul du temps de percée au puits en simple coning de gaz ou d'eau, ou en double coning de gaz et d'eau. A new analytical method for assessing water and/or gas coning in horizontal and vertical wells has been developed for infinite

A flux footprint analysis to understand ecosystem fluxes in an intensively managed landscape

Science.gov (United States)

Hernandez Rodriguez, L. C.; Goodwell, A. E.; Kumar, P.

2017-12-01

Flux tower studies in agricultural sites have mainly been done at plot scale, where the footprint of the instruments is small such that the data reveals the behaviour of the nearby crop on which the study is focused. In the Midwestern United States, the agricultural ecosystem and its associated drainage, evapotranspiration, and nutrient dynamics are dominant influences on interactions between the soil, land, and atmosphere. In this study, we address large-scale ecohydrologic fluxes and states in an intensively managed landscape based on data from a 25m high eddy covariance flux tower. We show the calculated upwind distance and flux footprint for a flux tower located in Central Illinois as part of the Intensively Managed Landscapes Critical Zone Observatory (IMLCZO). In addition, we calculate the daily energy balance during the summer of 2016 from the flux tower measurements and compare with the modelled energy balance from a representative corn crop located in the flux tower footprint using the Multi-Layer Canopy model, MLCan. The changes in flux footprint over the course of hours, days, and the growing season have significant implications for the measured fluxes of carbon and energy at the flux tower. We use MLCan to simulate these fluxes under land covers of corn and soybeans. Our results demonstrate how the instrument heights impact the footprint of the captured eddy covariance fluxes, and we explore the implication for hydrological analysis. The convective turbulent atmosphere during the daytime shows a wide footprint of more than 10 km2, which reaches 3km length for the 90% contribution, where buoyancy is the dominant mechanism driving turbulence. In contrast, the stable atmosphere during the night-time shows a narrower footprint that goes beyond 8km2 and grows in the direction of the prevalent wind, which exceeds 4 km in length. This study improves our understanding of agricultural ecosystem behaviour in terms of the magnitude and variability of fluxes and

Monte-Carlo method applied to the energy loss calculation of the gamma rays isotropic flux in the NaI(tau l) cylindrical scintillator between 0.5-20 MeV

International Nuclear Information System (INIS)

Martin, I.M.; Dutra, S.L.G.; Palmeira, R.A.R.

1975-01-01

Using the 'Monte Carlo' method, a determination was made of the response function of a NaI cylindrical crystal when exposed to an omnidirectional γ ray flux in the range 0.5 - 20 MeV. Improvements over previous similar calculations include considerations of the bremsstrahlung and multiple scattering processes in the slowing down of the secondary electrons. These calculations will be applied to the problem of determining the energy spectrum of an incident gamma ray flux from the measured response of the crystal in the space [pt

Three-Dimensional Temperature Field Calculation and Analysis of an Axial-Radial Flux-Type Permanent Magnet Synchronous Motor

Directory of Open Access Journals (Sweden)

Dong Li

2018-05-01

Full Text Available This article concentrates on the steady-state thermal characteristics of the Axial-Radial Flux-Type Permanent Magnet Synchronous Motor (ARFTPMSM. Firstly, the three-dimensional mathematical models for electromagnetic calculation and analyses are established, and the machine loss, including the stator loss, armature winding loss, rotor loss, and axial structure loss is calculated by using time-step Finite Element Method (FEM. Then, the loss distribution is assigned as the heat source for the thermal calculation. Secondly, the mathematical model for thermal calculation is also established. The assumptions and the boundary conditions are proposed to simplify the calculation and to improve convergence. Thirdly, the three-dimensional electromagnetic and thermal calculations of the machine, of which the armature winding and axial field winding are developed by using copper wires, are solved, from which the temperature distributions of the machine components are obtained. The experiments are carried out on the prototype with copper wires to validate the accuracy of the established models. Then, the temperature distributions of machine components under different Axial Magnetic Motive Force (AMMF are investigated. Since the machine is finally developing by using HTS wires, the temperature distributions of machine developed by utilizing High Temperature Superconducting (HTS wires, are also studied. The temperature distribution differences of the machine developed by using copper wires and HTS wires are drawn. All of these above will provide a helpful reference for the thermal calculation of the ARFTPMSM, as well as the design of the HTS coils and the cryogenic cooling system.

Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4

International Nuclear Information System (INIS)

Baur, A.; Dupont, C.; Totth, B.

1978-01-01

The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks

Création du groupe FMR (Flux, Matrices, Réseaux

Directory of Open Access Journals (Sweden)

Laurent Beauguitte

2013-06-01

Full Text Available Laurent Beauguitte et César Ducruetont pris l’initiative de créer un groupe de travail sur les réseaux en vue d’intégrer dans l’approche géographique les outils et méthodes issues de l’analyse des réseaux (planaires, sociaux, complexes. Ces derniers paraissent en effet fournir des pistes intéressantes pour l’analyse des flux de toute nature (commerciaux, humains, informationnels, financiers. Ce groupe tient une réunion mensuelle depuis le 26 octobre 2010 à Paris. Trois axes de travail sont...

L'action du CRDI — le développement du secteur privé

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

Cathy Egan

Le CRDI est du nombre. Il reconnaît depuis longtemps la valeur de l'industrie, des échanges et du commerce comme moteurs de la croissance économique. Grâce au soutien concret qu'il a accordé au développement du secteur privé, le CRDI a appris que deux types d'aide peuvent porter fruit : améliorer le con- texte dans ...

Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

2010-01-01

In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

Two further years of operation of the reactor G1 (july 1958 - july 1960); Deux nouvelles annees de fonctionnement du reacteur G1. (juillet 1958 - Juillet 1960)

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

The aim of the present report is to present certain observations and to give the results obtained during the period from july the 1{sup st} 1958 to july the 1{sup st} 1960. The main operations carried out during this period were, chronologically: - From july the 5{sup th} to october the 18{sup th} 1958: preparation and execution of the first annealing of the graphite. - From dec. the 15{sup th} 1958 to july the 15{sup th} 1959: a discharging campaign which resulted in the complete renewal of the fuel elements. During the monthly stoppages of this campaign, it was possible to make certain observations concerning the packing of the graphite, while at the same time measurements of the temperature of the element cans were made at an increased number of points. - From september the 25{sup th} 1959 to december the 9{sup th} 1959: preparation and execution of the second annealing. At the end of the annealing, the thorium lattice was modified and extra thermocouples were installed for measuring the temperature of the body of the graphite. An apparatus was built for measuring the radial flux. - From december the 9{sup th} 1959 to july 1960: a continuous operation campaign, with a minimum of stoppages. The experimental results are re-assembled, independently of their chronological order, under three main headings which describe the reactors history: - continuous operation, - discharges, - annealing of the reactor. (author) [French] Le but du present rapport est d'exposer certaines observations faites et les resultats obtenus au cours de la periode du 1{sup er} juillet 1958 au 1{sup er} juillet 1960. Cette periode a ete marquee chronologiquement par les operations essentielles suivantes: - du 5 juillet au 18 octobre 1958: preparation et execution du premier recuit du graphite. - du 15 decembre 1958 au 15 juillet 1959: campagne de dechargement entrainant un renouvellement total des cartouches de combustibles. Au cours des arrets mensuels de cette campagne, certaines

Measurements and calculations of neutron fluxes through a simulation of the CRBR upper axial shielding

International Nuclear Information System (INIS)

Maerker, R.E.; Muckenthaler, F.J.

1976-01-01

Measurements, using a 4-in. Bonner Ball, have been made of the neutron fluxes penetrating a simulation of CRBR upper axial biological shielding at the Tower Shielding Facility. The simulation consisted of a 45.7 cm thick slab of SS-304 followed by a series of sodium tanks having a total thickness of 457 cm followed by slabs of carbon steel up to 61.0 cm thick. Measurements were made behind the stainless steel, behind intermediate thicknesses of 152 cm, 305 cm, and 457 cm of sodium (with the stainless steel in place), and behind various thicknesses of the carbon steel following both 305 cm and 457 cm of sodium (also with the stainless steel in place). Calculated and measured data are presented and compared

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

Energy Technology Data Exchange (ETDEWEB)

Vermeeren, L. [SCK/CEN, B-2400 Mol (Belgium)

2001-07-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Absolute on-line in-pile measurement of neutron fluxes using self-powered neutron detectors: Monte Carlo sensitivity calculations

International Nuclear Information System (INIS)

Vermeeren, L.

2001-01-01

Self-powered neutron detectors (SPND) are well suited to monitor continuously the neutronic operating conditions of driver fuel of research reactors and to follow its burnup evolution. This is of particular importance when advanced or new MTR fuel designs need to be qualified. We have developed a detailed MCNP-4B based Monte Carlo approach for the calculation of neutron sensitivities of SPNDs. Results for the neutron sensitivity of a Rh SPND are in excellent agreement with experimental data recently obtained at the BR2 research reactor. A critical comparison of the Monte Carlo results with results from standard analytical methods reveals an important deficiency of the analytical methods in the description of the electron transport efficiency. Our calculation method allows a reliable on-line determination of the absolute in-pile neutron flux. (author)

Application of the Bowring correlation for calculating the critical heat flux

International Nuclear Information System (INIS)

Borges, R.C.; Freitas, R.L.

1986-01-01

The evaluation of the critical heat flux is of great importance for the nuclear reactor project, because it permits the verification of the safety margin with respect to fuel rod damage. This work presents a comparison of the original critical heat flux correlation proposed by Bowring with an alternative form derived from it presented in several papers. Very different results have been encountered from the application of the two correlation forms. Therefore, a criterious choice of the correlation form must be done avoid the violation of the project's safety margin. (Author) [pt

Identification des matériaux et étude du bâti : l’exemple du Clos du Cotentin

Directory of Open Access Journals (Sweden)

Julien Deshayes

2012-04-01

Full Text Available Partant de l’architecture vernaculaire de la Presqu’île du Cotentin, cette étude s’attache à montrer comment l'identification des matériaux de construction, replacée dans une problématique historique, peut fournir un outil d'analyse privilégié du bâti, notamment en matière de chronologie. L'identification des principaux centres carriers et de leur aire d'exportation permet de mieux appréhender l'évolution des modénatures et de raisonner sur des typologies cohérentes. L'acquisition de tels critères d'analyse, peu généralisables hors de régions très délimitées, est nécessairement liée à une expérience prolongée du terrain. L’exemple du Clos du Cotentin offre, en matière d'identification des matériaux et d'étude du bâti, un manifeste en faveur d'une approche topographique approfondie, susceptible de replacer l'architecture rurale dans le cadre d'une évolution historique.Based on the vernacular architecture of the Cotentin peninsula, this article sets out to show how the identification of construction materials, placed in a historical context, can offer a particularly useful tool for analysing buildings, notably for their chronology. The identification of the principle stone quarries and their zones of commercial influence gives information on the evolution of architectural proportions and permits the development of coherent typologies. The acquisition of such analytical criteria is difficult to generalise beyond strictly limited regions and are necessarily linked with in-depth familiarity with the territory concerned. The example that the Clos du Cotentin offers for the identification of building materials and the study of buildings may be seen as a manifesto in favour of a detailed topographical approach, capable of placing rural architecture in the framework of its historical evolution.

Thermal and hydrodynamic study of a whirling liquid hydrogen layer under high heat flux; Etude thermique et hydrodynamique d'une couche tourbillonnaire d'hydrogenen liquide sous flux de chaleur eleve

Energy Technology Data Exchange (ETDEWEB)

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

1968-03-01

In order to achieve a cold neutrons source ({lambda} {>=} 4.10{sup -10} m) in a high flux reactor ({approx} 10{sup 15} neutrons/cm{sup 2}.s), a whirling liquid hydrogen layer (145 mm OD, effective thickness 15 mm, height about 180 mm) was formed, out-of-pile, in a cylindrical transparent glass vessel. The whirling motion was obtained by tangential injection of the liquid, near the wall. Thermal and hydrodynamical conditions of formation and laws of similarity of such a layer were studied. The characteristics of this whirling flow were observed as a function of mass flow rate (5 to 27 g/s; 4.3 to 23 l/mn), and of spillway width (18 and 25 mm). Six different nozzles were used : 1.0; 1.5; 1.9; 2.25; 2.65 and 3.0 mm ID. The total heat influx was found between 8.6 and 10.4 kW. The heat flux density was about 9.4 W/cm{sup 2} and the mean layer density around 80 per cent of that of the liquid hydrogen at 20.4 Kelvin. High speed movies were used to analyze the boiling regime. (author) [French] En vue de realiser une source de neutrons froids ({lambda} {>=} 4.10{sup -10} m) dans un reacteur a haut flux ({approx} 10{sup 15} neutrons thermiques/cm{sup 2}.s), on a forme dans un vase cylindrique transparent en verre, hors-pile, une couche tourbillonnaire ('vortex') d'hydrogene liquide (diametre exterieur 145 mm, epaisseur effective 15 mm, hauteur 180 mm environ). Le mouvement giratoire est obtenu par injection tangentielle du liquide pres de la paroi. L'etude porte sur la determination des conditions thermiques et hydrodynamiques de la formation d'une telle couche et sur les regles de similitude de ce phenomene. On a observe les caracteristiques de l'ecoulement giratoire en fonction du debit (de 5 a 27 g/s, soit de 4.3 a 23 1/mn), de la vitesse d'injection (entre 10 et 110 m/s) et de la largeur du deversoir (18 et 25 mm), ceci pour six diametres differents d'injecteur (1.0 ; 1.5; 1.9; 2.25; 2.65 et 3.0 mm). Le flux de chaleur total mesure est compris entre 8.6 et 10.4 kW et la

Le discours rappporté comme effet de montage du discours citant et du segment citationnel. Contribution à l’étude du discours journalistique

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Biardzka Elżbieta

2012-07-01

Full Text Available Dans notre étude, nous nous proposons un vaste retour aux données empiriques qui nous permettra de repérer et de décrire un dispositif sémantico-énonciatif et grammatical qui est à l'origine de la prolifération de pratiques du discours rapporté (désormais DR, codifiées et non codifiées, particulièrement répandues dans la presse écrite. Nous avons calculé nos résultats de recherche sur l'analyse d'un corpus journalistique d'à peu près 1500 exemples. Nous considérons le DR comme une séquence textuelle binaire, embrassant deux segments: le discours citant (DC qui verbalise les données situationnelles de l'énonciation primaire, et le segment citationnel (Cit qui représente les paroles. Compris de la sorte, le DR peut se figurer sous l'équation suivante: DR= DC + Cit. Nous envisageons le DC et la Cit comme des sortes de briques, mettons comme des Lego, qui entrent dans plusieurs combinaisons possibles pour donner naissance aux séquences du DR. Pour étudier le mécanisme qui les engendre, nous inventorions et décrivons les propriétés formelles et sémantico-énonciatives des segments DC et Cit. L'inventaire des formes grammaticales que peut revêtir le DC embrasse 6 cas de figures: la phrase introductive, la phrase complète, l'interrogation, l'incise, les syntagmes en "selon A", ensuite d'autres syntagmes, comme les syntagmes nominaux, adjectivaux, représentant des cas de différentes "incomplétudes" formelles et de réductions. La Citation, terme générique dans notre étude, recouvre trois types d'occurrences: les Citations reproductions, les Citations reformulations et les Citations mixtes. Les relations syntaxiques entres les deux segments du DR dépendent de la forme du DC et se définissent par deux cas de figure. Soit le journaliste rapporte les paroles de différentes personnes sans les assimiler à la syntaxe de son propre énoncé (combinatoire libre, soit il les intègre dans son discours conformément aux

Fast neutron fluxes distribution in Egyptian ilmenite concrete

International Nuclear Information System (INIS)

Megahed, R.M.; Abou El-Nasr, T.Z.; Bashter, I.I.

1978-01-01

This work is concerned with the study of the distribution of fast neutron fluxes in a new type of heavy concrete made from Egyptian ilmenite ores. The neutron source used was a collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor. Measurements were carried-out using phosphorous activation detectors. Iso-flux curves were represented which give directly the shape and thickness required to attenuate the emitted fast neutron flux to a certain value. The relaxation lengths were also evaluated from the measured data for both disc monodirectional source and infinite plane monodirectional source. The obtained values were compared with that calculated using the derived values of relative number densities and microscopic removal cross-sections of the different constituents. The obtained data show that ilmenite concrete attenuates fast neutron flux more strongly than ordinary concrete. A semiemperical formula was derived to calculate the fast neutron flux at different thicknesses along the beam axis. Another semiemperical formula was also derived to calculate the fast neutron flux in ordinary concrete along the beam axis using the corresponding value in ilmenite concrete

Some Problems of Calculation and Design of High-Activity Gamma Units; Quelques aspects du calcul et de l'etablissement de projets d'installations puissantes emettrices de rayonnement gamma; Nekotorye voprosy rascheta i proektirovaniya moshchnykh gamma-ustanovok; Algunos aspectos del calculo y construccion de instalaciones de irradiacion gamma de elevada intensidad

Energy Technology Data Exchange (ETDEWEB)

Bibergal, A V; Leshchinsky, N I; Margulis, U Ya; Khrushev, V G [Academy of Sciences of the USSR, Moscow, Union of Soviet Socialist Republics (Russian Federation)

1960-07-15

The report describes the principal requirements for gamma units intended for various purposes. Several methods of calculating the dose fields for various forms of irradiators are given, as well as graphs, nomograms and formulae to estimate the amount of gamma-ray dose absorbed by the irradiated object from cobalt-60 and caesium-137. Some of the calculated data have been confirmed by experiment. The advantages of irradiators of various geometry employed in experimental and commercial units are discussed. The irradiation technique for various objects is analyzed and the optimum irradiation conditions (radiation utilization factor, dose field homogeneity, etc.) are discussed. Several rational shielding systems are suggested to simplify irradiation process and recharging, and to reduce the cost of design and operation. (author) [French] Resume Cette communication expose les conditions fondamentales auxquelles doivent satisfaire les installations emettrices de rayonnement gamma, destinees, a divers usages. Elle explique certaines methodes pour le calcul des champs recevant une dose determinee en fonction de la forme du dispositif d'irradiation; elle contient des courbes, des nomogrammes et des formules permettant de calculer la valeur de la dose de rayons gamma du cobalt-60 et du cesium-137 recue par l'echantillon soumis a l'irradiation. Certaines donnees etablies par le calcul ont ete confirmees experimentalement. On examine l'utilite d'employer des dispositifs d'irradiation de configurations variees dans des installations experimentales et des installations industrielles. Les auteurs examinent la technologie de l'irradiation de divers echantillons en vue de choisir les conditions optimales d'irradiation (facteur d'utilisation du rayonnement, uniformite du champ recevant une dose determinee, etc.). Quelques systemes rationnels de protection, simplifiant le processus d'irradiation, sont exposes ainsi que des methodes permettant un echange de charge dans les

Dynamique des populations du foreur des tiges du cacaoyer ...

African Journals Online (AJOL)

Objectif : Le foreur de tiges du cacaoyer, Eulophonotus myrmeleon Felder cause aujourd'hui d'énormes dégâts dans les cacaoyères ivoiriennes. La présente étude vise à déterminer les périodes de fortes attaques de ce déprédateur dans la région du Haut-Sassandra, la deuxième plus grande région de production de ...

Optimization of Hybrid Power Trains by Mechanistic System Simulations Optimisation de groupes motopropulseurs électriques hybrides par simulation du système mécanique

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Katrašnik T.

2013-05-01

écifiques au domaine et sur un couplage inter-domaines cohérent ultérieur des flux. En outre, une structure de calculs efficace permettant la simulation du transport d’espèces gazeuses actives et passives est introduite de manière à combiner l’efficacité des calculs à la nécessité d’une modélisation du transport des polluants clans le circuit des gaz. L’applicabilité et la versatilité du modèle de simulation au niveau mécanique sont présentées au moyen d’analyses des phénomènes transitoires provoqués par l’interdépendance élevée des sous-systèmes, c’est-à-dire les domaines. Les résultats obtenus pour les véhicules hybrides sont comparés â ceux obtenus pour les véhicules conventionnels afin de mettre l’accent sur les différences des régimes opératoires de composants particuliers inhérents à une topologie particulière du groupe motopropulseur.

Methode des Rayons avec Calcul d’Intensite Appliquee a la Propagation Anormale (Method of Ray Intensity Calculation as Applied to Anomalous Propagation),

Science.gov (United States)

1986-06-01

plus rapproch~es. Cette question, qui constitue en elle-mgme un domaine de recherche, d~ borde du cadre du present rapport. * Il est possible d’gtendre...par la commande JCL "PROPEX:T.’ qui produit le module ex~cutable "PROPAG:EX". TABLEAU I Bloc principal PROPAK *NOM DES ROUTINES FONCTION PROPAG...alise la r &mission des rayons r~fl~chis e77 SANS CLASSIFICATION 22 TABLEAU II Bloc utilitaire UTILPAK NOM DES ROUTINES FONCTION H - Calcule r la

Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

2016-01-01

In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

Toward an estimation of daily european CO{sub 2} fluxes at high spatial resolution by inversion of atmospheric transport; Vers une estimation des flux de CO{sub 2} journaliers europeens a haute resolution par inversion du transport atmospherique

Energy Technology Data Exchange (ETDEWEB)

Carouge, C

2006-04-15

distance between pixels, climate and vegetation distribution over Europe. To study the potential of this method, we used synthetic data generated from forward simulations of LMDZt (driven by flux fields generated from the biosphere model ORCHIDEE). We have found that the current network is not dense enough to constrain fluxes at model resolution. However, fluxes that are aggregated spatially over a region of 850 x 850 km in the Western Europe and temporally over 8-10 days compare very well with the ORCHIDEE fluxes. Preliminary inversion results using real data indicate that synoptic variations of estimated fluxes are in phase with the variations of the ORCHIDEE biosphere model flux and the variations observed in atmospheric concentrations. However, the quality of the flux estimates are highly dependant on transport model errors and in particular, on the quality of modelling small scale transport. Moreover, fossil fuel emissions are prescribed in this inverse model and the quality of their distribution is shown to be crucial. Data selection also has a large impact on estimated fluxes. The use of the daytime only data to calculate daily averaged concentrations greatly improves the estimated fluxes by reducing bias inferred from model transport errors. (author)

Nouvelles approches en theorie du champ moyen dynamique: le cas du pouvoir thermoelectrique et celui de l'effet orbital d'un champ magnetique

Science.gov (United States)

Arsenault, Louis-Francois

Les applications reliees a la generation d'energie motivent la recherche de materiaux ayant un fort pouvoir thermoelectrique (S). De plus, S nous renseigne sur certaines proprietes fondamentales des materiaux, comme, par exemple, la transition entre l'etat coherent et incoherent des quasi-particules lorsque la temperature augmente. Empiriquement, la presence de fortes interactions electron-electron peut mener a un pouvoir thermoelectrique geant. Nous avons donc etudie le modele le plus simple qui tient compte de ces fortes interactions, le modele de Hubbard. La theorie du champ moyen dynamique (DMFT) est tout indiquee dans ce cas. Nous nous sommes concentres sur un systeme tridimensionnel (3d) cubique a face centree (fcc), et ce, pour plusieurs raisons. A) Ce type de cristal est tres commun dans la nature. B) La DMFT donne de tres bons resultats en 3d et donc ce choix sert aussi de preuve de principe de la methode. C) Finalement, a cause de la frustration electronique intrinseque au fcc, celui-ci ne presente pas de symetrie particule-trou, ce qui est tres favorable a l'apparition d'une grande valeur de S. Ce travail demontre que lorsque le materiau est un isolant a demi-remplissage a cause des fortes interactions (isolant de Mott), il est possible d'obtenir de grands pouvoirs thermoelectriques en le dopant legerement. C'est un resultat pratique important. Du point de vue methodologique, nous avons montre comment la limite de frequence infinie de S et l'approche dite de Kelvin, qui considere la limite de frequence nulle avant la limite thermodynamique pour S, donnent des estimations fiables de la vraie limite continue (DC) dans les domaines de temperature appropriee. Ces deux approches facilitent grandement les calculs en court-circuit ant la necessite de recourir a de problematiques prolongements analytiques. Nous avons trouve que la methode de calcul a frequence infinie fonctionne bien lorsque les echelles d'energie sont relativement faibles. En d'autres termes

Le Developmental Entrepreneurship Program du Massachusetts ...

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

Faire en sorte que les étudiants venant des pays en développement pour étudier au MIT retournent ensuite dans leurs pays respectifs afin d'y concrétiser leurs idées est un souci majeur du MIT, des bailleurs de fonds et du milieu du développement. Ce projet palliera à ce souci en soutenant la formation des fellows du ...

Differential systems of flux and concentrations of a mixture in the separation by a plant (1963); Systemes differente des flux et des concentrations d'un melange en separation par une usine etagee (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-11-15

The study of transient flux and concentrations of a mixture in the separation by a plant with different interdependent stages leads to an examination of the solutions of certain types of differential systems. These systems are obtained from the representative graph of the plant and have a structural form. By Olga TAUSSKY's theorem and the introduction of steerable distances, the solutions of these systems are chiefly examined in their asymptotic behaviour. The appendix shows that solutions of partial differential equations relative to plants made of cascades with a slight separation have analogous properties. (author) [French] La recherche des regimes transitoires des flux et des concentrations d'un melange en separation par une usine etagee de forme quelconque conduit a etudier les solutions de certains types de systemes differentiels. Ces systemes sont deduits du graphe representatif de l'usine, et apparaissent sous une forme structuree. A l'aide du theoreme d'Olga TAUSSKY et par l'introduction d'ecarts orientes, les solutions de ces systemes sont principalement examinees dans leur comportement asymptotique. Un appendice montre que les solutions des systemes d'equations aux derivees partielles des usines en cascades a tres faible separation admettent des proprietes analogues. (auteur)

MOTUS, MEUTE, ÉMEUTE: FORMES DU MOUVEMENT SAUVAGE

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Diogo Sardinha

2015-07-01

Full Text Available L'article traite de la politique comme un mouvement ou flux de soulèvement indescriptible en termes d'identité, car elle est distincte de la stabilisation et de la domination du pouvoir constitué. L'article analyse l'ordre et le désordre, la raison et les émeutes, le gouvernement et la révolte, de la police et de la politique en vue de réfléchir sur l'identité comme une normalisation. De différents auteurs, le texte aborde les sujets suivants: 1. Une comparaison historique; 2. Espace mystique et de l'espace logique: Wittgenstein Althusser et 3. Désidentification: l '"Je est un autre” à “ni moi, ni l'autre".

Improving the Terrain-Based Parameter for the Assessment of Snow Redistribution in the Col du Lac Blanc Area and Comparisons with TLS Snow Depth Data

Science.gov (United States)

Schön, Peter; Prokop, Alexander; Naaim-Bouvet, Florence; Nishimura, Kouichi; Vionnet, Vincent; Guyomarc'h, Gilbert

2014-05-01

Wind and the associated snow drift are dominating factors determining the snow distribution and accumulation in alpine areas, resulting in a high spatial variability of snow depth that is difficult to evaluate and quantify. The terrain-based parameter Sx characterizes the degree of shelter or exposure of a grid point provided by the upwind terrain, without the computational complexity of numerical wind field models. The parameter has shown to qualitatively predict snow redistribution with good reproduction of spatial patterns, but has failed to quantitatively describe the snow redistribution, and correlations with measured snow heights were poor. The objective of our research was to a) identify the sources of poor correlations between predicted and measured snow re-distribution and b) improve the parameters ability to qualitatively and quantitatively describe snow redistribution in our research area, the Col du Lac Blanc in the French Alps. The area is at an elevation of 2700 m and particularly suited for our study due to its constant wind direction and the availability of data from a meteorological station. Our work focused on areas with terrain edges of approximately 10 m height, and we worked with 1-2 m resolution digital terrain and snow surface data. We first compared the results of the terrain-based parameter calculations to measured snow-depths, obtained by high-accuracy terrestrial laser scan measurements. The results were similar to previous studies: The parameter was able to reproduce observed patterns in snow distribution, but regression analyses showed poor correlations between terrain-based parameter and measured snow-depths. We demonstrate how the correlations between measured and calculated snow heights improve if the parameter is calculated based on a snow surface model instead of a digital terrain model. We show how changing the parameter's search distance and how raster re-sampling and raster smoothing improve the results. To improve the parameter

"Cirque du Freak."

Science.gov (United States)

Rivett, Miriam

2002-01-01

Considers the marketing strategies that underpin the success of the "Cirque du Freak" series. Describes how "Cirque du Freak" is an account of events in the life of schoolboy Darren Shan. Notes that it is another reworking of the vampire narrative, a sub-genre of horror writing that has proved highly popular with both adult and…

Precise calculation of the transmission coefficient of a potential barrier. Study of the error in the B K W approximation; Calcul exact du coefficient de transmission d'une barriere de potentiel. Etude de l'erreur de l'approximation B K W

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

Following on to work started in a previous report, the author carries out in the case of few examples, the calculation of the transmission coefficient T using accurate methods. He then deduces from this the error in the B K W method. The calculations are carried out for values of T ranging down to 10{sup -200}. The use of modern computers makes it possible to obtain values of T to eight decimal places in a few seconds and the practical advantage of the B K W approximation appears therefore considerably reduced. The author gives also a method which may be used for an exact calculation of the energy levels of a potential well. (author) [French] Poursuivant une etude commencee dans une note anterieure, l'auteur effectue, sur quelques exemples, le calcul du coefficient de transmission T par des methodes exactes. Il en deduit ensuite l'erreur de la methode B K W. Les calculs sont faits pour des valeurs de T allant jusqu'a 10{sup -200}. L'utilisation des machines a calculer modernes permettant d'obtenir en quelques secondes, la valeur de T avec 8 decimales exactes, l'interet pratique de l'approximation B K W semble considerablement diminue. L'auteur indique egalement une methode qui peut servir a calculer exactement les niveaux d'energie d'un puits de potentiel. (auteur)

Surface fluxes in heterogeneous landscape

Energy Technology Data Exchange (ETDEWEB)

Bay Hasager, C

1997-01-01

The surface fluxes in homogeneous landscapes are calculated by similarity scaling principles. The methodology is well establish. In heterogeneous landscapes with spatial changes in the micro scale range, i e from 100 m to 10 km, advective effects are significant. The present work focus on these effects in an agricultural countryside typical for the midlatitudes. Meteorological and satellite data from a highly heterogeneous landscape in the Rhine Valley, Germany was collected in the large-scale field experiment TRACT (Transport of pollutants over complex terrain) in 1992. Classified satellite images, Landsat TM and ERS SAR, are used as basis for roughness maps. The roughnesses were measured at meteorological masts in the various cover classes and assigned pixel by pixel to the images. The roughness maps are aggregated, i e spatially averaged, into so-called effective roughness lengths. This calculation is performed by a micro scale aggregation model. The model solves the linearized atmospheric flow equations by a numerical (Fast Fourier Transform) method. This model also calculate maps of friction velocity and momentum flux pixel wise in heterogeneous landscapes. It is indicated how the aggregation methodology can be used to calculate the heat fluxes based on the relevant satellite data i e temperature and soil moisture information. (au) 10 tabs., 49 ills., 223 refs.

Suivi après le traitement du cancer du sein

Science.gov (United States)

Sisler, Jeffrey; Chaput, Geneviève; Sussman, Jonathan; Ozokwelu, Emmanuel

2016-01-01

Résumé Objectif Offrir aux médecins de famille un résumé des recommandations fondées sur les données probantes pour guider les soins aux survivantes traitées pour le cancer du sein. Qualité des données Une recherche documentaire a été effectuée dans MEDLINE entre 2000 et 2016 à l’aide des mots-clés anglais suivants : breast cancer, survivorship, follow-up care, aftercare, guidelines et survivorship care plans, en se concentrant sur la revue des lignes directrices publiées récemment par les organismes nationaux de cancérologie. Les données étaient de niveaux I à III. Message principal Les soins aux survivantes comportent 4 facettes : surveillance et dépistage, prise en charge des effets à long terme, promotion de la santé et coordination des soins. La surveillance des récidives ne se traduit que par une mammographie annuelle, et le dépistage d’autres cancers doit suivre les lignes directrices basées sur la population. La prise en charge des effets à long terme du cancer et de son traitement aborde des problèmes courants tels la douleur, la fatigue, le lymphœdème, la détresse et les effets indésirables des médicaments, de même que les préoccupations à long terme comme la santé du cœur et des os. La promotion de la santé met en relief les bienfaits de l’activité chez les survivantes du cancer, avec l’accent mis sur l’activité physique. Les soins aux survivantes sont de meilleure qualité lorsque divers services et professionnels de la santé participent aux soins, et le médecin de famille joue un rôle important dans la coordination des soins. Conclusion Les médecins de famille sont de plus en plus souvent les principaux fournisseurs de soins de suivi après le traitement du cancer du sein. Le cancer du sein doit être considéré comme une affection médicale chronique, même chez les femmes en rémission, et les patientes profitent de la même approche que celle utilisée pour les autres affections chroniques en

ETUDE DU COMPORTEMENT MECANIQUE DU BETON CELLULAIRE AUTOCLAVE PRODUIT EN ALGERIE

Directory of Open Access Journals (Sweden)

R BELOUETTAR

2002-12-01

Full Text Available Ce travail présente une étude expérimentale du comportement mécanique du béton cellulaire autoclavé. L’étude est portée essentiellement sur une série d’essais mécaniques en compression quasistatique à différentes vitesses de déformation variables entre 10-4 s-1 et 10 s-1 et à deux états différents (état sec et état saturé d’eau. En général, l’augmentation de la vitesse de déformation donne une augmentation de la contrainte critique du béton cellulaire autoclavé. Le béton cellulaire autoclavé présente une sensibilité à la vitesse de déformation positive. La valeur du module d’élasticité est proche de la valeur standard (1.5 GPa pour un béton cellulaire autoclavé de masse volumique égale à 500-550 kg/m3.

Numerical effects in the neutron flux calculations into WWER-type reactor vessels by Monte Carlo method

International Nuclear Information System (INIS)

Alvarez Cardona, C.M.; Rodriguez Gual, M.; Hernandez Valle, S.

2001-01-01

The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested. (authors)

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

Dynamique des populations du foreur des tiges du cacaoyer ...

African Journals Online (AJOL)

SARAH

30 nov. 2014 ... déprédateur dans la région du Haut-Sassandra, la deuxième plus grande région de production de cacao en. Côte d'ivoire. Méthodologie et résultats : L'étude a été réalisée de 2009 à 2013 dans les plantations villageoises de la région du Haut-Sassandra en Côte d'Ivoire. Les variations des taux d'attaques ...

Unfolded ANO-1 fluxes using the LEPRICON methodology

International Nuclear Information System (INIS)

Wagschal, J.J.; Maerker, R.E.; Broadhead, B.L.; Williams, M.L.

1985-01-01

Calculated fluxes at a one-quarter depth midplane location in the pressure vessel of the ANO-1 reactor were adjusted using the LEPRICON methodology. The adjustment was based on the LEPRICON benchmark database in combination with the ANO-1 cavity activation measurements. The LEPRICON database has been expanded and now also includes the PSF field, of which the analysis has been recently completed. The measured, calculated and adjusted ANO-1 activations are presented together with their corresponding uncertainties. However, the main purpose of this work is to obtain more credible flux values in the pressure vessel. The pressure-vessel flux adjustments and the reduction in the pressure vessel flux uncertainties are elaborated on in the concluding section

Plasma crowbars in cylindrical flux compression experiments

International Nuclear Information System (INIS)

Suter, L.J.

1979-01-01

We have done a series of one- and two-dimensional calculations of hard-core Z-pinch flux compression experiments in order to study the effect of a plasma on these systems. These calculations show that including a plasma can reduce the amount of flux lost during the compression. Flux losses to the outer wall of such experiments can be greatly reduced by a plasma conducting sheath which forms along the wall. This conducting sheath consists of a cold, dense high β, unmagnetized plasma which has enough pressure to balance a large field gradient. Flux which is lost into the center conductor is not effectively stopped by this plasma sheath until late in the implosion, at which time a layer similar to the one formed at the outer wall is created. Two-dimensionl simulations show that flux losses due to arching along the sliding contact of the experiment can be effectively stopped by the formation of a plasma conducting sheath

La longue marche du mouvement sportif camerounais : l'émergence du Comité Olympique Camerounais (1946–1964

Directory of Open Access Journals (Sweden)

Charitas Pascal

2016-01-01

Dans ces conditions et selon le statut particulier de ce territoire du « pré-carré » africain de la France, il s'agit de comprendre les étapes de l'institutionnalisation du sport camerounais au sein des atermoiements de la colonisation française puis les processus à l'œuvre dans l'émergence puis la création du Comité Olympique Camerounais (COC, symbole de l'émancipation du sport camerounais et marqueur de son intégration sur la scène sportive internationale, plus particulièrement olympique. Autrement dit, de 1946 et la mise sous tutelle par l'ONU du Cameroun français pour sa partie orientale à 1964 et la première participation du CNO du Cameroun aux Jeux olympiques de Tokyo, l'objectif est de saisir ici le passage d'un sport colonial à un sport postcolonial mettant en exergue le rôle de l'influence française et le rôle des cadres camerounais du secteur sportif.

Sådan kan du bruge optioner og futures

DEFF Research Database (Denmark)

Kohl, Niklas

2017-01-01

Masterclass. Har du brug for at forsikre værdierne i din portefølje i en periode, hvor du forventer stor uro på finansmarkederne, kan du gøre det med optioner eller futures. Her kan du læse om de forskellige typer – og om de muligheder de giver dig.......Masterclass. Har du brug for at forsikre værdierne i din portefølje i en periode, hvor du forventer stor uro på finansmarkederne, kan du gøre det med optioner eller futures. Her kan du læse om de forskellige typer – og om de muligheder de giver dig....

Paysans du Brésil

Directory of Open Access Journals (Sweden)

Dominique Temple

2007-01-01

Full Text Available Eric Sabourin, « Paysans du Brésil : Entre échange marchand et réciprocité » Paris, Editions Quae, 241p, 30 euros, (préface de Maxime Haubert, 2007Dans la présentation du livre, Maxime Haubert dit :«Cet ouvrage propose une analyse socio-anthropologique et agronomique des sociétés rurales et paysannes du Brésil et des transformations qu'elles ont connues ces dernières décennies, en particulier face aux interventions de l'Etat et à l'expansion du marché capitaliste (.... «Le livre pose d'abor...

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1977-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount

EVALUATION DE L’IMPACT ENVIRONNEMENTAL : Evaluation des impacts du flux de transgènes de tolérance à différents herbicides à large spectre

Directory of Open Access Journals (Sweden)

Astoin Marie-Florence

2000-07-01

Full Text Available Ce texte est tiré du rapport « Introduction de variétés génétiquement modifiées de colza tolérantes à différents herbicides : évaluation des impacts agro-environnementaux et propositions de scénarios de gestion » établi par le Cetiom dans le cadre du moratoire sur les variétés génétiquement modifiées de colza. Les auteurs se réservent la possibilité d’ici publication définitive du rapport d’apporter des modifications à ce texte.

Communication du Service juridique et du Département HR à l'attention des membres du personnel domiciliés en France

CERN Multimedia

HR Department

2007-01-01

Déclaration de revenus 2006 Les membres du personnel trouveront ci-après les informations nécessaires sur la manière de remplir la déclaration de revenus 2006, qui doit être retournée au plus tard le 31 mai 2007. Qui doit remplir la déclaration? Comment l'obtenir? Tous les membres du personnel1) domiciliés en France2), qu'ils soient ou non de nationalité française, doivent remplir la déclaration de revenus 2006 sur la base des présentes instructions et la renvoyer signée, au plus tard le 31 mai 2007, au centre des impôts de leur domicile. Les membres du personnel devraient recevoir la déclaration de revenus 2006 fin avril - début mai. Ceux qui ne la recevraient pas directement devront la retirer au centre des impôts, à la trésorerie ou à la mairie de leur domicile ou bien encore la télécharger sur le site internet du Ministère des Finances (www.finances.gouv.fr). Comment remplir la déclaration Les membres du personnel doivent remplir la DECLARATION PAPIER uniquement. Dé...

Forward flux sampling calculation of homogeneous nucleation rates from aqueous NaCl solutions.

Science.gov (United States)

Jiang, Hao; Haji-Akbari, Amir; Debenedetti, Pablo G; Panagiotopoulos, Athanassios Z

2018-01-28

We used molecular dynamics simulations and the path sampling technique known as forward flux sampling to study homogeneous nucleation of NaCl crystals from supersaturated aqueous solutions at 298 K and 1 bar. Nucleation rates were obtained for a range of salt concentrations for the Joung-Cheatham NaCl force field combined with the Extended Simple Point Charge (SPC/E) water model. The calculated nucleation rates are significantly lower than the available experimental measurements. The estimates for the nucleation rates in this work do not rely on classical nucleation theory, but the pathways observed in the simulations suggest that the nucleation process is better described by classical nucleation theory than an alternative interpretation based on Ostwald's step rule, in contrast to some prior simulations of related models. In addition to the size of NaCl nucleus, we find that the crystallinity of a nascent cluster plays an important role in the nucleation process. Nuclei with high crystallinity were found to have higher growth probability and longer lifetimes, possibly because they are less exposed to hydration water.

Analytic flux formulas and tables of shielding functions

International Nuclear Information System (INIS)

Wallace, O.J.

1981-06-01

Hand calculations of radiation flux and dose rates are often useful in evaluating radiation shielding and in determining the scope of a problem. The flux formulas appropriate to such calculations are almost always based on the point kernel and allow for at most the consideration of laminar slab shields. These formulas often require access to tables of values of integral functions for effective use. Flux formulas and function tables appropriate to calculations involving homogeneous source regions with the shapes of lines, disks, slabs, truncated cones, cylinders, and spheres are presented. Slab shields may be included in most of these calculations, and the effect of a cylindrical shield surrounding a cylindrical source may be estimated. Detector points may be located axially, laterally, or interior to a cylindrical source. Line sources may be tilted with respect to a slab shield. All function tables are given for a wide range of arguments

Calculation of fluxes through a repository caused by a local well

International Nuclear Information System (INIS)

Thunvik, R.

1983-05-01

The purpose of the present study is to roughly estimate the ground water flux through a radioactive repository in relation to the flux into a local well under various conditions. The well is assumed to be located at a depth of either 60 or 200 metres below the ground surface. Two main settings are considered, in one the well is located in a vertical fracture zone at a distance of 100 metres from one of the outer edges of the repository. The withdrawal from the well is assumed to be 6 m 3 /day. The flow domain is characterized by a rather low permeability. The boundary conditions considered are either a continuously saturated upper boundary and impervious lateral boundaries, or a phreatic upper boundary and hydrostatic lateral boundaries. The ratio of the flux through the repository to the flux into the well was obtained to be in the range from about 10- 5 to 10- 3 , depending on the boundary conditions and the depth of the well. The lowest figures were obtained in the examples, in which the upper boundary was assumed to be continually saturated. It is concluded that these examples may be considered representative of the actual flow problem. This conclusion is based upon the fact that in the case of a phreatic boundary the drawdown caused by the well was very small and the flow responses were very slow, implying that rather a small infiltration rate is requied to maintain saturated conditions at the upper boundary. The regional gradients caused by the well were rather small in comparison with the typically naturally occurring gradients. The flow to the well will therefore have little influence on the regional flow pattern in most practical cases. (author)

The beam-kicker system of the synchrotron Saturne. Magnetic field and particle orbit computations. Experimental results (1963); Le percuteur de faisceau de Saturne. Calcul du champ magnetique et des trajectoires. Verifications experimentales (1963)

Energy Technology Data Exchange (ETDEWEB)

Gouttefangeas, M; Katz, A; Rastoix, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

In this report is briefly described the beam-kicker system of the synchrotron Saturne. An analysis of its operation based on the sampling method is given, as well as two methods for computing toe magnetic field produced by a set of endless conductors in the neighbourhood of a conducting shield where eddy currents are circulating. The first method leads to the resolution of a bi-dimensional Laplace equation with first kind boundary conditions (Dirichlet problem); the second one translates to electromagnetism the electrical images method currently used in electrostatics and yields the magnetic field as the sum of a triple series expansion in the general case of a set of conductors located in a parallelepipedal box. Finally are given the results obtained in computing on IBM 7090 the perturbation of the particle motion due to the beam-kicker. These results are compared with the experimental data. (authors) [French] Ce rapport decrit brievement le dispositif percuteur de faisceau mis en place sur le synchrotron Saturne. On y trouvera une analyse de se fonctionnement a partir de la theorie des echantillonnages. On indique egalment deux methodes de calcul du champ magnetique produit par un system de conducteurs indefinis en presence d'un blindage conducteur parcouru par des courants de Foucault: la premiere se ramene a la resolution d'une equation de Laplace a deux dimensions avec des conditions aux limites de premiere espece (probleme de Dirichlet), la seconde transpose en electromagnetisme la methode des images electriques classique en electrostatique et permet d'exprimer le champ magnetique sous la forme de la somme d'une serie triple dans le cas general d'un systeme de conducteurs contenus dans un blindage parallelepipedique. Pour terminer, on mentionne les resultats du calcul numerique de la perturbation de la trajectoire des particules sous l'effet du percuteur et on compare ces resultats aux resultats experimentaux. (auteurs)

Tests on dynamic corrosion by water. Influence of the passage of a heat flux on the corrosion kinetics. pH measurement in water at high temperature; Essais de corrosion dynamique par l'eau. Influence du passage d'un flux thermique sur la cinetique de corrosion. Mesure du pH dans l'eau a haute temperature

Energy Technology Data Exchange (ETDEWEB)

Coriou, H.; Grall, L.; Hure, J.; Saint-James, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France); Le peintre [Centre National de la Recherche Scientifique (CNRS), 38 - Grenoble (France)

1958-07-01

The passage of a heat flux through the surface of a metal placed in a corrosive medium influences the rate of corrosion, these being higher than under adiabatic conditions. The apparatus developed for corrosion tests is described, it is possible to obtain with this equipment: 1) Heat fluxes greater than 200 W/cm{sup 2}, across aluminium canning, which is cooled by water (temperature 50 deg. C), circulating with flow rates of the order of 5 to 6 m/s. 2) Heat fluxes which can go up to 150 W/cm{sup 2}, across canning of zircaloy or stainless steel. The cooling fluid is pressurized water at a temperature around 280 deg. C, the flow-rate of circulation reaching 6 m/s. The results obtained on aluminium canning are studied from the viewpoint of corrosion, paying particular attention to cavitation phenomena which can cause serious damage in certain special circumstances. After developing a glass electrode system capable of supporting high pressures, the authors have investigated materials capable of functioning as a hydrogen electrode and of resisting satisfactorily corrosion by water at 200 deg. C. Various possibilities have been examined: electrodes of special glasses, quartz, metals, with a membrane etc... The results of the various tests and the practical limits of utilisation are given. (author)Fren. [French] Le passage d'un flux thermique a travers la surface d'un metal place dans un milieu corrosif influence les vitesses de corrosion, celles-ci etant plus elevees que dans des conditions adiabatiques. On decrit les appareils mis au point, pour essais de corrosion. Ils permettent d'obtenir: 1) A travers des gaine aluminium des flux thermiques depassant 200 W /cm{sup 2}. Les gaines sont refroidies par l'eau (temperature 50 deg. C), circulant a des vitesses de l'ordre de 5 a 6 m/s. 2) A travers des gaines en zircaloy ou acier inoxydable des flux thermiques pouvant s'elever a 150 W/cm{sup 2}. Le fluide de refroidissement est de l

Anatomic-physiological schema of the gastrointestinal tract, to be taken in account in determining the levels of radioactive contamination; Schema anatomo-physiologique du tractus gastro-intestinal a prendre en consideration pour le calcul des niveaux de contamination Radioactive

Energy Technology Data Exchange (ETDEWEB)

Fabry, C [Commissariat a l' Energie Atomique, Cadarache, Association EURATOM-CEA, Niveaux de Contamination (France). Centre d' Etudes Nucleaires

1964-07-01

Anatomical and physiological data of the gastrointestinal tract of adults and children were summarized in a standard schema, to be used in calculating the levels of radioactive contamination, in the food chain. (author) [French] Cette note a pour objet de rassembler les donnees anatomiques et physiologiques du tractus gastrointestinal, chez l'adulte et chez l'enfant, et d'en deduire un schema standard qui puisse servir de base au calcul des niveaux de contamination de la chaine alimentaire par les substances radioactives. (auteur)

Ethanol-metabolizing pathways in deermice. Estimation of flux calculated from isotope effects

International Nuclear Information System (INIS)

Alderman, J.; Takagi, T.; Lieber, C.S.

1987-01-01

The apparent deuterium isotope effects on Vmax/Km (D(V/K] of ethanol oxidation in two deermouse strains (one having and one lacking hepatic alcohol dehydrogenase (ADH] were used to calculate flux through the ADH, microsomal ethanol-oxidizing system (MEOS), and catalase pathways. In vitro, D(V/K) values were 3.22 for ADH, 1.13 for MEOS, and 1.83 for catalase under physiological conditions of pH, temperature, and ionic strength. In vivo, in deermice lacking ADH (ADH-), D(V/K) was 1.20 +/- 0.09 (mean +/- S.E.) at 7.0 +/- 0.5 mM blood ethanol and 1.08 +/- 0.10 at 57.8 +/- 10.2 mM blood ethanol, consistent with ethanol oxidation principally by MEOS. Pretreatment of ADH- animals with the catalase inhibitor 3-amino-1,2,4-triazole did not significantly change D(V/K). ADH+ deermice exhibited D(V/K) values of 1.87 +/- 0.06 (untreated), 1.71 +/- 0.13 (pretreated with 3-amino-1,2,4-triazole), and 1.24 +/- 0.13 (after the ADH inhibitor, 4-methylpyrazole) at 5-7 mM blood ethanol levels. At elevated blood ethanol concentrations (58.1 +/- 2.4 mM), a D(V/K) of 1.37 +/- 0.21 was measured in the ADH+ strain. For measured D(V/K) values to accurately reflect pathway contributions, initial reaction conditions are essential. These were shown to exist by the following criteria: negligible fractional conversion of substrate to product and no measurable back reaction in deermice having a reversible enzyme (ADH). Thus, calculations from D(V/K) indicate that, even when ADH is present, non-ADH pathways (mostly MEOS) participate significantly in ethanol metabolism at all concentrations tested and play a major role at high levels

Improved HOR fuel management by flux measurement data feedback

Energy Technology Data Exchange (ETDEWEB)

Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

1997-07-01

Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

Improved HOR fuel management by flux measurement data feedback

International Nuclear Information System (INIS)

Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M.

1997-01-01

Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author)

Cartes infographiques du Symposium Afrique de l'Ouest et du Centre

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

fdieudonne

VILLES SÛRES ET INCLUSIVES | AFRIQUE DE L'OUEST ET DU CENTRE. Spécificités de violence, d'exclusion, et de lutte contre la criminalité. République démocratique du Congo: Taux de natalité par femme, en moyenne 10 enfants, très élevé. Exclusion sociale exacerbée : des jeunes enfants accusés de sorcellerie.

Solar neutrino flux at keV energies

Science.gov (United States)

Vitagliano, Edoardo; Redondo, Javier; Raffelt, Georg

2017-12-01

We calculate the solar neutrino and antineutrino flux in the keV energy range. The dominant thermal source processes are photo production (γ e→ e νbar nu), bremsstrahlung (e+Ze→ Ze+e+νbar nu), plasmon decay (γ→νbar nu), and νbar nu emission in free-bound and bound-bound transitions of partially ionized elements heavier than hydrogen and helium. These latter processes dominate in the energy range of a few keV and thus carry information about the solar metallicity. To calculate their rate we use libraries of monochromatic photon radiative opacities in analogy to a previous calculation of solar axion emission. Our overall flux spectrum and many details differ significantly from previous works. While this low-energy flux is not measurable with present-day technology, it could become a significant background for future direct searches for keV-mass sterile neutrino dark matter.

Predicting radon flux from uranium mill tailings

International Nuclear Information System (INIS)

Freeman, H.D.; Hartley, J.N.

1983-11-01

Pacific Northwest Laboratory (PNL), under contract to the US Department of Energy (DOE) Uranium Mill Tailings Remedial Action Project (UMTRAP) office, is developing technology for the design of radon barriers for uranium mill tailings piles. To properly design a radon cover for a particular tailings pile, the radon flux emanating from the bare tailings must be known. The tailings characteristics required to calculate the radon flux include radium-226 content, emanating power, bulk density, and radon diffusivity. This paper presents theoretical and practical aspects of estimating the radon flux from an uranium tailings pile. Results of field measurements to verify the calculation methodology are also discussed. 24 references, 4 figures, 4 tables

Power of the uranium plate of Naiade; Puissance de la plaque d'uranium de Naiade

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

1. Calculation. Knowing the incident current of thermal neutrons, the sources of fast neutrons inside the plate are deduced. The fast flux leaving the plate is then determined. Isotropically, 0.57 fast neutrons come out for each incident thermal neutron. 2. Experiments. The incident thermal flux is measured by means of manganese detectors on reaching the plate, and estimated from the flux leaving the graphite. The fast flux coming out is measured with dose-rate films. The absorption of the fast flux in the swimming-pool water is measured. 3. Results. The uranium plate is equivalent to a disc placed at its surface emitting isotropically 3.8 x 10{sup 7} fast neutrons per cm{sup 2}/s, when Zoe operates at 100 kW. This result is valid to about 10 per cent. (author) [French] 1.Calcul. Connaissant le courant de neutrons thermiques incident, on deduit les sources de rapides a l'interieur de la plaque. Puis on determine le flux de rapides sortants de la plaque. Il sort, de facon isotrope, 0,57 neutrons rapides par neutron thermique incident. 2. Experiences. Le flux thermique incident est mesure par des detecteurs au manganese a l'arrivee sur la plaque et estime a partir du flux a la sortie du graphite. Le flux rapide sortant est mesure avec des films de dose. On mesure l'absorption du flux rapide dans l'eau de la piscine. 3. Resultats. La plaque d'uranium est equivalente a un disque place a sa surface emettant isotropiquement, 3,8 x 10{sup 7} neutrons rapides par cm{sup 2}/s, lorsque ZOE marche a 100 kW. Ce resultat est valable a 10 pour cent pres. (auteur)

Protéger les collectivités côtières du nord du Maroc | CRDI - Centre ...

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

3 févr. 2011 ... Élévation du niveau de la mer, ondes de tempête, inondations littorales : les ... Un plan d'action en matière de gestion intégrée du littoral et une nouvelle ... du territoire sans recourir à un guide de la dynamique du milieu côtier. ... local, et encore moins de contrôle sur les terres dont elles sont tributaires.

One dimensional benchmark calculations using diffusion theory

International Nuclear Information System (INIS)

Ustun, G.; Turgut, M.H.

1986-01-01

This is a comparative study by using different one dimensional diffusion codes which are available at our Nuclear Engineering Department. Some modifications have been made in the used codes to fit the problems. One of the codes, DIFFUSE, solves the neutron diffusion equation in slab, cylindrical and spherical geometries by using 'Forward elimination- Backward substitution' technique. DIFFUSE code calculates criticality, critical dimensions and critical material concentrations and adjoint fluxes as well. It is used for the space and energy dependent neutron flux distribution. The whole scattering matrix can be used if desired. Normalisation of the relative flux distributions to the reactor power, plotting of the flux distributions and leakage terms for the other two dimensions have been added. Some modifications also have been made for the code output. Two Benchmark problems have been calculated with the modified version and the results are compared with BBD code which is available at our department and uses same techniques of calculation. Agreements are quite good in results such as k-eff and the flux distributions for the two cases studies. (author)

Apparatus for the measurement of the distribution of neutron flux (band and wire continuous recorders); Appareillage pour la mesure des repartitions de flux neutroniques (derouleurs a bandes et a fils)

Energy Technology Data Exchange (ETDEWEB)

Derrien, H; Lemerle, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

The apparatus of which we describe the principles and the operation, makes it possible to know the thermal neutron flux distribution in a pile channel or along a fuel element by the measurement of the {beta} activity of continuous detectors. It is also possible thereby to show the presence of a very localised sudden discontinuity in the distribution curve. (authors) [French] Les appareils dont nous decrivons le principe et le fonctionnement, permettent de connaitre la repartition du flux de neutrons thermiques dans un canal de pile ou le long d'un element combustible, par la mesure de l'activite {beta} de detecteurs continus. Ils permettent en outre de mettre en evidence un accident brutal, tres localise, de la courbe de repartition. (auteurs)

The sensitivity of the Antares detector to the galactic neutrino flux; Sensibilite du telescope Antares au flux diffus de neutrinos galactiques

Energy Technology Data Exchange (ETDEWEB)

Jouvenot, F

2005-06-15

The Antares european collaboration builds an underwater neutrinos telescope which will be deployed in the Mediterranean by 2500 m depth. This detector consists of a three-dimensional network of 900 photomultipliers which detects the Cherenkov light produced in water by muons created from the interaction of neutrinos in the Earth. Cosmic rays are confined in the Galaxy and interact with the interstellar matter producing charged pions which decay into neutrinos. The observation of the sky with high energy neutrinos (> 100 GeV) could open a new window on the Galaxy, in particular, the detection of these neutrinos may make it possible to directly observe the dense parts of the Galaxy. In this work, corresponding fluxes have been calculated using a simulation program GALPROP, for several models, constrained by various gamma and cosmic rays observations. The expected sensitivity of the Antares detector to these models was reviewed, as well as a first estimation of the performances of what would give a future km{sup 3} scale detector. A shape recognition algorithm was also developed: it would permit to highlight the structures of the Galaxy in the optimistic case which the number of events detected would be sufficient. This work shows that Antares has an insufficient size for observing the galactic plane. It was also demonstrated that a new generation of neutrino telescope having an effective area at least 40 times larger will be needed to detect the hardest spectrum model and put limits on the other models. (author)

Effets du travail du sol sur le comportement chimique et biologique ...

African Journals Online (AJOL)

SARAH

31 juil. 2017 ... RESUME. Objectif : L'objectif de cette étude est de comparer les effets de six techniques culturales de mise en place du blé tendre sur certaines propriétés chimiques et biologiques du sol et les conséquences sur le rendement grain et ses composantes dans la région «non chernozem» en 7ème années ...

Anatomic-physiological schema of the gastrointestinal tract, to be taken in account in determining the levels of radioactive contamination; Schema anatomo-physiologique du tractus gastro-intestinal a prendre en consideration pour le calcul des niveaux de contamination Radioactive

Energy Technology Data Exchange (ETDEWEB)

Fabry, C. [Commissariat a l' Energie Atomique, Cadarache, Association EURATOM-CEA, Niveaux de Contamination (France). Centre d' Etudes Nucleaires

1964-07-01

Anatomical and physiological data of the gastrointestinal tract of adults and children were summarized in a standard schema, to be used in calculating the levels of radioactive contamination, in the food chain. (author) [French] Cette note a pour objet de rassembler les donnees anatomiques et physiologiques du tractus gastrointestinal, chez l'adulte et chez l'enfant, et d'en deduire un schema standard qui puisse servir de base au calcul des niveaux de contamination de la chaine alimentaire par les substances radioactives. (auteur)

Merging a Terrain-Based Parameter and Snow Particle Counter Data for the Assessment of Snow Redistribution in the Col du Lac Blanc Area

Science.gov (United States)

Schön, Peter; Prokop, Alexander; Naaim-Bouvet, Florence; Vionnet, Vincent; Guyomarc'h, Gilbert; Heiser, Micha; Nishimura, Kouichi

2015-04-01

Wind and the associated snow drift are dominating factors determining the snow distribution and accumulation in alpine areas, resulting in a high spatial variability of snow depth that is difficult to evaluate and quantify. The terrain-based parameter Sx characterizes the degree of shelter or exposure of a grid point provided by the upwind terrain, without the computational complexity of numerical wind field models. The parameter has shown to qualitatively predict snow redistribution with good reproduction of spatial patterns. It does not, however, provide a quantitative estimate of changes in snow depths. The objective of our research was to introduce a new parameter to quantify changes in snow depths in our research area, the Col du Lac Blanc in the French Alps. The area is at an elevation of 2700 m and particularly suited for our study due to its consistently bi-modal wind directions. Our work focused on two pronounced, approximately 10 m high terrain breaks, and we worked with 1 m resolution digital snow surface models (DSM). The DSM and measured changes in snow depths were obtained with high-accuracy terrestrial laser scan (TLS) measurements. First we calculated the terrain-based parameter Sx on a digital snow surface model and correlated Sx with measured changes in snow-depths (Δ SH). Results showed that Δ SH can be approximated by Δ SHestimated = α * Sx, where α is a newly introduced parameter. The parameter α has shown to be linked to the amount of snow deposited influenced by blowing snow flux. At the Col du Lac Blanc test side, blowing snow flux is recorded with snow particle counters (SPC). Snow flux is the number of drifting snow particles per time and area. Hence, the SPC provide data about the duration and intensity of drifting snow events, two important factors not accounted for by the terrain parameter Sx. We analyse how the SPC snow flux data can be used to estimate the magnitude of the new variable parameter α . To simulate the development

Improved HOR fuel management by flux measurement data feedback

Energy Technology Data Exchange (ETDEWEB)

Serov, I.V.; Leege, P.F.A. de; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft University of Technology, Reactor Physics Dep., Interfaculty Reactor Inst., Delft (Netherlands)

1997-07-01

Flux distribution in a nuclear reactor can be obtained by utilizing different calculational and experimental methods. The obtained flux distributions are associated with uncertainties and therefore always differ from each other. By combining information from the calculation and experiment using the confluence method, it is possible to obtain a more reliable estimate of the flux distribution than exhibited by the calculation or experiment separately. As a feedback, the fuel burnup distribution, which is used as initial data to the calculation can be improved as well. The confluence method is applied to improvement of the burnup distribution estimates for the HOR research reactor of the Delft University of Technology. An integrated code system CONHOR is developed to match the CITATION results of in-core foil activation rate calculations with in-core experimental data through confluence. The system forms the basis for the advanced fuel management of the reactor. (author) 1 fig., 8 refs.

Theory of flux cutting and flux transport at the critical current of a type-II superconducting cylindrical wire

International Nuclear Information System (INIS)

Clem, John R.

2011-01-01

I introduce a critical-state theory incorporating both flux cutting and flux transport to calculate the magnetic-field and current-density distributions inside a type-II superconducting cylinder at its critical current in a longitudinal applied magnetic field. The theory is an extension of the elliptic critical-state model introduced by Romero-Salazar and Perez-Rodriguez. The vortex dynamics depend in detail on two nonlinear effective resistivities for flux cutting (ρ(parallel)) and flux flow (ρ(perpendicular)), and their ratio r = ρ(parallel)/ρ(perpendicular). When r c (φ) that makes the vortex arc unstable.

Coordonnateur du bureau d'assistance | CRDI - Centre de ...

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

Résumé des fonctions. En tant que membre de l'équipe du Bureau d'assistance au sein de la Section du service à la clientèle (SSC), le coordonnateur du Bureau d'assistance offre au personnel du Centre et aux autres utilisateurs reconnus des installations TI du Centre un soutien de première ligne en matière d'utilisation ...

Reactor calculation benchmark PCA blind test results

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

Reactor calculation benchmark PCA blind test results

Energy Technology Data Exchange (ETDEWEB)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

Prends soin du métro. Le contrôle-soin, forme émergente des régulations et menaces souterraines

Directory of Open Access Journals (Sweden)

Yves Jouffe

2015-05-01

Full Text Available Le métro de Santiago a subi une dégradation significative de ses conditions de transport du fait de son intégration au Transantiago, le réseau de bus urbains. L’opérateur du métro a donné beaucoup de visibilité à ses dispositifs de contrôle des nouvelles masses d’usagers : barrières optimisées, signalétique omniprésente, agents démultipliés, etc. L’étude du discours officiel de l’entreprise et surtout l’observation ethnographique de ces dispositifs révèlent que les modalités de contrôle ont elles-mêmes évoluées. En plus de la menace légale exercée contre les étrangers et de la régulation des flux par une discipline des corps, un contrôle-soin s’exerce sur une population à protéger : la population qui bénéficie du soin de l’opérateur lui est redevable d’un contre-soin. Le contrôle-soin constitue non seulement une modalité émergente de contrôle mais aussi la forme générale du contrôle dans le métro de Santiago.

Calculation of the working capital invested in fuel cycles and its interest charges (1963); Calcul des immobilisations financieres des cycles de combustible (1963)

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

All the processes undergone by the nuclear material, including the various steps of fuel element manufacturing and of irradiated fuel reprocessing lead to working capital investments varying with the type of reactor, that must be taken into account in the kWh cost calculation. The author deals with a calculation method called: 'present worth method' and gives some examples concerning reactors the main fuel of which being either natural uranium or enriched uranium or plutonium. He especially points out the importance these investments may take in the case of fast breeder reactors. (author) [French] L'ensemble des etapes parcourues par la matiere fissile comprenant les divers stades d'elaboration des elements combustibles et de leur traitement apres irradiation, implique des immobilisations financieres tres differentes d'un type de reacteur a l'autre, dont il convient de tenir compte dans le calcul du cout du kWh. L'auteur expose une methode de calcul dite 'd'actualisation des couts' et donne quelques exemples relatifs aux reacteurs utilisant l'uranium naturel, l'uranium enrichi et le plutonium comme combustible principal. Il montre en particulier l'importance que peuvent avoir ces immobilisations dans le cas des reacteurs surregenerateurs. (auteur)

Comparison of calculational methods for liquid metal reactor shields

International Nuclear Information System (INIS)

Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

1985-09-01

A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

Lettre ouverte au Président du Conseil du CERN

CERN Document Server

Association du personnel

2010-01-01

Genève, le 1er février 2010 Monsieur le Président, Les membres de la Caisse de pensions, actifs et pensionnés, sont à la fois très inquiets et indignés par la manière avec laquelle le Conseil traite la question, devenue alarmante, de l’équilibre de la Caisse. Dès 2004 en effet, l’étude actuarielle avait alerté le Conseil au sujet du déficit technique du régime et de l’insuffisance du financement de son plan de prestations. Le Conseil s’est alors contenté de mettre en place un mécanisme de sous indexation des pensions – arbitraire et discriminatoire dans la mesure où il ne fait appel qu’à la solidarité des pensionnés – et de décider une augmentation minime des contributions à la Caisse, une mesurette ...

Heat exchanges during the re-flooding of a water reactor core - within the framework of the 'reference accident'; Echanges thermiques lors du renoyage d'un coeur de reacteur a eau - dans le cadre de 'l'accident de reference'

Energy Technology Data Exchange (ETDEWEB)

Andreoni, Daniel

1975-11-28

After a brief presentation of reported studies made in different countries and regarding the so-called 'reference accident', this research thesis reports the study of reactor re-flooding when the reactor is completely dried and heating elements have reached a temperature between 300 and 900 C, with a constant water flow rate entering the test section, with a constant dissipated electrical power, and by using very simple geometries. After a first part addressing the experimental study, the author reports the development of conduction calculation codes used to compute the flow extracted by the two-phase flow, present the thermal-hydraulic code used to compute local values and to study the correlation of the upstream area exchange coefficient. The author finally reports an analysis of the different existing models and the study of a re-flooding model [French] La presente etude est consacree a l'un des aspects de la surete des reacteurs a eau sous pression, et plus precisement a l'accident tres important qui consiste en une perte de fluide caloporteur (Loss of Coolant Accident - 'LOCA'). Le but de l'etude est de fournir des renseignements necessaires a l'interpretation des experiences effectuees sur des grappes, de donner une correlation de coefficient d'echange dans la zone aval, et de donner aussi un modele de progression du front de trempe pour les analyses de surete. Une etude bibliographique preliminaire nous a permis de faire le point sur les experiences entreprises concernant le refroidissement de secours. Ensuite, les chapitres suivants seront decrits: 1) Le chapitre II, consacre a l'etude experimentale (boucle, sections d'essais, resultats globaux). 2) Le chapitre III ou seront presentes les codes de calcul de conduction, necessaires au calcul du flux extrait par le melange diphasique, le code de thermohydraulique necessaire au calcul des grandeurs locales et l'etude de la correlation du coefficient d'echange de la zone aval. 3) Enfin le chapitre IV ou, apres

A method to calculate flux distribution in reactor systems containing materials with grain structure

International Nuclear Information System (INIS)

Stepanek, J.

1980-01-01

A method is proposed to compute the neutron flux spatial distribution in slab, spherical or cylindrical systems containing zones with close grain structure of material. Several different types of equally distributed particles embedded in the matrix material are allowed in one or more zones. The multi-energy group structure of the flux is considered. The collision probability method is used to compute the fluxes in the grains and in an ''effective'' part of the matrix material. Then the overall structure of the flux distribution in the zones with homogenized materials is determined using the DPN ''surface flux'' method. Both computations are connected using the balance equation during the outer iterations. The proposed method is written in the code SURCU-DH. Two testcases are computed and discussed. One testcase is the computation of the eigenvalue in simplified slab geometry of an LWR container of one zone with boral grains equally distributed in an aluminium matrix. The second is the computation of the eigenvalue in spherical geometry of the HTR pebble-bed cell with spherical particles embedded in a graphite matrix. The results are compared to those obtained by repeated use of the WIMS Code. (author)

Seasonal and vertical variations of sinking particle fluxes in the West Caroline Basin

OpenAIRE

Kawahata, H; Yamamuro, M; Ohta, H

1998-01-01

Des pièges à sédiments ont été déployés dans le bassin des Carolines Occidentales (ouest du Pacifique équatorial), entre la zone d'influence de la mousson d'Asie et l'océan ouvert. À la station 1, le flux annuel au piège le moins profond est 57, 10 g m-2 an-1. Les flux élevés de matière organique sont généralement associés au développement des communautés planctoniques à test silicieux et carbonaté. De plus, le rapport carbone organique/carbone minéral tend à augmenter avec la teneur en matiè...

Ved du, hvad der er i dine varer, når du handler?

DEFF Research Database (Denmark)

Phillip, Anja; Smith, Viktor

2016-01-01

I supermarkedet på vej hjem køber du ind til madpakken. Ved kølemontren med pålæg ser du en pakke Gårdlykke Hønsesalat. ... Måske vælger du den uden at tænke nærmere over hvorfor, eller fordi emballagen og navnet får hønsesalaten til at fremstå lokal, hjemmelavet eller økologisk. ... Alt dette ba...... bakkes op af resultaterne fra årelang forskning på CBS, hvor universitetets forskergruppe Fair-Speak har undersøgt emballagernes måde at påvirke forbrugerne til at købe produktet....

Étude de la qualité des eaux usées des deux retenues du bassin du ...

African Journals Online (AJOL)

SARAH

31 mai 2017 ... Gourou au carrefour de l'Indénié à Abidjan en 2012. 11138. Journal of Applied .... retenues sont nées du projet de gestion intégrée du bassin du Gourou (CI-FAD, ..... Contrôle de la qualité microbiologique des eaux usées ...

fibrosarcome du larynx

African Journals Online (AJOL)

pie du lit tumoral est employée comme complément thé- rapeutique [9] alors que la chimiothérapie est générale- ment indiquée dans les formes métastatiques. Le pronos- tic dépend essentiellement du degré de différentiation his- tologique. En fait, le fibrosarcome bien différencié est caractérisé par la fréquence de récidive ...

Au fil du temps (1976 ou la loi du seuil

Directory of Open Access Journals (Sweden)

Isabelle Singer

2010-04-01

Full Text Available Ayant choisi de vivre dans un camion, les héros d’Au fil du temps (film de Wim Wenders de 1976 font du seuil une expérience particulière. Celle de la rupture avec toute idée de foyer et celle du refus d’appartenance à la terre natale, cette terre allemande traversée par une frontière le long de laquelle ils vont voyager, et qui les renvoie à chaque instant aux traumatismes de l’Histoire. Sur le pare-brise du camion, l’extérieur (campagnes indifférenciées, villes à l’abandon… et l’intérieur se superposent. Au fil du temps questionne le paysage : il s’agit d’en décoller un à un les mythes qui le recouvrent. L’image alors n’est plus surface mais volume à traverser, à lacérer et découvrir ce qui est tissé dans le paysage. La démarche de Wenders est alors proche de celle d’un de ses contemporains : Anselm Kiefer. Le paysage allemand provoque le rejet parce qu’il y a là toujours plus que le visible : des strates et des strates de culpabilité que le mythe - et c’est sa fonction - a recouvert. Et qu’il s’agira ici, de soulever. Le choix du nomadisme, c’est celui d’un état de l’humanité antérieur à l’idée de patrie. Et c’est aussi celui de la solitude, comme prix à payer à ce refus d’appartenance et à cette mise à jour des mythes. L’appartenance à la terre allemande et à son Histoire est profondément problématique parce que les pères sont fondamentalement coupables. On se reconnaîtra alors des pères de substitution : des pères de cinéma (Nicholas Ray ou Fritz Lang. Et l’on substituera l’Histoire du cinéma à l’Histoire. Bruno est réparateur ambulant de projecteurs et Au fil du temps dressera, au gré de ses pérégrinations, un état des lieux du cinéma allemand des années soixante-dix : déliquescent, colonisé par les images hollywoodiennes. Il faut que cela change : état du cinéma ; état des protagonistes solitaires en quête d’une identit

Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

Directory of Open Access Journals (Sweden)

Yang Seong Woo

2016-01-01

Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

Hausse du niveau des océans et perte de terres dans le delta du Nil ...

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

9 juin 2016 ... Au moyen de systèmes d'information géographique (SIG), des chercheurs subventionnés par le CRDI ont déterminé les zones littorales du delta du Nil qui pourraient être touchées par la hausse du niveau des océans. Il pourrait en résulter d'importantes pertes de terres d'ici 2100. Selon les plus récentes ...

Les mots du jazz

OpenAIRE

Roueff, Olivier

2007-01-01

L’ouvrage d’André Schaeffner constitue la première analyse savante du jazz (1926). Il a marqué une étape importante dans le processus de réinvention du jazz en France en contribuant notamment, par sa réception et les polémiques qu’il a suscitées, à transformer l’identification du jazz d’une musique « américaine » à une musique « noire-américaine » (c’est-à-dire aux « racines » africaines). Les analyses proposées dans cet ouvrage, alors qu’elles désignaient des musiques que la critique de jazz...

23 octobre 2012 - Le Président du Conseil général de la Haute-Savoie C. Monteil signe le livre d'or en présence du Directeur général R. Heuer, la chef des Relations Internationales F. Pauss, le chef du département Physique P. Bloch et le chef du département Technologie F. Bordry; visite du hall de test des aimants supraconducteurs du LHC avec F. Bordry.

CERN Multimedia

Maximilien Brice

2012-01-01

23 octobre 2012 - Le Président du Conseil général de la Haute-Savoie C. Monteil signe le livre d'or en présence du Directeur général R. Heuer, la chef des Relations Internationales F. Pauss, le chef du département Physique P. Bloch et le chef du département Technologie F. Bordry; visite du hall de test des aimants supraconducteurs du LHC avec F. Bordry.

Calcul statistique du volume des blocs matriciels d'un gisement fissuré The Statistical Computing of Matrix Block Volume in a Fissured Reservoir

Directory of Open Access Journals (Sweden)

Guez F.

2006-11-01

Full Text Available La recherche des conditions optimales d'exploitation d'un gisement fissuré repose sur une bonne description de la fissuration. En conséquence il est nécessaire de définir les dimensions et volumes des blocs matriciels en chaque point d'une structure. Or la géométrie du milieu (juxtaposition et formes des blocs est généralement trop complexe pour se prêter au calcul. Aussi, dans une précédente communication, avons-nous dû tourner cette difficulté par un raisonnement sur des moyennes (pendages, azimuts, espacement des fissures qui nous a conduits à un ordre de grandeur des volumes. Cependant un volume moyen ne peut pas rendre compte d'une loi de répartition des volumes des blocs. Or c'est cette répartition qui conditionne le choix d'une ou plusieurs méthodes successives de récupération. Aussi présentons-nous ici une méthode originale de calcul statistique de la loi de distribution des volumes des blocs matriciels, applicable en tout point d'un gisement. La part de gisement concernée par les blocs de volume donné en est déduite. La connaissance générale du phénomène de la fracturation sert de base au modèle. Les observations de subsurface sur la fracturation du gisement en fournissent les données (histogramme d'orientation et d'espacement des fissures.Une application au gisement d'Eschau (Alsace, France est rapportée ici pour illustrer la méthode. The search for optimum production conditions for a fissured reservoir depends on having a good description of the fissure pattern. Hence the sizes and volumes of the matrix blocks must be defined at all points in a structure. However, the geometry of the medium (juxtaposition and shapes of blocks in usually too complex for such computation. This is why, in a previous paper, we got around this problem by reasoning on the bases of averages (clips, azimuths, fissure spacing, and thot led us to an order of magnitude of the volumes. Yet a mean volume cannot be used to explain

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1976-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount. For example, a deuterium plasma with 1.3 percent oxygen, one of the particle transport coefficients is increased by a factor of about four. The transport coefficients for the toroidal magnetic flux are reduced by about 20 percent. The increase in the particle transport coefficient is due to the collisional scattering of the deuterons by the heavy oxygen ions which is larger than the deuteron electron scattering, the normal process for particle transport in a two species plasma. The reduction in the toroidal magnetic flux transport coefficients are left unexplained

Flux-limited diffusion models in radiation hydrodynamics

International Nuclear Information System (INIS)

Pomraning, G.C.; Szilard, R.H.

1993-01-01

The authors discuss certain flux-limited diffusion theories which approximately describe radiative transfer in the presence of steep spatial gradients. A new formulation is presented which generalizes a flux-limited description currently in widespread use for large radiation hydrodynamic calculations. This new formation allows more than one Case discrete mode to be described by a flux-limited diffusion equation. Such behavior is not extant in existing formulations. Numerical results predicted by these flux-limited diffusion models are presented for radiation penetration into an initially cold halfspace. 37 refs., 5 figs

Control of pile power by measurement of the activity of the cooling fluid. The particular case of EL3; Controle de la puissance d'une pile par la mesure de l'activite du fluide de refroidissement. Cas particulier de EL 3

Energy Technology Data Exchange (ETDEWEB)

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

1959-07-01

In a high flux pile the presence of local absorbers makes the power difficult to measure. In any case thermodynamic measurements are sometimes not very accurate. Another possible method could be the measurement of the cooling fluid activity. This has been tried on the heavy water circuit of EL3. In the first part of this report we give some qualitative indications of the various activities present in the heavy water in circulation. After this, the activity of the element chosen has been calculated. Finally, the results obtained from EL3 are given. (author) [French] Dans une pile a haut flux, la presence d'absorbeurs localises rend la mesure de la puissance difficile. Les mesures thermodynamiques sont par ailleurs parfois peu precises. Un autre moyen pourrait etre la mesure de l'activite du fluide de refroidissement. C'est ce qui a ete essaye sur le circuit eau lourde de EL3. Dans une premiere partie, nous donnons quelques indications qualitatives sur les diverses activites presentes dans l'eau lourde en circulation. Ensuite, nous avons fait le calcul de l'activite de l'element choisi. Enfin, nous avons mentionne les resultats qui ont ete obtenus EL3. (auteur)

Les outils du CERN

CERN Multimedia

1999-01-01

C'est le plus grand centre mondial de recherche en physique des particules. Les outils du Laboratoire, accélérateurs et détecteurs de particules, figurent parmi les instruments scientifiques les plus complexes au monde. Des prix Nobels ont d'ailleurs été attribués aux physiciens du CERN pour leurs développements.

Évaluation du nouveau programme d'études du College of Health ...

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

Ces changements étaient jugés nécessaires pour préparer les professionnels de la santé aux nouvelles exigences de la prestation de services de santé décentralisés, pour juguler les nouvelles pandémies comme celles du VIH/sida et du virus Ébola, ainsi que pour répondre aux besoins des malades dans les zones ...

On the kinetics of the aluminum-water reaction during exposure in high-heat flux test loops: 1, A computer program for oxidation calculations

International Nuclear Information System (INIS)

Pawel, R.E.

1988-01-01

The ''Griess Correlation,'' in which the thickness of the corrosion product on aluminum alloy surfaces is expressed as a function of time and temperature for high-flux-reactor conditions, was rewritten in the form of a simple, general rate equation. Based on this equation, a computer program that calculates oxide-layer thickness for any given time-temperature transient was written. 4 refs

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

L’écriture du temps dans Robinson Crusoe

OpenAIRE

Bulckaen, Denise

2018-01-01

“Le récit est une séquence deux fois temporelle... il y a le temps de la chose-racontée et le temps du récit (temps du signifié et temps du signifiant)”. Cette citation se trouve au début du chapitre que Genette consacre à “Ordre” dans Figures III. Elle s'applique à de nombreux romans, mais elle est particulièrement intéressante quand on considère le roman de forme autobiographique. Le décalage entre “le temps du signifié” et “le temps du signifiant” est plus ou moins grand selon les romans. ...

Atmospheric neutrino fluxes

International Nuclear Information System (INIS)

Perkins, D.H.

1984-01-01

The atmospheric neutrino fluxes, which are responsible for the main background in proton decay experiments, have been calculated by two independent methods. There are discrepancies between the two sets of results regarding latitude effects and up-down asymmetries, especially for neutrino energies Esub(ν) < 1 GeV. (author)

Établissements recevant du public

CERN Document Server

2014-01-01

Synthèse pour les installations électriques, des règles de sécurité contre les risques d’incendie et de panique dans les établissements recevant du public (ERP), selon l’arrêté du 25 juin 1980 modifié. L'ouvrage traite à la fois des dispositions communes aux établissements du 1er groupe (1ère à 4ème catégorie), du 2ème groupe (5ème catégorie) et des règles particulières applicables aux différents types d'établissements (structures d'accueil pour personnes âgées ou handicapées, hôtels, magasins de vente, restaurants et débit de boisson, établissements de soins, administration, châpiteaux, etc...) Sont ainsi notamment traités, dans le cadre des dispositions communes aux établissements du 1er groupe, pour les installations électriques normales : les installations des appareils, tableaux et canalisations, les locaux électriques et installation de machines, l'éclairage normal des locaux accessibles au public, le chauffage et la ventilation, l'installation d'eau chaude sanita...

Thermal and hydrodynamic study of a whirling liquid hydrogen layer under high heat flux; Etude thermique et hydrodynamique d'une couche tourbillonnaire d'hydrogenen liquide sous flux de chaleur eleve

Energy Technology Data Exchange (ETDEWEB)

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

1968-03-01

In order to achieve a cold neutrons source ({lambda} {>=} 4.10{sup -10} m) in a high flux reactor ({approx} 10{sup 15} neutrons/cm{sup 2}.s), a whirling liquid hydrogen layer (145 mm OD, effective thickness 15 mm, height about 180 mm) was formed, out-of-pile, in a cylindrical transparent glass vessel. The whirling motion was obtained by tangential injection of the liquid, near the wall. Thermal and hydrodynamical conditions of formation and laws of similarity of such a layer were studied. The characteristics of this whirling flow were observed as a function of mass flow rate (5 to 27 g/s; 4.3 to 23 l/mn), and of spillway width (18 and 25 mm). Six different nozzles were used : 1.0; 1.5; 1.9; 2.25; 2.65 and 3.0 mm ID. The total heat influx was found between 8.6 and 10.4 kW. The heat flux density was about 9.4 W/cm{sup 2} and the mean layer density around 80 per cent of that of the liquid hydrogen at 20.4 Kelvin. High speed movies were used to analyze the boiling regime. (author) [French] En vue de realiser une source de neutrons froids ({lambda} {>=} 4.10{sup -10} m) dans un reacteur a haut flux ({approx} 10{sup 15} neutrons thermiques/cm{sup 2}.s), on a forme dans un vase cylindrique transparent en verre, hors-pile, une couche tourbillonnaire ('vortex') d'hydrogene liquide (diametre exterieur 145 mm, epaisseur effective 15 mm, hauteur 180 mm environ). Le mouvement giratoire est obtenu par injection tangentielle du liquide pres de la paroi. L'etude porte sur la determination des conditions thermiques et hydrodynamiques de la formation d'une telle couche et sur les regles de similitude de ce phenomene. On a observe les caracteristiques de l'ecoulement giratoire en fonction du debit (de 5 a 27 g/s, soit de 4.3 a 23 1/mn), de la vitesse d'injection (entre 10 et 110 m/s) et de la largeur du deversoir (18 et 25 mm), ceci pour six diametres differents d'injecteur (1.0 ; 1.5; 1.9; 2.25; 2.65 et 3.0 mm). Le flux de chaleur total mesure

Analysis of the two-fluid model and the drift-flux model for numerical calculation of two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Munkejord, Svend Tollak

2006-05-11

This thesis analyses models for two-phase flows and methods for the numerical resolution of these models. It is therefore one contribution to the development of reliable design tools for multiphase applications. Such tools are needed and expected by engineers in a range of fields, including in the oil and gas industry. The approximate Riemann solver of Roe has been studied. Roe schemes for three different two-phase flow models have been implemented in the framework of a standard numerical algorithm for the solution of hyperbolic conservation laws. The schemes have been analysed by calculation of benchmark tests from the literature, and by comparison with each other. A Roe scheme for the four-equation one-pressure two-fluid model has been implemented, and a second-order extension based on wave decomposition and flux-difference splitting was shown to work well and to give improved results compared to the first-order scheme. The convergence properties of the scheme were tested on smooth and discontinuous solutions. A Roe scheme has been proposed for a five-equation two-pressure two-fluid model with pressure relaxation. The use of analogous numerical methods for the five-equation and four-equation models allowed for a direct comparison of a method with and without pressure relaxation. Numerical experiments demonstrated that the two approaches converged to the same results, but that the five-equation pressure-relaxation method was significantly more dissipative, particularly for contact discontinuities. Furthermore, even though the five-equation model with instantaneous pressure relaxation has real eigenvalues, the calculations showed that it produced oscillations for cases where the four-equation model had complex eigenvalues. A Roe scheme has been constructed for the drift-flux model with general closure laws. For the case of the Zuber-Findlay slip law describing bubbly flows, the Roe matrix is completely analytical. Hence the present Roe scheme is more efficient than

du Chott Marouane

African Journals Online (AJOL)

plancton de 90 µm de vide de maille. Ils ont été conservés dans du formol à 5%. L'identification de l'espèce est basée sur des critères morphologiques [20]: la forme des furcas, les lobes frontaux des antennes des mâles, de l'organe copulateur (pénis) et du sac ovigère. Le comptage des soies furcales a été réalisé. L'étude ...

CONDOR: neutronic code for fuel elements calculation with rods

International Nuclear Information System (INIS)

Villarino, E.A.

1990-01-01

CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author) [es

Modélisation du champ magnétique d'un propulseur M.H.D. annulaire

OpenAIRE

Kom , C.; Brunet , Y.

1995-01-01

Pour des raisons de discrétion, les champs de fuite doivent être minimisés en propulsion M.H.D. où les champs magnétiques doivent être intenses. Le calcul du champ magnétique d'un propulseur M.H.D. naval annulaire, constitué de secteurs inducteurs supraconducteurs est représenté. Dans la zone active, hors des têtes de bobines, une formulation analytique peut être utilisée. Une méthode analytique utilisant le développemment en série de Fourier du courant est adoptée pour les industeurs cylindr...

A Mathematical model for gastrointestinal tract dosimetry, and its application to adults; Modele mathematique destine a la dosimetrie du tractus gastrointestinal - application a l'adulte

Energy Technology Data Exchange (ETDEWEB)

Vacca, G; Bazin, J P; Karhausen, L [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1969-07-01

A new method is proposed for the computation of the dose delivered to any level of the gastrointestinal tract after ingestion of radioactive substances. The same method is used as that proposed by ICRP. The concentration of the radionuclide during the progression of the meal is the central concept of this model. A graph relates the MPC with the radioactive half life of the radionuclide and with any level of the gastrointestinal tract of adults.Soluble and insoluble radionuclides are considered. The case of daughter products is discussed in the Appendix. (author) [French] Une methode nouvelle permet le calcul de la dose delivree a un niveau quelconque du tractus gastrointestinal apres ingestion de substances radioactives. Elle utilise d une part la technique de calcul proposee par l'ICRP et de l'autre la notion de concentration du radionuclide au cours de la progression du repas. Un graphique simple met en relation la valeur de la concentration maximale admissible avec la periode radioactive du radionuclide et le niveau considere du tractus gastrointestinal de l'adulte. Le cas d'un radionuclide soluble et celui d'un radionuclide insoluble sont envisages. Un appendice traite le cas des filiations. (auteur)

OGM : vers une définition commune du contenu du problème ?

Directory of Open Access Journals (Sweden)

Icart Jean-Claude

2001-07-01

Full Text Available Les divergences croissantes depuis des années entre experts, responsables économiques, et acteurs sociaux et « citoyens » au sujet des OGM, ont pu être analysées à juste titre, comme exprimant la difficulté de dégager une « définition commune du contenu du problème »1.

L’Association du personnel (AP) en réunion du Directorat élargi (ED) !

CERN Multimedia

Staff Association

2017-01-01

Le 3 avril dernier, la Vice-Présidente et le Président de l’Association du personnel ont présenté en réunion du Directorat élargi (Directeurs et Chefs de départements et d’unités) le plan des activités de l’Association du personnel pour 2017 et ont fait part des préoccupations de l’AP. Cinq sujets ont été abordés en commençant par la mise en œuvre des décisions prises dans le cadre de l’examen quinquennal de 2015. Examen quinquennal – suivi (voir Echo n° 257) 2016 – Principales mises en œuvre De nombreux changements ont déjà été mis en place en 2016 : Révision des Statut et Règlement du personnel en janvier 2016, pour les aspects de diversité, et en septembre 2016, pour la ...

Atmosphere–Surface Fluxes of CO2 using Spectral Techniques

DEFF Research Database (Denmark)

Sørensen, Lise Lotte; Larsen, Søren Ejling

2010-01-01

Different flux estimation techniques are compared here in order to evaluate air–sea exchange measurement methods used on moving platforms. Techniques using power spectra and cospectra to estimate fluxes are presented and applied to measurements of wind speed and sensible heat, latent heat and CO2...... fluxes. Momentum and scalar fluxes are calculated from the dissipation technique utilizing the inertial subrange of the power spectra and from estimation of the cospectral amplitude, and both flux estimates are compared to covariance derived fluxes. It is shown how even data having a poor signal......-to-noise ratio can be used for flux estimations....

Les Cahiers du CREAD

African Journals Online (AJOL)

Admin

6 juil. 2007 ... La problématique du développement du secteur de l'artisanat en. Algérie a été très peu abordée par les chercheurs, qu'ils soient universitaires ou .... La loi a institué une taxe d'apprentissage dont le taux a été fixé à. 1% de la ...

Consolidation du leadership en recherche en écosanté en Asie du ...

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

1 juin 2016 ... Financée par le CRDI, l'Initiative de renforcement du leadership en recherche en écosanté en Asie du Sud‑Est appuie l'exécution, dans la région, de travaux de recherche qui visent à permettre de mieux comprendre les effets de la transformation de l'agriculture sur les écosystèmes et sur la santé humaine ...

Flow-excursion-induced dryout at low-heat-flux

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.G.

1983-01-01

Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures

Le Silurien du Synclinorium de Moncorvo (NE du Portugal): Biostratigraphie et Importance Paléogéographique

Czech Academy of Sciences Publication Activity Database

Sarmiento, G. N.; Picarra, J. M.; Rebelo, J. A.; Robardet, M.; Gutiérrez-Marco, J. C.; Štorch, Petr; Rábano, I.

1999-01-01

Roč. 32, č. 5 (1999), s. 749-767 ISSN 0016-6995 R&D Projects: GA AV ČR Projects 351 et 421 du Programme International de Corrélation Géologique et au Projet Iberian Variscides du programme Europrobe, Projet 061-B0 du Programme de Coopération Scientifique et Technique Franco-Portugais Subject RIV: DB - Geology ; Mineralogy Impact factor: 0.736, year: 1999

Notes on neutron flux measurement

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs

Tabular method of critical heat flux description in square packing rod bundles

International Nuclear Information System (INIS)

Bobkov, V.P.; Smogalev, I.P.

2003-01-01

Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

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)

Le Flaubert de Charles Du Bos

Directory of Open Access Journals (Sweden)

Jacques Neefs

2009-01-01

Full Text Available Charles Du Bos a porté une attention constante à l’œuvre de Flaubert (à l’exclusion de Bouvard et Pécuchet qui semble ne pas exister pour lui, à Madame Bovary et à L’Éducation sentimentale en particulier. La mise en relation de son étude : « Sur le milieu intérieur chez Flaubert », écrite en 1921, avec des textes du Journal de 1923 et de 1937, les rapprochements avec Gogol, Thomas Hardy, Tolstoï, Baudelaire, Henry James qui traversent les écrits de Du Bos, permettent de suivre ce que celui-ci décrit comme « l’expérience spirituelle » d’une matérialité comprise dans la conquête de la triple exigence du Beau, du Vivant et du Vrai. Du Bos décèle la force de l’œuvre de Flaubert dans la « disproportion » du style, et dans la puissance d’absorption qui fait la densité de cette prose, et qui désigne un extraordinaire travail de conversion. L’obscure expérience spirituelle ainsi poursuivie est celle d’un absolu de l’art, expérience paradoxale d’un « mystique qui ne croit à rien » (comme se désignait Flaubert lui-même, que le critique lie à une interrogation sur sa propre conversion.Charles Du Bos devoted an unflagging attention to Flaubert’s work (except for Bouvard et Pécuchet, which, apparently, according to him did not exist, to Madame Bovary and in particular L’Éducation sentimentale. The connection between his essay “Sur le milieu intérieur chez Flaubert”, written in 1921, and extracts from his Journal, from 1923 to 1937, the comparisons with Gogol, Thomas Hardy, Tolstoy, Baudelaire, and Henry James that run through the writings of Du Bos, allow us to follow what he terms “the spiritual experience” of a materiality encompassed in the conquest of the triple demand of the Beautiful, the Living, the Truth. Du Bos detects the power of Flaubert’s work in the “disproportion” of his style, and the power of absorption that forms the density of his prose, showing an

Unconventional application of the two-flux approximation for the calculation of the Ambartsumyan-Chandrasekhar function and the angular spectrum of the backward-scattered radiation for a semi-infinite isotropically scattering medium

Science.gov (United States)

Remizovich, V. S.

2010-06-01

It is commonly accepted that the Schwarzschild-Schuster two-flux approximation (1905, 1914) can be employed only for the calculation of the energy characteristics of the radiation field (energy density and energy flux density) and cannot be used to characterize the angular distribution of radiation field. However, such an inference is not valid. In several cases, one can calculate the radiation intensity inside matter and the reflected radiation with the aid of this simplest approximation in the transport theory. In this work, we use the results of the simplest one-parameter variant of the two-flux approximation to calculate the angular distribution (reflection function) of the radiation reflected by a semi-infinite isotropically scattering dissipative medium when a relatively broad beam is incident on the medium at an arbitrary angle relative to the surface. We do not employ the invariance principle and demonstrate that the reflection function exhibits the multiplicative property. It can be represented as a product of three functions: the reflection function corresponding to the single scattering and two identical h functions, which have the same physical meaning as the Ambartsumyan-Chandrasekhar function ( H) has. This circumstance allows a relatively easy derivation of simple analytical expressions for the H function, total reflectance, and reflection function. We can easily determine the relative contribution of the true single scattering in the photon backscattering at an arbitrary probability of photon survival Λ. We compare all of the parameters of the backscattered radiation with the data resulting from the calculations using the exact theory of Ambartsumyan, Chandrasekhar, et al., which was developed decades after the two-flux approximation. Thus, we avoid the application of fine mathematical methods (the Wiener-Hopf method, the Case method of singular functions, etc.) and obtain simple analytical expressions for the parameters of the scattered radiation

Statistical mechanics of flux lines in high-temperature superconductors

International Nuclear Information System (INIS)

Dasgupta, C.

1992-01-01

The shortness of the low temperature coherence lengths of high T c materials leads to new mechanisms of pinning of flux lines. Lattice periodic modulations of the order parameters itself acts to pin vortex lines in regions of the unit cell were the order parameter is small. A presentation of flux creep and flux noise at low temperature and magnetic fields in terms of motion of simple metastable defects on flux lines is made, with a calculation of flux lattice melting. 12 refs

Droit du dommage corporel systèmes d'indemnisation

CERN Document Server

Lambert-Faivre, Yvonne

2015-01-01

Le Droit du dommage corporel est aujourd'hui une spécialité reconnue, à la convergence du droit civil et du droit pénal, du droit de la consommation et du droit de l'environnement, du droit de la sécurité sociale et du droit des transports, etc. La sécurité et l'intégrité de la personne humaine sont des droits fondamentaux où l'exigence éthique et l'équité confortent la règle juridique pour tenter d'indemniser les victimes de dommages corporels avec efficacité et transparence. Dans tous ces domaines, cette 8e édition rend compte de l'actualité du droit du dommage corporel : jurisprudence détaillée sur la nomenclature des préjudices réparables, à l'heure où sa consécration par décret reste en suspens, évaluation médicale et monétaire des préjudices, actualité jurisprudentielle du recours des tiers payeurs (article 25 de la loi du 21 décembre 2006 et ses difficultés d'application), sécurité sociale, responsabilité médicale et indemnisations des victimes d'accidents médica...

De Paris à Lyon. Les mutations éditoriales du «Lancelot du Lac»

Directory of Open Access Journals (Sweden)

Gaëlle Burg

2015-07-01

Full Text Available Lancelot du Lac est le premier roman arthurien imprimé à la Renais­sance. Dans sa première édition parue en 1488, qui réunit deux imprimeurs (Jean Le Bourgeois à Rouen et Jean Du Pré à Paris, un découpage et un prologue inédits sont ajoutés par le remanieur. Le texte sera réédité six fois par divers imprimeurs-libraires parisiens jusqu’en 1533. L’édition de luxe d’Antoine Vérard (1494, destinée au roi Charles VIII, présente d’importantes modifications effectuées dans un but commercial. Après une longue période d’accalmie qui signe le début du déclin de la vogue des romans de chevalerie médiévaux, Benoît Rigaud publie à Lyon, sous une forme considérablement abrégée, la dernière édition connue du Lancelot au XVIe siècle (1591. Si elle ne présente que peu d’intérêt littéraire, elle apporte cependant des informations concernant les pratiques éditoriales et les goûts du  lecteur de la fin du XVIe siècle. De Paris à Lyon, entre renaissance et déclin, le parcours éditorial d’un incontournable roman arthurien.Lancelot du Lac is the editio princeps of an Arthurian romance in Renaissance France. The first edition in 1488, which brings together two printers (Jean Le Bourgeois from Rouen and Jean Du Pré from Paris, offers original arrangement and prologue added by the compositor. The text will be published six times by various printers and booksellers in Paris until 1533. The luxurious edition from Antoine Vérard (1494 dedicated to King Charles VIII provides interesting transformations in commercial purposes. After a long time without edition, showing the beginning of chivalry literature’s decline, Benoît Rigaud publish in Lyon, in a greatly abbreviated form, the last known edition of Lancelot in the XVIth century (1591. If it presents no literary interest, it provides nevertheless informations about editorial practices and reader’s tastes from the end of Renaissance France. From Paris to

Spectrophotometric study of the complexes of cerium and uranides with diethylenetetraaminepentaacetic acid (DTPA); Contribution a l'etude de la complexation du cerium et des uranides par l'acide diethylenetetraminepentaacetique (DTPA)

Energy Technology Data Exchange (ETDEWEB)

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

1968-03-01

A spectrophotometric determination was made of the complexes of cerium and uranides in their different degrees of oxidation with diethylenetriaminepentaacetic acid (DTPA). Two physico-chemical states, the ionic and hydrolysis ones, were considered. In the former case, we studied the influence of the pH on the formation of the complex, and whenever the complexes were stable, we determined the molecular ratio and calculated their stability constants. In the latter case, we studied the conditions of solubilization of the precipitate by formation of complexes soluble in water as a function of pH, age of the precipitate and time of precipitate-chelate contact. (authors) [French] Etude spectrophotometrique de la complexation du cerium et des uranides (uranium, neptunium, plutonium et americium) dans leurs differents etats d'oxydation par l'acide diethylenetriaminepentaacetique (DTPA). Deux etats physicochimiques ont retenu notre attention, l'etat ionique et l'etat precipite d'hydrolyse. Dans le premier cas, nous avons etudie l'influence du pH sur la formation du complexe, et dans tous les cas ou le complexe est stable, nous avons determine le rapport moleculaire (5) [element]/[chelatant] du complexe forme et nous avons calcule leurs constantes de stabilite. Dans le deuxieme cas, l'etat de precipite d'hydrolyse, nous avons etudie les conditions de solubilisation du precipite, par formation du complexe, en fonction du pH, de l'age du precipite et du temps de contact precipite - chelatant. (auteurs)

Cool seafloor hydrothermal springs reveal global geochemical fluxes

Science.gov (United States)

Wheat, C. Geoffrey; Fisher, Andrew T.; McManus, James; Hulme, Samuel M.; Orcutt, Beth N.

2017-10-01

We present geochemical data from the first samples of spring fluids from Dorado Outcrop, a basaltic edifice on 23 M.y. old seafloor of the Cocos Plate, eastern Pacific Ocean. These samples were collected from the discharge of a cool hydrothermal system (CHS) on a ridge flank, where typical reaction temperatures in the volcanic crust are low (2-20 °C) and fluid residence times are short. Ridge-flank hydrothermal systems extract 25% of Earth's lithospheric heat, with a global discharge rate equivalent to that of Earth's river discharge to the ocean; CHSs comprise a significant fraction of this global flow. Upper crustal temperatures around Dorado Outcrop are ∼15 °C, the calculated residence time is V, U, Mg, phosphate, Si and Li are different. Applying these observed differences to calculated global CHS fluxes results in chemical fluxes for these ions that are ≥15% of riverine fluxes. Fluxes of K and B also may be significant, but better analytical resolution is required to confirm this result. Spring fluids also have ∼50% less dissolved oxygen (DO) than bottom seawater. Calculations of an analytical model suggest that the loss of DO occurs primarily (>80%) within the upper basaltic crust by biotic and/or abiotic consumption. This calculation demonstrates that permeable pathways within the upper crust can support oxic water-rock interactions for millions of years.

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

Les premiers tours du monde à forfait. L’exemple de la Société des Voyages d’Etudes Autour du Monde (1878)

OpenAIRE

Gauthier, Lionel

2012-01-01

Dans la seconde moitié du XIXe siècle, avec le développement du bateau à vapeur et du chemin de fer, et l’ouverture de nouveaux axes de transport comme le Canal de Suez, le voyage autour du monde devient envisageable pour les touristes. Ainsi, dès 1872 des tours du monde à forfait sont organisés, ce qui témoigne de l’entrée du voyage autour du monde dans l’industrie touristique naissante. Cet article étudie ces premiers voyages d’un nouveau genre à travers l’exemple de la Société des Voyages ...

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-07-01

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

Les dispositifs du Net art

OpenAIRE

Fourmentraux, Jean-Paul

2010-01-01

La pratique du Net art radicalise la question du potentiel communicationnel d’un média —Internet— qui constitue tout à la fois le support technique, l’outil créatif et le dispositif social de l’œuvre. Les technologies de l’information et de la communication (TIC) placent en effet l’œuvre d’art au cœur d’une négociation socialement distribuée entre l’artiste et le public. L’article est focalisé sur cette construction collective du Net art et sur ses mises en scènes. Il montre le travail artist...

Bassins versants du Loup, de la Cagne et du Malvan

OpenAIRE

Lepère, Cédric; Lautier, Laurence; Pellegrino, Emmanuel

2013-01-01

Identifiant de l'opération archéologique : 8453 Date de l'opération : 2007 (PC) ; 2007 (PI) Inventeur(s) : Lepère Cédric (AUT) ; Lautier Laurence (AUT) ; Pellegrino Emmanuel (AUT) Une campagne de prospection inventaire a été effectuée pendant trois mois, dans les bassins-versants de la Cagne, du Loup et du Malvan qui regroupent les communes de Cagnes-sur-Mer, Villeneuve-Loubet, La Colle-sur-Loup, Saint-Paul-de-Vence, Vence, Saint-Jeannet, Bezaudun, Tourrettes-sur-Loup, Roquefort-les-Pins, Le ...

Houilleres du Bassin de Lorraine - calculation of AFC drive power

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

Calculation method drawn up as a result of a test programme carried out by the CoRT working group on 'Coal and dirt clearance from the face - Face-ends' on 5 faces in the H.B.L. Describes the features of the conveyors. Presents a general formula for calculating the power requirement; how this is applied. Sets forth the experimental method for determining the formula coefficients plus annotations. Includes a table summarizing the various power factors (measured and theoretically calculated). Concludes that from the proposed methods, the best is that used by Potasses d'Alsace, on condition that certain of the coefficients are empirically validated. Includes a table giving features of the workings and the conveyors used.

Visite du Conseil des gouverneurs du CRDI en Afrique de l'Est ...

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

17 oct. 2017 ... La délégation dans la communauté La délégation à l'école féminine. La délégation avec les membres du Women's Group. Durant la visite, Jean Lebel et Molapo Qhobela, directeur général de la Fondation nationale de la recherche de l'Afrique du Sud, ont annoncé les projets retenus au Kenya et en ...

La L.O.L.F. et les projets annuels de performance (P.A.P.) : Elaboration des figures du citoyen, de l'usager, du contribuable et du service public

OpenAIRE

Eyraud, Corine

2006-01-01

Nous nous intéresserons ici à la fois aux questions de la mesure des effets de l'action publique - ici l'action publique éducative universitaire -, aux dispositifs qui « construisent » les figures de l'usager (en tant que client ?), du citoyen, du contribuable et du service public, et aux nouvelles formes de démocratie que ces dispositifs génèrent (ou pas).

Flux at a point in MCNP

International Nuclear Information System (INIS)

Cashwell, E.D.; Schrandt, R.G.

1980-01-01

The current state of the art of calculating flux at a point with MCNP is discussed. Various techniques are touched upon, but the main emphasis is on the fast improved version of the once-more-collided flux estimator, which has been modified to treat neutrons thermalized by the free gas model. The method is tested on several problems on interest and the results are presented

COMMUNICATION DU CREDIT AGRICOLE - French version only

CERN Multimedia

2003-01-01

La Direction du Crédit Agricole informe son aimable clientèle du CERN des jours et heures d'ouverture de l'agence du site de Prévessin à compter du mardi 14 janvier 2003 : 1. Horaires pour les opérations bancaires courantes 7 jours sur 7 et 24h/24 avec l'espace libre service bancaire. 2. Horaires conseil du mardi au vendredi - Mardi, de 9h. à 12h. et de 14h.15 à 16h.30. - Mercredi, jeudi et vendredi, de 9h. à 12h. et de 13h.30 à 16h.30. Deux collaboratrices au lieu d'une seront désormais présentes toute la journée du mardi au vendredi pour vous accueillir, vous informer et vous conseiller en crédits et placements (réception conseil sur rendez-vous). Autre nouveauté : les mêmes conseillers seront aussi à votre disposition le samedi, sur notre agence de Gex, de 8h.15 à 13h.05, notamment pour les études de financements habitat. La Direction et toute l'équipe de l'agence du Crédit Agricole vous souhaitent une excellent année 2003.

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

Energy Technology Data Exchange (ETDEWEB)

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

1964-07-01

tubes and the experimental equipment which can support doses much higher than the ones which are biologically permissible. The final part of the communication describes the studies carried out on the realization of a liquid hydrogen cold sink, one of the most important experimental devices envisaged. (authors) [French] Les besoins francais en canaux pour sortie de neutrons de differentes energies sont brievement indiques. L'interet bien connu des neutrons froids (plus de 4 Angstroem) est souligne. Les grandes lignes d'un reacteur permettant de satisfaire les physiciens sont esquissees. Ce sont les suivantes: 1 - Flux dans l'eau lourde du reflecteur de l'ordre de 7. 10{sup 14} thermiques. 2 - Souplesse d'emploi maximum obtenue par: - separation physique du coeur et du reflecteur, - independance des experiences entre elles, - possibilite de modification, sans interruption notable du fonctionnement de la pile, des experiences physiques jusqu'a - et y compris - la nature du reflecteur utilise, - reduction au minimum des protections fixes; emploi largement generalise des protections liquides (eau) et fluidisees (sables). 3 - Continuite technologique aussi grande que possible avec les reacteurs de recherche francais existant ou en construction (SILOE, PEGASE, OSIRIS). 4 - Surete de fonctionnement recherche par la simplicite de conception. 5 - Minimisation des frais de construction. La reduction des frais d'exploitation est recherchee plutot indirectement par la simplicite des solutions et la reduction du personnel d'exploitation, que directement par la minimisation des consommations d'elements combustibles et d'energie. La solution preconisee peut etre decrite comme un reacteur de type piscine a coeur clos, non pressurise, tres sous modere par l'eau legere de refroidissement. Entourant le reacteur, se trouvent un certain nombre de 'canaux boucles' comprenant chacun: - une portion du reflecteur (eau lourde dans l'exemple decrit), - une portion de canal d'extraction de neutrons

Paul Celan in Translation: "Du sei wie du"

Directory of Open Access Journals (Sweden)

John Felstiner

1983-09-01

Full Text Available Translating the lyric poetry of Paul Celan, especially his later poems, carries not only the endemic challenge and difficulty of any verse translation, but the added incentive of doing justice to a writer whose whole recourse after the Holocaust—whose sanctuary, if he was to have any at all—he sought in language itself, specifically in the Muttersprache , the mother tongue that was as well the tongue of those who murdered his mother and father. This essay exposes a process of translating "Du sei wie du" (1970, which perhaps more than any other poem by Celan, at once solicits and defies translation, moving as it does from modern to medieval German, and closing with Hebrew words from Isaiah— a messianic imperative that shows Celan verging as ever on his Jewish identity.

Gestion intégrée du mildiou du mil en station au centre régional de ...

African Journals Online (AJOL)

user

Le mil (Pennicetum glaucum) L.R.Br constitue 75% de la production céréalière du Niger. Cependant, son rendement est très faible dû à plusieurs types de contraintes. La maladie du mildiou du mil causé par un champignon Sclerospora graminicola (Sacc) Schroët, occupe une place importante. L'objectif de cette étude.

Festival du rire de Genève

CERN Document Server

Staff Association

2015-01-01

Connaissez-vous le Festival du rire de Genève ? La deuxième édition aura lieu du 25 au 28 mars 2015 au Casino-Théâtre à Carouge. Côté programmation, Marc Donnet-Monay ouvre les festivités avant trois autres soirées de folie et d’humour que nous vous laissons le soin de découvrir dans le programme : http://www.rire-geneve.ch/#programme. Réduction de 30% sur l’achat de places pour les membres du personnel du CERN. Pour cela, il suffit de se rendre sur la billetterie en ligne de notre site : www.rire-geneve.ch et d’utiliser le code promotionnel. Contacter le secrétariat de l’Association du personnel (Staff.Association@cern.ch) pour connaitre ce code promotionnel.

Use of potassium-42 in the study of kidney functioning; Emploi du patassium-12 pour l'etude du fonctionnement renal

Energy Technology Data Exchange (ETDEWEB)

Morel, F; Guinnebault, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

Following an intravenous injection of potassium-42 as indicator, an analysis of the specific activity vs. time curve in arterial plasma, in venous plasma efferent from the kidney, in urine and in various regions of the kidney of rabbits reveals that: 1) The turnover rate of potassium in the cortex cells (proximal and distal convoluted tubes) is very large, being limited only by renal blood flow. 2) The turnover rate of potassium in deep regions (Henle loops and collector tubules) is much smaller. 3) Potassium in the urine comes from cells of the convoluted tubes and not from cells of Henle loops, collector ducts, or glomerular filtrate. 4) Any potassium filtered at the level of the glomerules would be entirely reabsorbed at the level of the proximal tube, while total potassium in the urine results from a process of excretion by cells of the distal tube. These results are comparable with the assumption that the movement of potassium between interstitial medium and convoluted tube cells results from entirely passive processes. (author) [French] Apres injection intraveineuse au lapin de radiopotassium comme indicateur, l'analyse des courbes de la radioactivite specifique du potassium, mesuree en fonction du temps dans le plasma arteriel, dans le plasma veineux efferent du rein, dans l'urine et dans diverses regions du rein, lui-meme, permet de montrer: 1)que la vitesse de renouvellement du potassium contenu dans les cellules du cortex (tubes contournes proximaux et distaux), apparait tres grande et semble limitee par le debit sanguin renal. 2) que le vitesse de renouvellement du potassium contenu dans les regions profondes (anses de Henle et tubes collecteurs) est beaucoup plus faible. 3) que le potassium de l'urine a pour precurseur le potassium des cellules des tubes contournes et non celui des cellules des anses de Henle ou des canaux collecteurs, ni celui du filtrat glomerulaire. 4) que le potassium filtre au niveau des glomerules serait entierement reabsorbe au

Implementação computacional do modelo carga-fluxo de carga-fluxo de dipolo para cálculo e interpretação das intensidades do espectro infravermelho Computational implementation of the model charge-charge flux-dipole flux for calculation and analysis of infrared intensities

Directory of Open Access Journals (Sweden)

Thiago C. F. Gomes

2008-01-01

Full Text Available The first computational implementation that automates the procedures involved in the calculation of infrared intensities using the charge-charge flux-dipole flux model is presented. The atomic charges and dipoles from the Quantum Theory of Atoms in Molecules (QTAIM model was programmed for Morphy98, Gaussian98 and Gaussian03 programs outputs, but for the ChelpG parameters only the Gaussian programs are supported. Results of illustrative but new calculations for the water, ammonia and methane molecules at the MP2/6-311++G(3d,3p theoretical level, using the ChelpG and QTAIM/Morphy charges and dipoles are presented. These results showed excellent agreement with analytical results obtained directly at the MP2/6-311++G(3d,3p level of theory.

Bulletin du CRDI #127

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

La mise à l'échelle de la recherche et de l'innovation en vue de créer un impact social constitue une priorité pour la communauté du développement. Toutefois ... Nous avons renouvelé notre soutien à la recherche auprès du gouvernement de l'Inde ... Des femmes étudient à l'École supérieure d'infotronique d'Haïti.

Performances comparées du HDL-cholestérol et du ratio cholestérol ...

African Journals Online (AJOL)

Pour le dépistage du SMet, l'AUC du CT/HDL-C est de 0,69 (IC 95% 0,61-0,77) chez les ... high blood pressure (BP), high fasting glucose, low HDL-C and high triglycerides. Areas under the "Receiver operator characteristic" curves (AUC)

Neutrino fluxes produced by high energy solar flare particles

International Nuclear Information System (INIS)

Kolomeets, E.V.; Shmonin, V.L.

1975-01-01

In this work the calculated differential energy spectra of neutrinos poduced by high energy protons accelerated during 'small' solar flares are presented. The muon flux produced by neutrino interactions with the matter at large depths under the ground is calculated. The obtained flux of muons for the total number of solar flare accelerated protons of 10 28 - 10 32 is within 10 9 - 10 13 particles/cm 2 X s x ster. (orig.) [de

La reconstruction du sourcil par greffon composite du cuir chevelu: une astuce pour faciliter la technique

Science.gov (United States)

El Omari, Mounia; El Mazouz, Samir; Gharib, Noureddine; EL Abbassi, Abdallah

2015-01-01

Les sourcils jouent un rôle important dans l’équilibre esthétique du visage. Leur reconstruction ou ophriopoïése, après séquelle de brûlure fait partie intégrante du programme de réhabilitation de la face brûlée. Plusieurs techniques ont été décrites. Nous insistons ici sur l'intérêt d'une technique simple, à la portée de tous les chirurgiens, et dont la méthode et les résultats peuvent être améliorés par un dessin bien planifié des zones donneuse et receveuse: la greffe composite prélevée au niveau du cuir chevelu dessinée à l'aide d'un calque du sourcil controlatéral. PMID:26401195

Adjusted neutron spectra of STEK cores for reactivity calculations

International Nuclear Information System (INIS)

Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.

1978-02-01

Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK