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Sample records for calcul des fissions

  1. Equilibrium fission model calculations

    International Nuclear Information System (INIS)

    Beckerman, M.; Blann, M.

    1976-01-01

    In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling

  2. Fission neutron multiplicity calculations

    International Nuclear Information System (INIS)

    Maerten, H.; Ruben, A.; Seeliger, D.

    1991-01-01

    A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs

  3. 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)

  4. A revised calculational model for fission

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F

    1998-09-01

    A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)

  5. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    Science.gov (United States)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  6. Calculated fission properties of the heaviest elements

    International Nuclear Information System (INIS)

    Moeller, P.; Nix, J.R.; Swiatecki, W.J.

    1986-09-01

    A quantitative calculation is presented that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. For the macroscopic part a Yukawa-plus-exponential model is used and for the microscopic part a folded-Yukawa single-particle potential is used. The three-quadratic-surface parameterization generates shapes for which the potential-energy surfaces are calculated. The use of this parameterization and the use of the finite-range macroscopic model allows for the study of two touching spheres and similar shapes. The results of the calculations in terms of potential-energy surfaces and fission half-lives are presented for heavy even nuclei. The surfaces are displayed in the form of contour diagrams as functions of two moments of the shape. 53 refs., 15 figs., 1 tab

  7. Detection of fission fragments by secondary emission; Detection des fragments de fission par emission secondaire

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    This fission fragment detecting apparatus is based on the principle that fragments traversing a thin foil will cause emission of secondary electrons. These electrons are then accelerated (10 kV) and directly detected by means of a plastic scintillator and associated photomultiplier. Some of the advantages of such a detector are, its rapidity, its discriminating power between alpha particles and fission fragments, its small energy loss in detecting the fragments and the relatively great amount of fissionable material which it can contain. This paper is subdivided as follows: a) theoretical considerations b) constructional details of apparatus and some experimental details and c) a study of the secondary emission effect itself. (author) [French] Le detecteur de fragments de fission que nous avons realise est base sur le principe de l'emission secondaire produite par les fragments de fission traversant une feuille mince: les electrons secondaires emis sont acceleres a des tensions telles (de l'ordre de 10 kV), qu'ils soient directement detectables par un scintillateur plastique associe a un photomultiplicateur. L'interet d'un tel detecteur reside: dans sa rapidite, sa tres bonne discrimination alpha, fission, la possibilite de detecter les fragments de fission avec une perte d'energie pouvant rester relativement faible, et la possibilite d'introduire des quantites de matiere fissile plus importantes que dans les autres types de detecteurs. Ce travail comporte: -) un apercu bibliographique de la theorie du phenomene, -) realisation et mise au point du detecteur avec etude experimentale de quelques parametres intervenant dans l'emission secondaire, -) etude de l'emission secondaire (sur la face d'emergence des fragments de fission) en fonction de l'energie du fragment et en fonction de l'epaisseur de matiere traversee avant emission secondaire, et -) une etude comparative de l'emission secondaire sur la face d'incidence et sur la face d'emergence des fragments de

  8. Model for fission-product calculations

    International Nuclear Information System (INIS)

    Smith, A.B.

    1984-01-01

    Many fission-product cross sections remain unmeasurable thus considerable reliance must be placed upon calculational interpolation and extrapolation from the few available measured cross sections. The vehicle, particularly for the lighter fission products, is the conventional optical-statistical model. The applied goals generally are: capture cross sections to 7 to 10% accuracies and inelastic-scattering cross sections to 25 to 50%. Comparisons of recent evaluations and experimental results indicate that these goals too often are far from being met, particularly in the area of inelastic scattering, and some of the evaluated fission-product cross sections are simply physically unreasonable. It is difficult to avoid the conclusion that the models employed in many of the evaluations are inappropriate and/or inappropriately used. In order to alleviate the above unfortunate situations, a regional optical-statistical (OM) model was sought with the goal of quantitative prediction of the cross sections of the lighter-mass (Z = 30-51) fission products. The first step toward that goal was the establishment of a reliable experimental data base consisting of energy-averaged neutron total and differential-scattering cross sections. The second step was the deduction of a regional model from the experimental data. It was assumed that a spherical OM is appropriate: a reasonable and practical assumption. The resulting OM then was verified against the measured data base. Finally, the physical character of the regional model is examined

  9. Adsorption of fission products on mediterranean mud; Adsorption des produits de fission sur des vases de mediterranee

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, P; Gailledreau, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    Partition coefficients of some fission products have been measured in sea water on mud taken from the bottom of the Mediterranean sea. A discussion follows on the behaviour of these radioisotopes. (author) [French] On a mesure les coefficients de partage dans l'eau de mer de quelques produits de fission a longue periode sur des echantillons de vase preleves en Mediterranee. Les valeurs trouvees sont elevees. Le comportement de ces isotopes radioactifs est discutee. (auteur)

  10. On the calculation of multi-group fission spectrum vectors

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1984-05-01

    In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

  11. Calculated apparent yields of rare gas fission products

    International Nuclear Information System (INIS)

    Delucchi, A.A.

    1975-01-01

    The apparent fission yield of the rare gas fission products from four mass chains is calculated as a function of separation time for six different fissioning systems. A plot of the calculated fission yield along with a one standard deviation error band is given for each rare gas fission product and for each fissioning system. Those parameters in the calculation that were major contributors to the calculated standard deviation at each separation time were identified and the results presented on a separate plot. To extend the usefulness of these calculations as new and better values for the input parameters become available, a third plot was generated for each system which shows how sensitive the derived fission yield is to a change in any given parameter used in the calculation. (U.S.)

  12. Fission yield covariance generation and uncertainty propagation through fission pulse decay heat calculation

    International Nuclear Information System (INIS)

    Fiorito, L.; Diez, C.J.; Cabellos, O.; Stankovskiy, A.; Van den Eynde, G.; Labeau, P.E.

    2014-01-01

    Highlights: • Fission yield data and uncertainty comparison between major nuclear data libraries. • Fission yield covariance generation through Bayesian technique. • Study of the effect of fission yield correlations on decay heat calculations. • Covariance information contribute to reduce fission pulse decay heat uncertainty. - Abstract: Fission product yields are fundamental parameters in burnup/activation calculations and the impact of their uncertainties was widely studied in the past. Evaluations of these uncertainties were released, still without covariance data. Therefore, the nuclear community expressed the need of full fission yield covariance matrices to be able to produce inventory calculation results that take into account the complete uncertainty data. State-of-the-art fission yield data and methodologies for fission yield covariance generation were researched in this work. Covariance matrices were generated and compared to the original data stored in the library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235 U. Calculations were carried out using different libraries and codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the libraries. The uncertainty quantification was performed first with Monte Carlo sampling and then compared with linear perturbation. Indeed, correlations between fission yields strongly affect the uncertainty of decay heat. Eventually, a sensitivity analysis of fission product yields to fission pulse decay heat was performed in order to provide a full set of the most sensitive nuclides for such a calculation

  13. Some methods for the detection of fissionable matter; Quelques methodes de detection des corps fissiles

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-03-01

    A number of equipments or processes allowing to detect uranium or plutonium in industrial plants, and in particular to measure solution concentrations, are studied here. Each method has its own field of applications and has its own performances, which we have tried to define by calculations and by experiments. The following topics have been treated: {gamma} absorptiometer with an Am source, detection test by neutron multiplication, apparatus for the measurement of the {alpha} activity of a solution, fissionable matter detection by {gamma} emission, fissionable matter detection by neutron emission. (author) [French] On examine ici plusieurs appareils ou procedes qui permettent de detecter l'uranium ou le plutonium dans les installations industrielles, et en particulier de mesurer les concentrations de solutions. Chacune des methodes a son domaine d'application et ses performances, qu'on a tente de definir par le calcul et par des experiences. Les sujets traites sont les suivants: absorptiometre {gamma} a source d'americium, essais de detection par multiplication neutronique, appareil de mesure de l'activite {alpha} d'une solution, detection des matieres fissiles par leur emission {gamma}, detection des matieres fissiles par leur emission neutronique. (auteur)

  14. Sensitivity and uncertainty analysis for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Rebah, J.; Lee, Y.K.; Nimal, J.C.; Nimal, B.; Luneville, L.; Duchemin, B.

    1994-01-01

    The calculated uncertainty in decay heat due to the uncertainty in basic nuclear data given in the CEA86 Library, is presented. Uncertainties in summation calculation arise from several sources: fission product yields, half-lives and average decay energies. The correlation between basic data is taken into account. The uncertainty analysis were obtained for thermal-neutron-induced fission of U235 and Pu239 in the case of burst fission and irradiation time. The calculated decay heat in this study is compared with experimental results and with new calculation using the JEF2 Library. (from authors) 6 figs., 19 refs

  15. ELSA: A simplified code for fission product release calculations

    International Nuclear Information System (INIS)

    Manenc, H.; Notley, M.J.

    1996-01-01

    During a light water reactor severe accident, fission products are released from the overheated core as it progressively degrades. A new computer module named ELSA is being developed to calculate fission product release. The authors approach is to model the key phenomena, as opposed to more complete mechanistic approaches. Here they present the main features of the module. Different release mechanisms have been identified and are modeled in ELSA, depending on fission product volatility: diffusion seems to govern the release of the highly volatile species if fuel oxidation is properly accounted for, whereas mass transport governs that of lower volatility fission products and fuel volatilization that of the practically involatile species

  16. Trajectory calculations for the ternary cold fission of 252Cf

    International Nuclear Information System (INIS)

    Misicu, S.

    1998-01-01

    We compute the final kinetic energies of the fragments emitted in the light charged particle accompanied by cold fission of 252 Cf taking into account the deformation and the finite-size effects of the fragments and integrating the equations of motion for a three-body system subjected only to Coulomb forces. The initial conditions for the trajectory calculations were derived in the framework of a deformed cluster model which includes also the effect due to the absorbative nuclear part. Although the distributions of initial kinetic energies are rather broad we show that in cold fission the initial conditions can be better determined than in the usual spontaneous fission

  17. Fission properties of superheavy nuclei for r -process calculations

    Science.gov (United States)

    Giuliani, Samuel A.; Martínez-Pinedo, Gabriel; Robledo, Luis M.

    2018-03-01

    We computed a new set of static fission properties suited for r -process calculations. The potential energy surfaces and collective inertias of 3640 nuclei in the superheavy region are obtained from self-consistent mean-field calculations using the Barcelona-Catania-Paris-Madrid energy density functional. The fission path is computed as a function of the quadrupole moment by minimizing the potential energy and exploring octupole and hexadecapole deformations. The spontaneous fission lifetimes are evaluated employing different schemes for the collective inertias and vibrational energy corrections. This allows us to explore the sensitivity of the lifetimes to those quantities together with the collective ground-state energy along the superheavy landscape. We computed neutron-induced stellar reaction rates relevant for r -process nucleosynthesis using the Hauser-Feshbach statistical approach and study the impact of collective inertias. The competition between different reaction channels including neutron-induced rates, spontaneous fission, and α decay is discussed for typical r -process conditions.

  18. Prompt neutron emission; Emission des neutrons prompts de fission

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

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

  19. Fission cross-section calculations and the multi-modal fission model

    International Nuclear Information System (INIS)

    Hambsch, F.J.

    2004-01-01

    New, self consistent, neutron-induced reaction cross section calculations for 235,238 U, 237 Np have been performed. The statistical model code STATIS was improved to take into account the multimodality of the fission process. The three most dominant fission modes, the two asymmetric standards I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode have been taken into account. De-convoluted fission cross sections for those modes for 235,238 U(n,f) and 237 Np(n,f) based on experimental branching ratios, were calculated for the first time up to the second chance fission threshold. For 235 U(n,f), the calculations being made up to 28 MeV incident neutron energy, higher fission chances have been considered. This implied the need for additional calculations for the neighbouring isotopes. As a side product also mass yield distributions could be calculated at energies hitherto not accessible by experiment. Experimental validation of the predictions is being envisaged

  20. On the Calculation of the Fast Fission Factor

    Energy Technology Data Exchange (ETDEWEB)

    Almgren, B

    1960-06-15

    Definitions of the fast fission factor {epsilon} are discussed. Different methods of calculation of {epsilon} are compared. Group constants for one - , two- and three-group calculations have been evaluated using the best obtainable basic data. The effects of back-scattering, coupling and (n,2n) reactions are discussed.

  1. Map of calculated radioactivity of fission product, (4)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1978-07-01

    The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

  2. Separation of fission products by the use of recoil; Separation des produits de fission par utilisation du recul

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R; Beydon, J; Bardy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    We have studied fission recoil in U{sub 3}O{sub 8} organic solvent mixtures. The organic phase chosen was first naphtalene then terphenyl. Graphite and activated carbon were also tried out as recoil media. We first verified that the fission fragments are ejected from the uranium oxide particles under our experimental conditions. The retention phenomenon observed is due to an adsorption occurring either during irradiation or during the chemical treatment. Using naphthalene or terphenyl, the individual separation of the fission products has made it possible to show the influence of the chemical nature of the recoil medium on the retention of each fission product. We put forward a hypothesis concerning this phenomenon: experiments carried out using 'scavengers', together with kinetic studies make it possible to explain the retention phenomenon and to choose the most favorable conditions for reducing this retention to a low value. The thermal recombination kinetics demonstrate the influence of the fission ion charge on the final value of the retention for a given temperature. The origins of this thermal recombination are discussed. (author) [French] On a etudie le recul de fission dans les melanges U{sub 3}0{sub 8}, phase organique. La phase organique choisie a ete le naphtalene puis le terphenyle. Le graphite et le charbon actif ont egalement ete essayes comme milieux de recul. On a d'abord determine que les fragments de fission sortent des particules d'oxyde d'uranium avec un rendement de 100 pour cent dans nos conditions experimentales. Le phenomene de retention observe est du a une adsorption ayant lieu pendant l'irradiation ou pendant le traitement chimique. Dans le naphtalene et le terphenyle, la separation individuelle des produits de fission a permis de mettre en evidence l'influence de la nature chimique du milieu de recul sur la retention de chaque produit de fission. On avance une hypothese sur ce phenomene: des experiences effectuees avec des 'scavengers

  3. New Burnup Calculation System for Fusion-Fission Hybrid System

    International Nuclear Information System (INIS)

    Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

    2006-01-01

    Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise

  4. Calculation of fission gases internal pressure in nuclear fuel rods

    International Nuclear Information System (INIS)

    Vasconcelos Santana, M. de.

    1981-12-01

    Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

  5. Calculation of vapor pressure of fission product fluorides and oxyfluorides

    International Nuclear Information System (INIS)

    Roux, J.P.

    1976-03-01

    The equilibrium diagrams of the condensed phases - solid and liquid - and vapor phase are collected for the principal fluorides and oxyfluorides of fission product elements (atomic number from 30 to 66). These diagrams are used more particularly in fuel reprocessing by fluoride volatility process. Calculations and curves (vapor pressure in function of temperature) are processed using a computer program given in this report [fr

  6. Quantitative analysis of fission products by {gamma} spectrography; Analyse quantitative des produits de fission par spectrographie {gamma}

    Energy Technology Data Exchange (ETDEWEB)

    Malet, G

    1962-07-01

    The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio ({sup 144}Ce + {sup 144}Pr activity/{sup 137}Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By {gamma}-scintillation spectrography it was possible to estimate the following elements individually: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author) [French] L'activite des produits de fission presents dans les solutions de traitement de combustibles irradies est donnee en fonction du temps de refroidissement et du temps d'irradiation. On etudie de plus la variation du rapport Activite du {sup 144}Ce + {sup 144}Pr /Activite du {sup 137}Cs en fonction de ces memes parametres. De ces resultats, on deduit une methode donnant l'age de la solution analysee. La spectrographie {gamma} a scintillation a permis le dosage individuel des produits suivants: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Des courbes de rendement sont donnees dans le cas d'un emetteur unique. Des differentes methodes existantes, la methode des moindres carres a ete employee pour l'analyse quantitative des produits de fission precites. La precision obtenue varie entre 3 et 10 pour cent. (auteur)

  7. Recent progress in fission product separation; Progres recents de la separation des produits de fission

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    saturated with magnesium in order to avoid as far as possible the extraction of strontium, extraction of strontium at pH 4. (author) [French] La methode decrite a Geneve en 1958 a ete experimentee avec succes. Des ameliorations importantes ont ete apportees a ce procede: 1 - Initialement, la precipitation phosphotungstique du cesium etait suivie d'une reprise a la baryte en centrifugeuse et d'une distillation de l'ammoniac dans un concentrateur. Ensuite, l'hydroxyde de baryum etait elimine par precipitation carbonique et centrifugation. 1 - Il a ete prouve que la distillation de l'ammoniac pouvait etre remplacee par une evaporation au cours de la centrifugation, ce qui supprime le concentrateur. Il a ensuite ete possible d'effectuer la carbonatation sur le melange solide-liquide provenant de l'attaque a la baryte. 2 - Pour appliquer le precedent procede au traitement des solutions provenant des combustibles uranium-molybdene, nous avons recommande une concentration qui maintient en solution le molybdene par complexation par l'acide phosphorique. Cette complexation fournit une suspension de phosphate de zirconium et de phosphotungstate d'ammonium. Ils sont separes en repassant en milieu basique qui fait precipiter la zircone, puis en revenant en milieu acide, la suite du traitement restant le meme. 3 - Les etudes sur les proprietes d'echange des sels d'hetero-polyacides poursuivies dans plusieurs pays ont toujours rencontre l'ecueil des mauvaises qualites mecaniques de ces produits. Cette difficulte a ete surmontee en emprisonnant le phosphotungstate d'ammonium dans une matrice en phosphate de zirconium. L'echangeur obtenu possede: - des proprietes mecaniques satisfaisantes - une capacite de 0,1 milliequivalent par gramme en milieu acide nitrique concentre. II est eluable et regenerable par une solution d'un sel d'ammonium. Le schema de recuperation de ces differents produits de fission est succinctement le suivant: - extraction des terres rares par l'acide di-2-ethyl hexyl

  8. Uncertainties in fission-product decay-heat calculations

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, K.; Ohta, H.; Miyazono, T.; Tasaka, K. [Nagoya Univ. (Japan)

    1997-03-01

    The present precision of the aggregate decay heat calculations is studied quantitatively for 50 fissioning systems. In this evaluation, nuclear data and their uncertainty data are taken from ENDF/B-VI nuclear data library and those which are not available in this library are supplemented by a theoretical consideration. An approximate method is proposed to simplify the evaluation of the uncertainties in the aggregate decay heat calculations so that we can point out easily nuclei which cause large uncertainties in the calculated decay heat values. In this paper, we attempt to clarify the justification of the approximation which was not very clear at the early stage of the study. We find that the aggregate decay heat uncertainties for minor actinides such as Am and Cm isotopes are 3-5 times as large as those for {sup 235}U and {sup 239}Pu. The recommended values by Atomic Energy Society of Japan (AESJ) were given for 3 major fissioning systems, {sup 235}U(t), {sup 239}Pu(t) and {sup 238}U(f). The present results are consistent with the AESJ values for these systems although the two evaluations used different nuclear data libraries and approximations. Therefore, the present results can also be considered to supplement the uncertainty values for the remaining 17 fissioning systems in JNDC2, which were not treated in the AESJ evaluation. Furthermore, we attempt to list nuclear data which cause large uncertainties in decay heat calculations for the future revision of decay and yield data libraries. (author)

  9. Fission of heavy nuclei: microscopic study of fission barriers and fragments angular momentum; Fission des noyaux lourds: etude microscopique des barrieres de fission et du moment angulaire des fragments

    Energy Technology Data Exchange (ETDEWEB)

    Bonneau, L

    2003-11-01

    A lot of experimental data on nuclear fission has been being collected for the last 65 years, allowing theoreticians to confront their models with reality. The first part of this work is dedicated to the computation of fission barriers. We have extended the HF + BCS (Hartree Fock + Bandeen-Cooper-Schrieffer) method in order to include a new set of polynomials on which wave functions can be broken to, more accurately than on Hermite's polynomials in the 2 fragment configuration. The fission barriers of 26 heavy nuclei from Thorium-230 to Nobelium-256 have been assessed and compared to experimental data, it appears that differences are no greater than 1 MeV. We have discovered a neat correlation between the variation of the experimental fission lifetimes of even Fermium isotopes and the computed heights of second barriers. Moreover our model reproduces the hyper-deformed well of Thorium-230 with a good agreement on the well depth. The second part deals with the scission region. We have performed Hartree-Fock calculations in order to explore different ways of fragmentation. We have shown that the harmonic oscillator gives a valid description of such ways. In order to compute the mean value of J{sup 2} in the fragments we have been driven to propose an adequate definition of that quantity consistent with the non-locality property of the J{sup 2} operator. (A.C.)

  10. Fission of heavy nuclei: microscopic study of fission barriers and fragments angular momentum; Fission des noyaux lourds: etude microscopique des barrieres de fission et du moment angulaire des fragments

    Energy Technology Data Exchange (ETDEWEB)

    Bonneau, L

    2003-11-01

    A lot of experimental data on nuclear fission has been being collected for the last 65 years, allowing theoreticians to confront their models with reality. The first part of this work is dedicated to the computation of fission barriers. We have extended the HF + BCS (Hartree Fock + Bandeen-Cooper-Schrieffer) method in order to include a new set of polynomials on which wave functions can be broken to, more accurately than on Hermite's polynomials in the 2 fragment configuration. The fission barriers of 26 heavy nuclei from Thorium-230 to Nobelium-256 have been assessed and compared to experimental data, it appears that differences are no greater than 1 MeV. We have discovered a neat correlation between the variation of the experimental fission lifetimes of even Fermium isotopes and the computed heights of second barriers. Moreover our model reproduces the hyper-deformed well of Thorium-230 with a good agreement on the well depth. The second part deals with the scission region. We have performed Hartree-Fock calculations in order to explore different ways of fragmentation. We have shown that the harmonic oscillator gives a valid description of such ways. In order to compute the mean value of J{sup 2} in the fragments we have been driven to propose an adequate definition of that quantity consistent with the non-locality property of the J{sup 2} operator. (A.C.)

  11. The MCEF code for nuclear evaporation and fission calculations

    International Nuclear Information System (INIS)

    Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.

    2001-11-01

    We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)

  12. Thermodynamic cycle calculations for a pumped gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.

    1991-01-01

    Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

  13. Experimental study of the thermal fission of uranium 235 in the region of symmetrical masses; Contribution a l'etude experimentale de la fission thermique de l'uranium 235 dans la region des masses symetriques

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-03-15

    Energy correlation experiments with fission fragments are strongly perturbed, in the symmetric region, by the detection of spurious events caused by the apparatus. We show that the measurement of an additional parameter, namely the difference in time-of-flight between the fragments, enables us to eliminate these difficulties. In this work we show also an original method of calibration of the time-of-flight set-up. For thermal fission of {sup 235}U, values of mass yields in the symmetric region are found to agree quantitatively with radiochemical values. Moreover, the average total kinetic energy distribution as a function of the pre-neutron emission masses of the fragments has been calculated. This curve presents in the symmetric region a large dip, the value of which takes on the value 21.2 {+-} 0.8 MeV. This value is smaller than previously published results. (author) [French] Les mesures correlees des energies cinetiques des fragments de fission sont fortement perturbees, dans la region symetrique, par la detection d'evenements aberrants d'origine instrumentale. Nous montrons que la mesure d'un parametre supplementaire, a savoir, la difference des temps de vol des deux fragments, nous permet d'eliminer ces difficultes. Dans ce travail, nous indiquons egalement une methode originale de calibration du dispositif de mesure des temps de vol. Dans le cas de la fission thermique de {sup 235}U, nous avons trouve, dans la region symetrique, une courbe de rendement des masses, en accord quantitatif avec les donnees radiochimiques. De plus, nous avons calcule la distribution de l'energie cinetique totale moyenne en fonction de la masse des fragments, avant emission neutronique. Cette courbe presente, dans la region symetrique, un creux important, dont la valeur atteint 21,2 {+-} 0,8 MeV. Cette valeur est inferieure aux resultats precedemment publies. (auteur)

  14. Study on the calculation method of source term from fission products

    International Nuclear Information System (INIS)

    Zhou Jing; Gong Quan; Qiu Haifeng

    2014-01-01

    As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

  15. Dynamical calculations of nuclear fission and heavy-ion reactions

    International Nuclear Information System (INIS)

    Nix, J.R.; Sierk, A.J.

    1984-01-01

    With the goal of determining the magnitude and mechanism of nuclear dissipation from comparisons of predictions with experimental data, we describe recent calculations in a unified macroscopic-microscopic approach to large-amplitude collective nuclear motion such as occurs in fission and heavy-ion reactions. We describe the time dependence of the distribution function in phase space of collective coordinates and momenta by a generalized Fokker-Planck equation. The nuclear potential energy of deformation is calculated as the sum of repulsive Coulomb and centrifugal energies and an attractive Yukawa-plus-exponential potential, the inertia tensor is calculated for a superposition of rigid-body rotation and incompressible, nearly irrotational flow by use of the Werner-Wheeler method, and the dissipation ensor that describes the conversion of collective energy into single-particle excitation energy is calculated for two prototype mechanisms that represent opposite extremes of large and small dissipation. We solve the generalized Hamilton equations of motion for the first moments of the distribution function to obtain the mean translational fission-fragment kinetic energy and mass of a third fragment that sometimes forms between the two end fragments, as well as dynamical thresholds, capture cross sections, and ternary events in heavy-ion reactions. 33 references

  16. Automatic fission source convergence criteria for Monte Carlo criticality calculations

    International Nuclear Information System (INIS)

    Shim, Hyung Jin; Kim, Chang Hyo

    2005-01-01

    The Monte Carlo criticality calculations for the multiplication factor and the power distribution in a nuclear system require knowledge of stationary or fundamental-mode fission source distribution (FSD) in the system. Because it is a priori unknown, so-called inactive cycle Monte Carlo (MC) runs are performed to determine it. The inactive cycle MC runs should be continued until the FSD converges to the stationary FSD. Obviously, if one stops them prematurely, the MC calculation results may have biases because the followup active cycles may be run with the non-stationary FSD. Conversely, if one performs the inactive cycle MC runs more than necessary, one is apt to waste computing time because inactive cycle MC runs are used to elicit the fundamental-mode FSD only. In the absence of suitable criteria for terminating the inactive cycle MC runs, one cannot but rely on empiricism in deciding how many inactive cycles one should conduct for a given problem. Depending on the problem, this may introduce biases into Monte Carlo estimates of the parameters one tries to calculate. The purpose of this paper is to present new fission source convergence criteria designed for the automatic termination of inactive cycle MC runs

  17. ALGORITHME POUR LE CALCUL DES COURBURES GENERALISEES

    Directory of Open Access Journals (Sweden)

    K MEZAGHCHA

    2004-06-01

    Full Text Available On sait qu’une courbe algébrique standard  d'équation f(x, y =0 admet un nombre fini de branches (nombre inférieur à l'ordre de f , dont les paramètrages peuvent être obtenus en particulier à partir de la décomposition de Goze itérée. On aimerait calculer  leur courbure généralisée sans les déterminer explicitement, la notion de courbure généralisée ayant fait l’objet d’un travail, publié dans les comptes rendus de l’Université de Cagliari (Italie [12]. Dans cet article, on se propose d'établir à cet effet un algorithme qui donnera à partir seulement des coefficients de f, la liste exhaustive des courbures généralisées de toutes les branches réelles. L’article se termine par la donnée d’un exemple pour montrer l’efficacité de l’algorithme proposé.

  18. An application program for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Pham, Ngoc Son; Katakura, Jun-ichi

    2007-10-01

    The precise knowledge of decay heat is one of the most important factors in safety design and operation of nuclear power facilities. Furthermore, decay heat data also play an important role in design of fuel discharges, fuel storage and transport flasks, and in spent fuel management and processing. In this study, a new application program, called DHP (Decay Heat Power program), has been developed for exact decay heat summation calculations, uncertainty analysis, and for determination of the individual contribution of each fission product. The analytical methods were applied in the program without any simplification or approximation, in which all of linear and non-linear decay chains, and 12 decay modes, including ground state and meta-stable states, are automatically identified, and processed by using a decay data library and a fission yield data file, both in ENDF/B-VI format. The window interface of the program is designed with optional properties which is very easy for users to run the code. (author)

  19. Dosage of fission products in irradiated fuel treatment effluents (radio-chemical method); Dosage des produits de fission dans les effluents du traitement des combustibles irradies (methode radiochimique)

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-07-01

    The dosage methods presented here are applicable to relatively long-lived fission products present in the effluents resulting from irradiated fuel treatment processes (Sr - Cs - Ce - Zr - Nb - Ru - I). The methods are based on the same principle: - addition of a carrying-over agent - chemical separation over several purification stages, - determination of the chemical yield by calorimetry - counting of an aliquot liquid portion. (author) [French] Les methodes de dosage presentees concernent les produits de fission a vie relativement longue presents dans les effluents de traitement des combustibles irradies (Sr - Cs - Ce - Zr - Nb - Ru - I). Elles sont toutes basees sur le meme principe: - addition d'entraineur, - separation chimique en plusieurs stades de purification, - determination du rendement chimique par calorimetrie, - comptage d'une aliquote liquide. (auteur)

  20. The use of recoil for the separation of uranium fission products; Utilisation du recul pour la separation des produits de fission de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    The recoil distance of fission fragments in U{sub 3}O{sub 8} is about 8 microns. By using highly diluted suspensions of uranium oxide particles having dimension much smaller than this figure (mean diameter 0,5 micron), we were able to study the re-adsorption of fission products on uranium oxide. Separation results have been studied as a function of the nature of the irradiation medium (solid or liquid) and the separation medium, of particle size and of concentration of particles in the dispersing medium. Decay curves can be used to discriminate between {sup 239}Np and mixed fission products. Most of the {sup 239}Np is found in the U{sub 3}O{sub 8} particles. The location of fission products in solid dispersing media has been determined, fission products being found always inside the dispersing medium particles. The results obtained can be applied to the rapid separation of short-lived fission products from a uranium-free starting material. (author) [French] Le parcours de recul des fragments de fission est en moyenne de 8 microns dans l'U{sub 3}O{sub 8}. En prenant des suspensions d'oxyde d'uranium dont les particules, tres diluees, ont des dimensions nettement inferieures a cette valeur (diametre moyen 0,5 micron), on a pu etudier directement la readsorption des produits de fission sur l'oxyde d'uranium. Les resultats de separation ont ete etudies en fonction de la nature du milieu d'irradiation (solide ou liquide) et du milieu de separation, de la taille des particules d'oxyde et de leur concentration dans le milieu dispersant. Les courbes de decroissance permettent de determiner la perturbation apportee dans les mesures par le {sup 239}Np qui reste en majorite dans les grains d'U{sub 3}O{sub 8}. On a determine enfin l'emplacement des produits de fission dans le cas des melanges solides; ils se trouvent toujours a l'interieur des grains du milieu recepteur. Les resultats obtenus permettent d'envisager la separation rapide de produits de fission a periode courte a

  1. Map of calculated radioactivity of fission product, 3

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

  2. Map of calculated radioactivity of fission product, (1)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr) Vol. II Maps of radioactivity of each nuclide (Nb - Sb) Vol. III Maps of radioactivity of each nuclide (Te - Tm) (auth.)

  3. Map of calculated radioactivity of fission product, 2

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-02-01

    In this work, the radioactivities of fission products were calculated and summarized in contour maps and tables depending on irradiation and cooling times. The irradiation condition and other parameters used for the present calculation are shown in the followings. Neutron flux (N sub(th)): 3x10 13 n/sec/cm 2 Atom number of uranium: 1 mole (6x10 23 , ca. 271 gUO 2 ) Enrichment of U-235: 2.7% Range of irradiation time: 60-6x10 7 sec (ca. 1.9 y) Range of cooling time: 60-6x10 7 sec (ca. 1.9 y). Values of the neutron flux and the enrichment treated here are representative for common LWRs. The maps and tables of 560 nuclides are divided and compiled into the following three volumes. Vol. I: Maps of radioactivity of overall total, element total and each nuclide (Ni - Zr), Vol. II: Maps of radioactivity of each nuclide (Nb - Sb), Vol. III: Maps of radioactivity of each nuclide (Te - Tm). (auth.)

  4. FISPRO: a simplified computer program for general fission product formation and decay calculations

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.; Bailey, P.G.

    1979-08-01

    This report describes a computer program that solves a general form of the fission product formation and decay equations over given time steps for arbitrary decay chains composed of up to three nuclides. All fission product data and operational history data are input through user-defined input files. The program is very useful in the calculation of fission product activities of specific nuclides for various reactor operational histories and accident consequence calculations

  5. Fission cross section calculations for 209Bi target nucleus based on fission reaction models in high energy regions

    Directory of Open Access Journals (Sweden)

    Kaplan Abdullah

    2015-01-01

    Full Text Available Implementation of projects of new generation nuclear power plants requires the solving of material science and technological issues in developing of reactor materials. Melts of heavy metals (Pb, Bi and Pb-Bi due to their nuclear and thermophysical properties, are the candidate coolants for fast reactors and accelerator-driven systems (ADS. In this study, α, γ, p, n and 3He induced fission cross section calculations for 209Bi target nucleus at high-energy regions for (α,f, (γ,f, (p,f, (n,f and (3He,f reactions have been investigated using different fission reaction models. Mamdouh Table, Sierk, Rotating Liquid Drop and Fission Path models of theoretical fission barriers of TALYS 1.6 code have been used for the fission cross section calculations. The calculated results have been compared with the experimental data taken from the EXFOR database. TALYS 1.6 Sierk model calculations exhibit generally good agreement with the experimental measurements for all reactions used in this study.

  6. Recent improvements in the calculation of prompt fission neutron spectra: Preliminary results

    International Nuclear Information System (INIS)

    Madland, D.G.; LaBauve, R.J.; Nix, J.R.

    1989-01-01

    We consider three topics in the refinement and improvement of our original calculations of prompt fission neutron spectra. These are an improved calculation of the prompt fission neutron spectrum N(E) from the spontaneous fission of 252 Cf, a complete calculation of the prompt fission neutron spectrum matrix N(E,E n ) from the neutron-induced fission of 235 U, at incident neutron energies ranging from 0 to 15 MeV, and an assessment of the scission neutron component of the prompt fission neutron spectrum. Preliminary results will be presented and compared with experimental measurements and an evaluation. A suggestion is made for new integral cross section measurements. (author). 45 refs, 12 figs, 1 tab

  7. Application of dynamic pseudo fission products and actinides for accurate burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

    1996-09-01

    The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

  8. Fission yield calculation using toy model based on Monte Carlo simulation

    International Nuclear Information System (INIS)

    Jubaidah; Kurniadi, Rizal

    2015-01-01

    Toy model is a new approximation in predicting fission yield distribution. Toy model assumes nucleus as an elastic toy consist of marbles. The number of marbles represents the number of nucleons, A. This toy nucleus is able to imitate the real nucleus properties. In this research, the toy nucleons are only influenced by central force. A heavy toy nucleus induced by a toy nucleon will be split into two fragments. These two fission fragments are called fission yield. In this research, energy entanglement is neglected. Fission process in toy model is illustrated by two Gaussian curves intersecting each other. There are five Gaussian parameters used in this research. They are scission point of the two curves (R c ), mean of left curve (μ L ) and mean of right curve (μ R ), deviation of left curve (σ L ) and deviation of right curve (σ R ). The fission yields distribution is analyses based on Monte Carlo simulation. The result shows that variation in σ or µ can significanly move the average frequency of asymmetry fission yields. This also varies the range of fission yields distribution probability. In addition, variation in iteration coefficient only change the frequency of fission yields. Monte Carlo simulation for fission yield calculation using toy model successfully indicates the same tendency with experiment results, where average of light fission yield is in the range of 90fission yield is in about 135

  9. Fission yield calculation using toy model based on Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Jubaidah, E-mail: jubaidah@student.itb.ac.id [Nuclear Physics and Biophysics Division, Department of Physics, Bandung Institute of Technology. Jl. Ganesa No. 10 Bandung – West Java, Indonesia 40132 (Indonesia); Physics Department, Faculty of Mathematics and Natural Science – State University of Medan. Jl. Willem Iskandar Pasar V Medan Estate – North Sumatera, Indonesia 20221 (Indonesia); Kurniadi, Rizal, E-mail: rijalk@fi.itb.ac.id [Nuclear Physics and Biophysics Division, Department of Physics, Bandung Institute of Technology. Jl. Ganesa No. 10 Bandung – West Java, Indonesia 40132 (Indonesia)

    2015-09-30

    Toy model is a new approximation in predicting fission yield distribution. Toy model assumes nucleus as an elastic toy consist of marbles. The number of marbles represents the number of nucleons, A. This toy nucleus is able to imitate the real nucleus properties. In this research, the toy nucleons are only influenced by central force. A heavy toy nucleus induced by a toy nucleon will be split into two fragments. These two fission fragments are called fission yield. In this research, energy entanglement is neglected. Fission process in toy model is illustrated by two Gaussian curves intersecting each other. There are five Gaussian parameters used in this research. They are scission point of the two curves (R{sub c}), mean of left curve (μ{sub L}) and mean of right curve (μ{sub R}), deviation of left curve (σ{sub L}) and deviation of right curve (σ{sub R}). The fission yields distribution is analyses based on Monte Carlo simulation. The result shows that variation in σ or µ can significanly move the average frequency of asymmetry fission yields. This also varies the range of fission yields distribution probability. In addition, variation in iteration coefficient only change the frequency of fission yields. Monte Carlo simulation for fission yield calculation using toy model successfully indicates the same tendency with experiment results, where average of light fission yield is in the range of 90fission yield is in about 135

  10. Calculation of Fission Observables Through Event-by-Event Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Randrup, J; Vogt, R

    2009-06-04

    The increased interest in more exclusive fission observables has demanded more detailed models. We present here a new computational model, FREYA, that aims to met this need by producing large samples of complete fission events from which any observable of interest can then be extracted consistently, including arbitrary correlations. The various model assumptions are described and the potential utility of the model is illustrated by means of several novel correlation observables.

  11. Fission product determination in irradiated fuel processing waste (electrophoresis); Dosage des produits de fission dans les effluents de traitement des combustibles irradies (electrophorese)

    Energy Technology Data Exchange (ETDEWEB)

    Auchapt, J M; Tret, J [Commissariat a l' Energie Atomique, Centre de Marcoule, 30 - Bagnols-sur-Ceze (France). Centre de Production de Plutonium de Marcoule. Services d' Extraction du Plutonium

    1966-07-01

    This dosage method concerns fission products present in the waste produced from the processing of cooled irradiated fuels. - Sr, Cs, Ce, Y, Ru by quantitative analysis; - Zr, Nb by qualitative analysis. It includes electrophoresis on paper strips one meter long which is then analysed between two window-less Geiger counters. For an activity of 10{sup -2} {mu}Ci of any cation in a 10 {mu}l spot, the standard error {sigma} if 3 to 4 per cent. complete analysis lasts about 5 hours. (authors) [French] Cette methode de dosage concerne les produits de fission presents dans les effluents de traitement des combustibles irradies refroidis: - Sr, Cs, Ce, Y, Ru en analyse quantitative; - Zr, Nb en analyse qualitative. Elle comporte une electrophorese sur bande de papier de un metre de longueur suivie d'un depouillement entre deux compteurs Geiger sans fenetre. Pour une activite de 10{sup -2} {mu}Ci d'un cation quelconque dans une tache de 10 {mu}l l'erreur standard {sigma} est de 3 a 4 pour cent. L'analyse complete demande environ 5 heures. (auteurs)

  12. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  13. Calculating the mass distribution of heavy nucleus fission product by neutrons

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

    1981-01-01

    The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

  14. Pion-induced fission of 209Bi and 119Sn: measurements, calculations, analyses and comparison

    International Nuclear Information System (INIS)

    Rana, M.A.; Sher, G.; Manzoor, S.; Shehzad, M.I.

    2011-01-01

    Cross-sections for the π - -induced fission of 209 Bi and 119 Sn have been measured using the most sensitive CR-39 solid-state nuclear track detector. In experiments, target–detector stacks were exposed to negative pions of energy 500, 672, 1068, and 1665 MeV at the Brookhaven National Laboratory, USA. An important aspect of the present paper is the comparison of pion-induced fission fragment spectra of above mentioned nuclei with the spontaneous fission fragment spectra of 252 Cf. This comparison is made in terms of fission fragment track lengths in the CR-39 detectors. Measurement results are compared with calculations of Monte Carlo and statistical weight functions methods using the computer code CEM95. Agreement between measurements and calculations is fairly good for 209 Bi target nuclei whereas it is indigent for the case of 119 Sn. The possibilities of the trustworthy calculations, using the computer code CEM95, comparable with measurements of pion-induced fission in intermediate and heavy nuclei are explored by employing various systematics available in the code. Energy dependence of pion-induced fission in 119 Sn and 209 Bi is analyzed employing a newly defined parameter geometric-size-normalized fission cross-section (χ f g ). It is found that the collective nuclear excitations, which may lead to fission, become more probable for both 209 Bi and 119 Sn nuclei with increasing energy of negative pions from 500 to 1665 MeV. (author)

  15. Prediction of fission mass-yield distributions based on cross section calculations

    International Nuclear Information System (INIS)

    Hambsch, F.-J.; G.Vladuca; Tudora, Anabella; Oberstedt, S.; Ruskov, I.

    2005-01-01

    For the first time, fission mass-yield distributions have been predicted based on an extended statistical model for fission cross section calculations. In this model, the concept of the multi-modality of the fission process has been incorporated. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode are taken into account. De-convoluted fission cross sections for S1, S2 and SL modes for 235,238 U(n, f) and 237 Np(n, f), based on experimental branching ratios, were calculated for the first time in the incident neutron energy range from 0.01 to 5.5 MeV providing good agreement with the experimental fission cross section data. The branching ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including mean values also for incident neutron energies hitherto not accessible to experiment

  16. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  17. MADNIX a code to calculate prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

    Merchant, A.C.

    1986-03-01

    A code has been written and tested on the CDC Cyber-170 to calculate the prompt fission neutron spectrum, N(E), as a function of both the fissioning nucleus and its excitation energy. In this note a brief description of the underlying physical principles involved and a detailed explanation of the required input data (together with a sample output for the fission of 235 U induced by 14 MeV neutrons) are presented. Weisskopf's standard nuclear evaporation theory provides the basis for the calculation. Two important refinements are that the distribution of fission-fragment residual nuclear temperature and the cooling of the fragments as neutrons are emitted approximately taken into account, and also the energy dependence of the cross section for the inverse process of compound nucleus formation is included. This approach is then used to calculate the average number of prompt neutrons emitted per fission, v-bar p . At high excitation energies, where fission is still possible after neutron emission, the consequences of the competition between first, second and third chance fission on N(E) and v-bar p are calculated. Excellent agreement with all the examples given in the original work of Madland and Nix is obtained. (author) [pt

  18. An investigation of fission models for high-energy radiation transport calculations

    International Nuclear Information System (INIS)

    Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

    1983-07-01

    An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

  19. MCNP6 Fission Cross Section Calculations at Intermediate and High Energies

    OpenAIRE

    Mashnik, Stepan G.; Sierk, Arnold J.; Prael, Richard E.

    2013-01-01

    MCNP6 has been Validated and Verified (V&V) against intermediate- and high-energy fission cross-section experimental data. An error in the calculation of fission cross sections of 181Ta and a few nearby target nuclei by the CEM03.03 event generator in MCNP6 and a "bug: in the calculation of fission cross sections with the GENXS option of MCNP6 while using the LAQGSM03.03 event generator were detected during our V&V work. After fixing both problems, we find that MCNP6 using CEM03.03 and LAQGSM...

  20. An assessment of fission product data for decay power calculation in fast reactors

    International Nuclear Information System (INIS)

    Sridharan, M.S.; Murthy, K.P.N.

    1987-01-01

    A review of our present capability at IGC, Kalpakkam to predict fission product decay power in fast reactors is presented. This is accomplished by comparing our summation calculations with the calculations of others and the reported experimental measurements. Our calculations are based on Chandy code developed at our Centre. The fission product data base of Chandy is essentially drawn from the yield data compiled by Crouch (1977) and the data on halflives etc. compiled by Tobias (1973). In general, we find good agreement amongst the different calculations (within ±5%) and our calculations also compare well with experimental measurements of AKIAMA et al and MURPHY et al

  1. Nuclear data requirements for fission reactor neutronics calculations

    International Nuclear Information System (INIS)

    Finck, P.

    1998-01-01

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data

  2. Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers

    International Nuclear Information System (INIS)

    Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.

    1989-01-01

    The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain

  3. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

  4. Fission product inventory calculation by a CASMO/ORIGEN coupling program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong; Jung, In Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14 x 14 PWR fuel assembly and the results are given in this paper. 3 refs., 1 fig., 1 tab. (Author)

  5. Optimal systematics of single-humped fission barriers for statistical calculations

    International Nuclear Information System (INIS)

    Mashnik, S.G.

    1993-01-01

    A systematic comparison of the existing phenomenological approaches and models for describing single-humped fast-computing fission barriers are given. The experimental data on excitation energy dependence of the fissility of compound nuclei are analyzed in the framework of the statistical approach by using different models for fission barriers, shell and pairing corrections and level-density parameter in order to identify their reliability and region of applicability for Monte Carlo calculations of evaporative cascades. The energy dependence of fission cross-sections for reactions induced by intermediate energy protons has been analyzed in the framework of the cascade-exiton model. 53 refs., 15 figs., 3 tabs

  6. 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

  7. Reliable method for fission source convergence of Monte Carlo criticality calculation with Wielandt's method

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori

    2004-01-01

    A new algorithm of Monte Carlo criticality calculations for implementing Wielandt's method, which is one of acceleration techniques for deterministic source iteration methods, is developed, and the algorithm can be successfully implemented into MCNP code. In this algorithm, part of fission neutrons emitted during random walk processes are tracked within the current cycle, and thus a fission source distribution used in the next cycle spread more widely. Applying this method intensifies a neutron interaction effect even in a loosely-coupled array where conventional Monte Carlo criticality methods have difficulties, and a converged fission source distribution can be obtained with fewer cycles. Computing time spent for one cycle, however, increases because of tracking fission neutrons within the current cycle, which eventually results in an increase of total computing time up to convergence. In addition, statistical fluctuations of a fission source distribution in a cycle are worsened by applying Wielandt's method to Monte Carlo criticality calculations. However, since a fission source convergence is attained with fewer source iterations, a reliable determination of convergence can easily be made even in a system with a slow convergence. This acceleration method is expected to contribute to prevention of incorrect Monte Carlo criticality calculations. (author)

  8. Structures and properties of (U,Pu)O{sub 2} containing non-active fission products. A simulation of irradiated nuclear fuel; Structure et proprietes de (U, Pu)O{sub 2} contenant des produits de fission sous forme inactive. Une simulation de combustible nucleaire irradie

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    We have made oxides with the same uranium and plutonium content, the same stoichiometry and the same fission product content as an oxide fuel (U{sub 0,8}PuO{sub 2})O{sub 1,96} after 2 per cent burn up. We have calculated the stoichiometry changes due to irradiation and checked the calculation by X rays parameters measurements. We have calculated and measured the contraction of the oxide lattice due to fission products in solid solution. Microprobe analysis of precipitates have been made and have lead to the identification of non metallic barium containing compounds and have shown the particular behaviour of molybdenum. Some physical properties have been measured especially the electrical resistivity, the thermal diffusivity and the vapour pressure of zirconium in solid solution. (author) [French] Nous avons fabrique des oxydes dont la composition en uranium et plutonium, la stoechiometrie et la teneur en produit de fission, sont identiques a celles d'un oxyde (U{sub 0,8}PuO{sub 2})O{sub 1,96} ayant subi 2 pour cent de combustion. Nous avons calcule les changements de stoechiometrie entraines par l'irradiation et controle ces calculs par des mesures de parametre. Nous avons calcule et mesure la contraction du reseau due aux produits de fissions solubles dans la matrice. Des analyses a la microsonde des precipites de produits de fission insolubles ont ete faites et ont conduit a l'identification de composes non metalliques contenant du baryum et a la mise en evidence du role particulier du molybdene. Certaines proprietes physiques ont ete mesurees sur ces composes, en particulier la resistivite electrique, la diffusivite thermique et la tension de vapeur du zirconium dissout dans la matrice. (auteur)

  9. On some uses of the recording op fission fragments occurring in insulators; Sur quelques utilisations de l'enregistrement des fragments de fission dans les isolants

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    The passage of fission fragments is manifested in insulators by the formation of damage lines which can easily be observed by transparence using the electron microscope after a suitable chemical attack. Since the recording efficiency is 100 per cent for mica and plastics this phenomenon has a certain number of applications. After briefly recalling the interaction processes between charged particles and matter, and giving a quantitative study of the relationships connecting the various parameters, the author considers here some of these applications: - thermal neutron dosimetry: it is possible to measure integrated fluxes of between lO{sup 3} and 10{sup 21} n/cm{sup 2}, according to the method used; - fast neutron dosimetry: by using fission fragment threshold sources it is easy to measure biological type doses of about 1 rem: - dosage of very low fissile impurity concentrations: atomic concentrations of about 10{sup -9}; can be measured; this limit has never been attained by conventional methods; - study of fissile elements occurring in atmospheric dusts. Results are then briefly given of an automated counting test for traces effected by measuring the electrical resistivity, of the irradiated membrane. Finally are given the advantages and disadvantages of these solid detectors, especially with respect to nuclear emulsions whose uses are approximately identical. (author) [French] Le passage des fragments de fission se materialise dans les isolants par des lignes de dommages que l'on peut facilement observer par transparence au microscope optique apres une attaque chimique appropriee. L'efficacite d'enregistrement etant de 100 pour cent dans le mica et les plastiques, ce phenomene peut avoir un certain nombre d'applications. Apres un bref rappel des processus d'interaction entre particules chargees et matiere, et une etude quantitative des relations unissant les differents parametres, on etudie ici quelques-unes de ces applications: - dosimetrie de neutrons

  10. Calculated leaching of certain fission products from a cylinder of French glass

    International Nuclear Information System (INIS)

    Blomqvist, G.

    1977-07-01

    The probable total leaching of the most important fission products and actinides have been tabulated for a cylinder of French HLW glass with approximately 9 percent fission products. The calculations cover the period between 30 and 10000 years after removal from the reactor. The cylinder is of the type planned for the introduction of the HLW into Swedish crystalline rocks. All the components are supposed to have the same leach rate. The calculations also include the probable thickness of eroded glass layer/year. (author)

  11. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-01-01

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

  12. 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)

  13. Calculational study for criticality safety data of fissionable actinides

    International Nuclear Information System (INIS)

    Nojiri, Ichiro; Fukasaku, Yasuhiro.

    1997-01-01

    This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

  14. First and second chance fission calculations for actinides and related topics

    International Nuclear Information System (INIS)

    Maino, G.; Menapace, E.; Motta, M.; Ventura, A.

    1980-01-01

    First and second chance contributions to neutron induced fission cross sections in an energy range of interest for reactor applications (E/sub n/less than or equal to 13 MeV) were obtained by extensive and consistent calculations for 241 Am; moreover, a simplified semiempirical approach was applied to 235 U and 239 Pu

  15. Inclusion of temperature dependence of fission barriers in statistical model calculations

    International Nuclear Information System (INIS)

    Newton, J.O.; Popescu, D.G.; Leigh, J.R.

    1990-08-01

    The temperature dependence of fission barriers has been interpolated from the results of recent theoretical calculations and included in the statistical model code PACE2. It is shown that the inclusion of temperature dependence causes significant changes to the values of the statistical model parameters deduced from fits to experimental data. 21 refs., 2 figs

  16. Fission barrier theory and its application to the calculation of actinide neutron cross-sections

    International Nuclear Information System (INIS)

    Lynn, J.E.

    1980-01-01

    The lectures discuss the possibilities and realisations of applying nuclear fission theory to the calculation of unknown nuclear data required for applications, principally in the nuclear power field. A brief description of the fundamentals of fission theory, the nature of the potential energy surface in the deformation plane, and of the inertial tensor, is given, and the accuracy of the theoretical calculations is discussed. It is concluded that it is impracticable to obtain required quantities such as neutron cross-sections from such fundamental calculations at present. On the other hand the fundamental theory reveals a wealth of phenomenological aspects of the fission process which can be incorporated into nuclear reaction theory. It is then shown how reaction theory thus extended to take correct account of the structured (''double-humped'') fission barrier can be used to parametrise the barrier by analysis of experimental data, and subsequently to calculate new data. Descriptions of computer programmes and illustrations of the application of the methods to actual physical examples are included in this account. (author)

  17. Study of a device for the direct measurement of the fission gas pressure inside an in-pile fuel element; Etude d'un dispositif pour la mesure directe de la pression des gaz de fission a l'interieur d'un element combustible en pile

    Energy Technology Data Exchange (ETDEWEB)

    Lavaud, B; Uschanoff, S [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The fission gas pressure inside a fuel element made of a refractory fuel constitutes an important limiting factor for the burn-up. Although it is possible to calculate approximately the volume of gas produced outside the fuel during its life-time; it is nevertheless very difficult to evaluate the pressure since the volume allowed to the fission gases, as well as their temperature are known only very approximately. This physical value, which is essential for the technologist, can only be known by direct in-pile measurement of the pressure. The report describes the equipment which has been developed for this test. (authors) [French] La pression des gaz de fission a l'interieur d'un element combustible a combustible refractaire constitue une des limitations importantes du taux de combustion. Si on peut approcher par calcul la determination du volume, des gaz degages hors du combustible au cours de sa vie, il est par contre tres difficile d'evaluer la pression car le volume alloue aux gaz de fission et leur temperature sont tres mal connus. Cette donnee essentielle pour le technologue ne peut etre atteinte que par une mesure directe en pile de la pression. Le rapport decrit l'appareillage qui a ete mis au point pour cet essai. (auteurs)

  18. Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors; Formulaire pour le calcul de la mecanique des empilements des reacteurs graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-07-01

    This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [French] Le domaine de ce formulaire est strictement limite aux effets mecaniques, pour les empilements, des deformations, thermiques ou autres, des structures metalliques de soutien (aire - support et corset). On propose un ensemble de relations qui ont ete etablies a la suite des essais de CHINON sur des maquettes de grande taille. Ces relations permettent le calcul des mouvements, des deformations et des contraintes dans les empilements du type EDF, a reseau horizontal triangulaire regulier. (auteurs)

  19. Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

    Science.gov (United States)

    Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

    2018-05-01

    Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  1. Fission gas pressure build-up and fast-breeder economy; Accumulation de la pression des gaz de fission et economie des reacteurs surgenerateurs a neutrons rapides; Nakoplenie davleniya gazov produktov deleniya i ehkonomika reaktorov-razmnozhitelej na bystrykh nejtronakh; Aumento de la presion de los gases de fision y economia de los reactores reproductores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Engelmann, P [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    Fuel-cycle costs and doubling time of fast-breeder reactors are strongly affected by the fuel-burn-up obtainable. Use of oxide or carbide fuel offers the possibility of reaching a burn-up of 100 000 MWd/t. In fuel-clad elements, a limiting factor is the fission-gas-pressure build-up. At the high burn-up considered, an appreciable fraction of the fission gases gets into the pores and thus contributes to the pressure on the can. Starting from the known fission-product yields and decay chains, gas production and pressure build-up have been calculated. Three physical models have been employed in calculating the pressure acting upon the can : the gas is contained either in interconnected pores, in separate pores, or in a central hole. The pressure-dependence upon free volume (fuel density) and temperature will be discussed. Cans made of high-strength materials as Ineonel-X and molybdenum could stand the fission-gas pressure at operating temperatures. Unfortunately, these materials have higher absorption cross-sections than stainless steel. Results of a multi-group calculation are given, showing the effect of using these can materials and of decreasing the fuel density on critical mass and breeding ratio in small and medium-size breeders. (author) [French] Le cout du cycle de combustible et la periode de doublement des reacteurs surgenerateurs a neutrons rapides dependent etroitement du taux de combustion. En utilisant pour combustible un oxyde ou un carbure, on peut atteindre un taux de combustion de 100 000 MW j/t. Avec des combustibles gaines, l'accumulation de la pression des gaz de fission est un facteur limitatif. Pour le fort taux de combustion envisage, une fraction non negligeable des gaz de fission penetre dans les interstices et contribue ainsi a la pression sur la gaine. A partir des rendements en produits de fission et des chaines de desintegration connus, l'auteur a calcule la production de gaz et l'accumulation de pression. Pour calculer la pression

  2. Burn-up calculation of fusion-fission hybrid reactor using thorium cycle

    International Nuclear Information System (INIS)

    Shido, S.; Matsunaka, M.; Kondo, K.; Murata, I.; Yamamoto, Y.

    2006-01-01

    A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated nuclear data. The 3-dimensional ITER model was used as a base fusion reactor. The nuclear fuel (reprocessed plutonium as the fission materials mixed with thorium as the fertile materials), transmutation materials (minor actinides and long-lived fission products) and tritium breeder were loaded into the blanket. Performances of gas-cooled and water-cooled blankets were compared with each other. As a result, the proposed reactor can meet the requirement for TBP and power density. As far as nuclear waste incineration is concerned, the gas-cooled blanket has advantages. On the other hand, the water cooled-blanket is suited to energy production. (author)

  3. Effect of fission yield libraries on the irradiated fuel composition in Monte Carlo depletion calculations

    International Nuclear Information System (INIS)

    Mitenkova, E.; Novikov, N.

    2014-01-01

    Improving the prediction of radiation parameters and reliability of fuel behaviour under different irradiation modes is particularly relevant for new fuel compositions, including recycled nuclear fuel. For fast reactors there is a strong dependence of nuclide accumulations on the nuclear data libraries. The effect of fission yield libraries on irradiated fuel is studied in MONTEBURNS-MCNP5-ORIGEN2 calculations of sodium fast reactors. Fission yield libraries are generated for sodium fast reactors with MOX fuel, using ENDF/B-VII.0, JEFF3.1, original library FY-Koldobsky, and GEFY 3.3 as sources. The transport libraries are generated from ENDF/B-VII.0 and JEFF-3.1. Analysis of irradiated MOX fuel using different fission yield libraries demonstrates the considerable spread in concentrations of fission products. The discrepancies in concentrations of inert gases being ∼25%, up to 5 times for stable and long-life nuclides, and up to 10 orders of magnitude for short-lived nuclides. (authors)

  4. Calculation of the fast multiplication factor by the fission matrix method

    International Nuclear Information System (INIS)

    Naumov, V.A.; Rozin, S.G.; Ehl'perin, T.I.

    1976-01-01

    A variation of the Monte Carlo method to calculate an effective breeding factor of a nuclear reactor is described. The evaluation procedure of reactivity perturbations by the Monte Carlo method in the first order perturbation theory is considered. The method consists in reducing an integral neutron transport equation to a set of linear algebraic equations. The coefficients of this set are elements of a fission matrix. The fission matrix being a Grin function of the neutron transport equation, is evaluated by the Monte Carlo method. In the program realizing the suggested algorithm, the game for initial neutron energy of a fission spectrum and then for the region of neutron birth, ΔVsub(f)sup(i)has been played in proportion to the product of Σsub(f)sup(i)ΔVsub(f)sup(i), where Σsub(f)sup(i) is a macroscopic cross section in the region numbered at the birth energy. Further iterations of a space distribution of neutrons in the system are performed by the generation method. In the adopted scheme of simulation of neutron histories the emission of secondary neutrons is controlled by weights; it occurs at every collision and not only in the end on the history. The breeding factor is calculated simultaneously with the space distribution of neutron worth in the system relative to the fission process and neutron flux. Efficiency of the described procedure has been tested on the calculation of the breeding factor for the Godiva assembly, simulating a fast reactor with a hard spectrum. A high accuracy of calculations at moderate number of zones in the core and reasonable statistics has been stated

  5. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  6. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  7. Uncertainty of decay heat calculations originating from errors in the nuclear data and the yields of individual fission products

    International Nuclear Information System (INIS)

    Rudstam, G.

    1979-01-01

    The calculation of the abundance pattern of the fission products with due account taken of feeding from the fission of 235 U, 238 U, and 239 Pu, from the decay of parent nuclei, from neutron capture, and from delayed-neutron emission is described. By means of the abundances and the average beta and gamma energies the decay heat in nuclear fuel is evaluated along with its error derived from the uncertainties of fission yields and nuclear properties of the inddividual fission products. (author)

  8. New results concerning the behaviour of fission gases in in-pile UO{sub 2} at high temperatures; Resultats nouveaux sur le comportement des gaz de fission a haute temperature dans l'UO{sub 2} en pile

    Energy Technology Data Exchange (ETDEWEB)

    Soulhier, R; Schurenkamper, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors consider in the first part the various phenomena giving rise to the production of fission gases towards the exterior of nuclear fuels. The following aspects are dealt with: diffusion, for which is considered the influence of the predecessors of the radioactive gases, the fission recoil, atom expulsion along the fission paths and the evaporation. In the second part the authors present the results obtained on UO{sub 2} samples subjected to irradiation at temperatures of between 150 and 2000 deg C: - At low temperatures the variation of the amount produced as a function of the half-life of the isotopes studied shows that recoil is hot the only cause of gas production. - Above 1800 deg C, a weight loss by evaporation has been observed and the influence of this phenomenon on gas liberation has been studied; thus the fraction of {sup 135}Xe liberated at 2000 deg C by processes other than evaporation is of the order of 10 per cent. - The influence of the various mechanism on the overall effect as a function of temperature is discussed. (authors) [French] Dans une premiere partie, les auteurs etudient les differents phenomenes pouvant donner lieu au degagement des gaz de fission hors d'un combustible. Sont traites successivement: la diffusion, pour laquelle on discute l'influence des predecesseurs des gaz radioactifs, le recul de fission, l'expulsion des atomes le long des trajets de fission et l'evaporation. Dans une deuxieme partie ils exposent les resultats obtenus sur des echantillons d'UO{sub 2} portes sous irradiation a des temperatures comprises entre 150 deg C et 2000 deg C: - A basse temperature la variation de la quantite degagee suivant la periode des isotopes etudies montre que le recul n'est pas la seule cause du degagement des gaz. - Au-dessus de 1800 deg C on a note une perte de poids par evaporation et on a evalue l'influence de ce phenomene sur la liberation des gaz: ainsi la fraction du {sup 135}Xe liberee a 2000 deg C par d'autres processus

  9. 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)

  10. Calculation of the Fission Product Release for the HTR-10 based on its Operation History

    International Nuclear Information System (INIS)

    Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

    2014-01-01

    Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

  11. Contribution to the study of the fission-gas release in metallic nuclear fuels; Contribution a l'etude du degagement des gaz de fission dans les combustibles nucleaires metalliques

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-10-01

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [French] Afin d'etudier l'effet d'une pression exterieure sur la limitation du gonflement due a la precipitation des gaz de fission, on a irradie a des taux de combustion d'environ 35.000 MWj/t et a des temperatures moyennes de 575 degres des echantillons d'uranium non allie et d'uranium-molybdene 8 pour cent contenus dans une gaine en acier inoxydable epaisse. Un trou cylindrique central permet au combustible de gonfler librement de 20 a 33 pour cent suivant les cas. Apres irradiation les echantillons d'uranium presentent deux types de ruptures de gaine: l'une due au gonflement du combustible, l'autre a la pression des gaz degages, ce degagement des gaz etant provoque par un reseau de micro-fissures. Les gaines des echantillons d'alliage uranium-molybdene sont toutes intactes et l'on montre que le relachement des gaz opere par interconnexion des bulles pour des valeurs de gonflement plus elevees que dans

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  13. Experiments on the incorporation of concentrated solutions of fission products in glasses and micas; Essais d'incorporation de solutions concentrees de produits de fission dans des verres et des micas

    Energy Technology Data Exchange (ETDEWEB)

    Bonniaud, R; Cohen, P; Sombret, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    1) The plants designed for extracting the plutonium of the G1, G2, G3 reactors fuel rods will produce extremely concentrated tions of fission products. 2) Let us consider a synthetic solution of the 'G2' type: (2N NO{sub 3}H - 2,25 meq/cm{sup 3} AI{sup +++} - 5.10{sup -3} {sup 134}Cs - {sup 137}Cs tracer). We made various glasses and micas by adding and mixing the necessary adjuvants and baking (900 to 1400 deg. C) in a graphite crucible. 3) The products obtained had either the shape of a cylindrical pellet or were reduced into a fine powder. They were mixed with 300 cm{sup 3} of synthetic sea water during variable periods of time in order to study leaching of the activity. 4) Experiments were first carried on caesium because of its solubility. 5) Fabrication of micas on a large scale sets many technological problems more difficult to solve (1400 deg. C) than in the case of glasses (1000 deg. C). A comparative study on both micas and glasses showed that leaching of activity was more important in the micas. (author)Fren. [French] 1) La mise en service de l'Usine d'Extraction du Plutonium destinee en particulier au traitement des barres de combustibles de G1, G2, G3 fournira des solutions residuaires concentrees de produits de fission. 2) Partant d'une solution synthetique 'G2': (2N NO{sub 3}H - 2,25meq/cm{sup 3} AI{sup +++} - 5.10{sup -3} {sup 134}Cs - {sup 137}Cs traceur) nous avens realise differents verres et micas apres adduction et melange des adjuvants necessaires suivis de cuisson (900 a 1400 deg. C) dans un creuset de graphite. 3) Les produits obtenus ont ete, soit sous forme d'eprouvette cylindrique, soit sous forme de poudre finement broyee, mis en contact et agites avec 300 cm{sup 3} d'eau de mer synthetique, durant des temps varies pour etudier la retention de l'activite. 4) Le cesium a ete choisi pour ces essais a cause de sa solubilite. 5) La fabrication de micas a grande echelle pose des problemes technologiques plus difficiles (1400 deg. C) a

  14. Calculation of energy transfer by fission fragments from plane uranium layer to thin wire

    International Nuclear Information System (INIS)

    Pikulev, A.A.

    2006-01-01

    Energy transfer from a flat fissile uranium slab to a fine wire via fission fragments is calculated. The rate of energy transfer versus the thicknesses of the slab and protecting aluminum film, as well as the wire-slab gap, is found. An expression for the absorption coefficient of the wire is derived, and the effect the thickness of the wire has on the energy transfer process is studied. The amount of the edge effect for a finite-size uranium slab is demonstrated with calculations for vacuum conditions and for argon under a pressure of 0.25 atm [ru

  15. Implementation of a Thermodynamic Solver within a Computer Program for Calculating Fission-Product Release Fractions

    Science.gov (United States)

    Barber, Duncan Henry

    During some postulated accidents at nuclear power stations, fuel cooling may be impaired. In such cases, the fuel heats up and the subsequent increased fission-gas release from the fuel to the gap may result in fuel sheath failure. After fuel sheath failure, the barrier between the coolant and the fuel pellets is lost or impaired, gases and vapours from the fuel-to-sheath gap and other open voids in the fuel pellets can be vented. Gases and steam from the coolant can enter the broken fuel sheath and interact with the fuel pellet surfaces and the fission-product inclusion on the fuel surface (including material at the surface of the fuel matrix). The chemistry of this interaction is an important mechanism to model in order to assess fission-product releases from fuel. Starting in 1995, the computer program SOURCE 2.0 was developed by the Canadian nuclear industry to model fission-product release from fuel during such accidents. SOURCE 2.0 has employed an early thermochemical model of irradiated uranium dioxide fuel developed at the Royal Military College of Canada. To overcome the limitations of computers of that time, the implementation of the RMC model employed lookup tables to pre-calculated equilibrium conditions. In the intervening years, the RMC model has been improved, the power of computers has increased significantly, and thermodynamic subroutine libraries have become available. This thesis is the result of extensive work based on these three factors. A prototype computer program (referred to as SC11) has been developed that uses a thermodynamic subroutine library to calculate thermodynamic equilibria using Gibbs energy minimization. The Gibbs energy minimization requires the system temperature (T) and pressure (P), and the inventory of chemical elements (n) in the system. In order to calculate the inventory of chemical elements in the fuel, the list of nuclides and nuclear isomers modelled in SC11 had to be expanded from the list used by SOURCE 2.0. A

  16. Development and application of the PBMR fission product release calculation model

    International Nuclear Information System (INIS)

    Merwe, J.J. van der; Clifford, I.

    2008-01-01

    At PBMR, long-lived fission product release from spherical fuel spheres is calculated using the German legacy software product GETTER. GETTER is a good tool when performing calculations for fuel spheres under controlled operating conditions, including irradiation tests and post-irradiation heat-up experiments. It has proved itself as a versatile reactor analysis tool, but is rather cumbersome when used for accident and sensitivity analysis. Developments in depressurized loss of forced cooling (DLOFC) accident analysis using GETTER led to the creation of FIssion Product RElease under accident (X) conditions (FIPREX), and later FIPREX-GETTER. FIPREX-GETTER is designed as a wrapper around GETTER so that calculations can be carried out for large numbers of fuel spheres with design and operating parameters that can be stochastically varied. This allows full Monte Carlo sensitivity analyses to be performed for representative cores containing many fuel spheres. The development process and application of FIPREX-GETTER in reactor analysis at PBMR is explained and the requirements for future developments of the code are discussed. Results are presented for a sample PBMR core design under normal operating conditions as well as a suite of design-base accident events, illustrating the functionality of FIPREX-GETTER. Monte Carlo sensitivity analysis principles are explained and presented for each calculation type. The plan and current status of verification and validation (V and V) is described. This is an important and necessary process for all software and calculation model development at PBMR

  17. Methodes de calcul des forces aerodynamiques pour les etudes des interactions aeroservoelastiques

    Science.gov (United States)

    Biskri, Djallel Eddine

    L'aeroservoelasticite est un domaine ou interagissent la structure flexible d'un avion, l'aerodynamique et la commande de vol. De son cote, la commande du vol considere l'avion comme une structure rigide et etudie l'influence du systeme de commande sur la dynamique de vol. Dans cette these, nous avons code trois nouvelles methodes d'approximation de forces aerodynamiques: Moindres carres corriges, Etat minimal corrige et Etats combines. Dans les deux premieres methodes, les erreurs d'approximation entre les forces aerodynamiques approximees par les methodes classiques et celles obtenues par les nouvelles methodes ont les memes formes analytiques que celles des forces aerodynamiques calculees par LS ou MS. Quant a la troisieme methode, celle-ci combine les formulations des forces approximees avec les methodes standards LS et MS. Les vitesses et frequences de battement et les temps d'executions calcules par les nouvelles methodes versus ceux calcules par les methodes classiques ont ete analyses.

  18. Multi-dimensional fission-barrier calculations from Se to the SHE; from the proton to the neutron drip lines

    International Nuclear Information System (INIS)

    Moeller, Peter; Sierk, Arnold J.; Bengtsson, Ragnar; Iwamoto, Akira

    2003-01-01

    We present fission-barrier-height calculations for nuclei throughout the periodic system based on a realistic theoretical model of the multi-dimensional potential-energy surface of a fissioning nucleus. This surface guides the nuclear shape evolution from the ground state, over inner and outer saddle points, to the final configurations of separated fission fragments. We have previously shown that our macroscopic-microscopic nuclear potential-energy model yields calculated 'outer' fission-barrier heights (E B ) for even-even nuclei throughout the periodic system that agree with experimental data to within about 1.0 MeV. We present final results of this work. Just recently we have enhanced our macroscopic-microscopic nuclear potential-energy model to also allow the consideration of axially asymmetric shapes. This shape degree of freedom has a substantial effect on the calculated height (E A ) of the inner peak of some actinide fission barriers. We present examples of fission-barrier calculations by use of this model with its redetermined constants. Finally we discuss what the model now tells us about fission barriers at the end of the r-process nucleosynthesis path. (author)

  19. Calculation of high-dimensional fission-fusion potential-energy surfaces in the SHE region

    International Nuclear Information System (INIS)

    Moeller, Peter; Sierk, Arnold J.; Ichikawa, Takatoshi; Iwamoto, Akira

    2004-01-01

    We calculate in a macroscopic-microscopic model fission-fusion potential-energy surfaces relevant to the analysis of heavy-ion reactions employed to form heavy-element evaporation residues. We study these multidimensional potential-energy surfaces both inside and outside the touching point.Inside the point of contact we define the potential on a multi-million-point grid in 5D deformation space where elongation, merging projectile and target spheroidal shapes, neck radius and projectile/target mass asymmetry are independent shape variables. The same deformation space and the corresponding potential-energy surface also describe the shape evolution from the nuclear ground-state to separating fragments in fission, and the fast-fission trajectories in incomplete fusion.For separated nuclei we study the macroscopic-microscopic potential energy, that is the ''collision surface'' between a spheroidally deformed target and a spheroidally deformed projectile as a function of three coordinates which are: the relative location of the projectile center-of-mass with respect to the target center-of-mass and the spheroidal deformations of the target and the projectile. We limit our study to the most favorable relative positions of target and projectile, namely that the symmetry axes of the target and projectile are collinear

  20. Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS

    International Nuclear Information System (INIS)

    Perkasa, Y. S.; Waris, A.; Kurniadi, R.; Su'ud, Z.

    2014-01-01

    Comparative studies of actinide and sub-actinide fission cross section calculation from MCNP6 and TALYS have been conducted. In this work, fission cross section resulted from MCNP6 prediction will be compared with result from TALYS calculation. MCNP6 with its event generator CEM03.03 and LAQGSM03.03 have been validated and verified for several intermediate and heavy nuclides fission reaction data and also has a good agreement with experimental data for fission reaction that induced by photons, pions, and nucleons at energy from several ten of MeV to about 1 TeV. The calculation that induced within TALYS will be focused mainly to several hundred MeV for actinide and sub-actinide nuclides and will be compared with MCNP6 code and several experimental data from other evaluator

  1. Calculation for fission decay from heavy ion reactions at intermediate energies

    International Nuclear Information System (INIS)

    Blaich, T.; Begemann-Blaich, M.; Fowler, M.M.; Wilhelmy, J.B.; Britt, H.C.; Fields, D.J.; Hansen, L.F.; Namboodiri, M.N.; Sangster, T.C.; Fraenkel, Z.

    1992-01-01

    A detailed deexcitation calculation is presented for target residues resulting from intermediate-energy heavy ion reactions. The model involves an intranuclear cascade, subsequent fast nucleon emission, and final decay by statistical evaporation including fission. Results are compared to data from bombardments with Fe and Nb projectiles on targets of Ta, Au, and Th at 100 MeV/nucleon. The majority of observable features are reproduced with this simple approach, making obvious the need for involving new physical phenomena associated with multifragmentation or other collective dissipation mechanisms

  2. LENMODEL: A forward model for calculating length distributions and fission-track ages in apatite

    Science.gov (United States)

    Crowley, Kevin D.

    1993-05-01

    The program LENMODEL is a forward model for annealing of fission tracks in apatite. It provides estimates of the track-length distribution, fission-track age, and areal track density for any user-supplied thermal history. The program approximates the thermal history, in which temperature is represented as a continuous function of time, by a series of isothermal steps of various durations. Equations describing the production of tracks as a function of time and annealing of tracks as a function of time and temperature are solved for each step. The step calculations are summed to obtain estimates for the entire thermal history. Computational efficiency is maximized by performing the step calculations backwards in model time. The program incorporates an intuitive and easy-to-use graphical interface. Thermal history is input to the program using a mouse. Model options are specified by selecting context-sensitive commands from a bar menu. The program allows for considerable selection of equations and parameters used in the calculations. The program was written for PC-compatible computers running DOS TM 3.0 and above (and Windows TM 3.0 or above) with VGA or SVGA graphics and a Microsoft TM-compatible mouse. Single copies of a runtime version of the program are available from the author by written request as explained in the last section of this paper.

  3. The present state of research on the vitrification of concentrated solutions of fission products (1962); Etat des etudes sur la vitrification des solutions concentrees de produits de fission (1962)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    The present report gives the actual point of studies on vitrification of concentrated solutions of fission products. An active cell, giving glasses in crucibles, permitted to study various glass compositions. The leaching rate from the glass raises 1 to 2 10{sup -7} g of glass/cm{sup 2}/day. Activity loss by volatility during vitrification remains weak and often below 0.1 per cent of total activity. Off gas cleaning is made easier by presence of filter which is compound of granules including iron oxide. After saturation the content of this filter can be melt. Moreover different processes are in experimentation for a more important production. Daily 72 liters of solution containing tracer activity are treated in a continuous calcination and vitrification plant. The loss in {sup 106}Ru is still important and a modification of installation has been necessary. A pot vitrification plant is in study. In order to reduce cost of processing the possibility to pour glass after melting is actuality in study. A production set of very active glass is also in project. (authors) [French] Le present rapport fait le point des etudes menees sur la vitrification des solutions concentrees de produits de fission. Une installation active, produisant des verres en creusets, a permis d'etudier plusieurs compositions de verres. Le taux de perte d'activite par lixiviation a l'eau atteint 1 a 2 10{sup -7} gramme de verre/cm{sup 2}/jour. Les pertes d'activite par volatilite au cours de la cuisson restent faibles et souvent inferieures a 0,1 pour cent de l'activite totale. L'epuration des gaz de cuisson est facilitee par la presence d'un filtre a granules riches en oxyde de fer, dont le contenu peut etre fondu apres saturation. Differentes techniques sont, en outre, en experimentation pour une production plus importante: Une installation de calcination et vitrification continue traite 72 litres par jour de solution contenant une activite traceur. La perte en Ru{sup 106} est encore importante

  4. Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

    International Nuclear Information System (INIS)

    Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

    2007-01-01

    The purpose of the present work is to develop an efficient solution method for the calculation of neutron importance function in fissionable assemblies for all criticality conditions, based on Monte Carlo calculations. The neutron importance function has an important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating the adjoint flux while solving the adjoint weighted transport equation based on deterministic methods. However, in complex geometries these calculations are very complicated. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on the physical concept of neutron importance has been introduced for calculating the neutron importance function in sub-critical, critical and super-critical conditions. For this propose a computer program has been developed. The results of the method have been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries. The correctness of these results has been confirmed for all three criticality conditions. Finally, the efficiency of the method for complex geometries has been shown by the calculation of neutron importance in Miniature Neutron Source Reactor (MNSR) research reactor

  5. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

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

  7. Study of the isomer nuclei produced in the spontaneous fission of {sup 252}Cf; Etude des noyaux isomeriques produits dans la fission spontanee de {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Gautherin, C.; Houry, M.; Korten, W.; Le Coz, Y.; Lucas, R.; Thiesen, Ch. [Dept. d`Astrophysique, de la Physique des Particules, de la Physique Nucleaire et de l`Instrumentation Associee, CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France); Badimon, C.; Barreau, G.; Doan, T.P.; Pedemay, G. [Centre d`Etudes Nucleaires, Bordeaux-1 Univ., 33 Gradignan (France); Belier, G.; Girod, M.; Meot, M.V. [CEA Centre d`Etudes de Bruyeres-le-Chatel, 91 (France); Astier, G.; Meot, V.; Peru, S. [CEA Centre d`Etudes de Bruyeres-le-Chatel, 91 (France); Astier, A.; Ducroux, L.; Meyer, M.; Redon, N. [Inst.de Physique Nucleaire, Lyon-1 Univ., 69 - Villeurbanne (France)

    1997-06-01

    Isomeric states have been studied in fission fragments produced by spontaneous fission of {sup 252}Cf. 34 isomeric nuclei have been identified by using coincidences between {gamma}-rays detected in EUROGAM II and fission fragments detected in photovoltaic cells. Lifetimes from 20 ns up to 2 {mu}s have been measured. Microscopic interpretation of the isomeric levels discovered has been tried by means of the Hartree-Fock-Bogolyubov procedure using Gogny force. It was found that the {sup 152,154,156}Nd nuclei have prolate deformation in their ground state; the rotational band built on this ground state is well reproduced by the calculation. For these nuclei the 2 quasi-particle excited states energies are above 1 MeV in agreement with the experiment. The computation confirms the similitude of the {sup 156}Nd and {sup 158}Sm isomeric states associated to neutron 2 quasi-particles of J{sup {pi}} K{sup {pi}} = 5{sup -} ({nu} 5/2 (642) x {nu} 5/2 (523)) while such similitude does not occur for the isotone nuclei {sup 154}Nd and {sup 156}Sm. The computation predicts a proton 2 quasi-particle excited states of J{sup {pi}} = 5{sup -}, near the isomeric level measured in {sup 156}Sm but not for that of {sup 154}Nd. Concerning the {sup 152}Nd the calculated level density is very near that of the states measured in the 1.6 to 2.3 MeV interval. However, the lack of precise information on spins and parities of the measured levels does make not possible a confrontation with the calculations. An up-graded equipment implying 32 photovoltaic cells instead of 2 cells is to be developed and installed by the EUROGAM 3 refs.

  8. Investigation of the fission yields of the fast neutron-induced fission of {sup 233}U; Mesure de la distribution en masse et en charge des produits de la fission rapide de l'{sup 233}U

    Energy Technology Data Exchange (ETDEWEB)

    Galy, J

    1999-09-01

    As a stars, a survey of the different methods of investigations of the fission product yields and the experimental data status have been studied, showing advantages and shortcomings for the different approaches. An overview of the existing models for the fission product distributions has been as well intended. The main part of this thesis was the measurement of the independent yields of the fast neutron-induced fission of{sup 233}U, never investigated before this work. The experiment has been carried out using the mass separator OSIRIS (Isotope Separator On-Line). Its integrated ion-source and its specific properties required an analysis of the delay-parameter and ionisation efficiency for each chemical species. On the other hand, this technique allows measurement of independent yields and cumulative yields for elements from Cu to Ba, covering most of the fission yield distribution. Thus, we measured about 180 independent yields from Zn (Z=30) to Sr (Z=38) in the mass range A=74-99 and from Pd (Z=46) to Ba (Z=56) in the mass range A=113-147, including many isomeric states. An additional experiment using direct {gamma}-spectroscopy of aggregates of fission products was used to determine more than 50 cumulative yields of element with half-life from 15 min to a several days. All experimental data have been compared to estimates from a semi-empirical model, to calculated values and to evaluated values from the European library JEF 2.2. Furthermore, a study of both thermal and fast neutron-induced fission of {sup 233}U measured at Studsvik, the comparison of the OSIRIS and LOHENGRIN facilities and the trends in new data for the Reactors Physics have been discussed. (author)

  9. Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

    International Nuclear Information System (INIS)

    Feghhi, S. A. H.; Afarideh, H.; Shahriari, M.

    2007-01-01

    The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor

  10. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  11. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  12. TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

    International Nuclear Information System (INIS)

    Smith, P.D.

    1978-02-01

    A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique

  13. 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)

  14. 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

  15. Macroscopic calculational model of fission gas release from water reactor fuels

    International Nuclear Information System (INIS)

    Uchida, Masaki

    1993-01-01

    Existing models for estimating fission gas release rate usually have fuel temperature as independent variable. Use of fuel temperature, however, often brings an excess ambiguity in the estimation because it is not a rigorously definable quantity as a function of heat generation rate and burnup. To derive a mathematical model that gives gas release rate explicitly as a function of design and operational parameters, the Booth-type diffusional model was modified by changing the character of the diffusion constant from physically meaningful quantity into a mere mathematical parameter, and also changing its temperature dependency into power dependency. The derived formula was found, by proper choice of arbitrary constants, to satisfactorily predict the release rates under a variety of irradiation histories up to a burnup of 60,000 MWd/t. For simple power histories, the equation can be solved analytically by defining several transcendental functions, which enables simple calculation of release rate using graphs. (author)

  16. Cluster expression in fission and fusion in high-dimensional macroscopic-microscopic calculations

    International Nuclear Information System (INIS)

    Iwamoto, Akira; Ichikawa, Takatoshi; Moller, Peter; Sierk, Arnold J.

    2004-01-01

    We discuss the relation between the fission-fusion potential-energy surfaces of very heavy nuclei and the formation process of these nuclei in cold-fusion reactions. In the potential-energy surfaces, we find a pronounced valley structure, with one valley corresponding to the cold-fusion reaction, the other to fission. As the touching point is approached in the cold-fusion entrance channel, an instability towards dynamical deformation of the projectile occurs, which enhances the fusion cross section. These two 'cluster effects' enhance the production of superheavy nuclei in cold-fusion reactions, in addition to the effect of the low compound-system excitation energy in these reactions. Heavy-ion fusion reactions have been used extensively to synthesize heavy elements beyond actinide nuclei. In order to proceed further in this direction, we need to understand the formation process more precisely, not just the decay process. The dynamics of the formation process are considerably more complex than the dynamics necessary to interpret the spontaneous-fission decay of heavy elements. However, before implementing a full dynamical description it is useful to understand the basic properties of the potential-energy landscape encountered in the initial stages of the collision. The collision process and entrance-channel landscape can conveniently be separated into two parts, namely the early-stage separated system before touching and the late-stage composite system after touching. The transition between these two stages is particularly important, but not very well understood until now. To understand better the transition between the two stages we analyze here in detail the potential energy landscape or 'collision surface' of the system both outside and inside the touching configuration of the target and projectile. In Sec. 2, we discuss calculated five-dimensional potential-energy landscapes inside touching and identify major features. In Sec. 3, we present calculated

  17. Fission observables from 4D Langevin calculations with macroscopic transport coefficients

    Directory of Open Access Journals (Sweden)

    Usang Mark D.

    2018-01-01

    Full Text Available We have extended the Langevin equations to 4 dimensions (4D by allowing the independent deformation for the left (δ1 and right fragments (δ2 of the fissioning nucleus. At the moment we are only able to use them in conjunction with the macroscopic transport coefficients. Nevertheless, we can see a considerable improvement in the preliminary results for the fission observables, especially those related to the total kinetic energy (TKE of fission fragments. By plotting the TKE distributions we have revealed the super-long fission modes in 236U and super-short fission modes in 257Fm. By plotting the distribution of δ against the fragment’s TKE we have noted a correlation between the values of δ and Brosa’s fission modes. We have found that the standard fission modes correspond to prolate tips of the light fragments while the complementary heavy fragments have oblate fission tips. On the other hand, if both fragments were prolate at the tips, we get super-long fission modes. If both fragments were oblate at the tips, we get super-short fission modes.

  18. Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

    International Nuclear Information System (INIS)

    Okuno, H.

    2002-01-01

    We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

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

  20. Variational RRKM calculation of thermal rate constant for C–H bond fission reaction of nitro methane

    Directory of Open Access Journals (Sweden)

    Afshin Taghva Manesh

    2017-02-01

    Full Text Available The present work provides quantitative results for the rate constants of unimolecular C–H bond fission reactions in the nitro methane at elevated temperatures up to 2000 K. In fact, there are three different hydrogen atoms in the nitro methane. The potential energy surface for each C–H bond fission reaction of nitro methane was investigated by ab initio calculations. The geometry and vibrational frequencies of the species involved in this process were optimized at the MP2 level of theory, using the cc-pvdz basis set. Since C–H bond fission channel is a barrierless reaction, we have used variational RRKM theory to predict rate coefficients. By means of calculated rate coefficients at different temperatures, the Arrhenius expression of the channel over the temperature range of 100–2000 K is k(T = 5.9E19∗exp(−56274.6/T.

  1. Impact of the Heat Transfer on the Performance Calculations of Automotive Turbocharger Compressor Influence des transferts thermiques sur le calcul des performances des compresseurs de suralimentation

    Directory of Open Access Journals (Sweden)

    Chesse P.

    2011-09-01

    Full Text Available Usually, turbochargers used within internal combustion engine simulation software are modelled in an adiabatic manner. However, during our experimental tests we found that this is not necessarily the case. The direct use of the manufacturer’s map is not possible anymore. A simple method which considers the heat transfers is proposed. It is based on experimental tests made on hot air supplied turbocharger test bench. The difference with the adiabatic model is considerable mainly for low compressor power. This corresponds to internal combustion engine low loads. En général, les turbocompresseurs pris en compte dans les logiciels de simulation moteur sont modélisés de façon adiabatique. Cependant, les tests expérimentaux effectués au laboratoire montrent que ce n’est pas toujours le cas. L’utilisation directe des champs de fonctionnement fournis par les constructeurs de turbomachines n’est alors plus possible. Une évaluation quantitative de ces transferts, basée sur des tests réalisés sur un banc d’essais turbo à air chaud, est présentée. Puis ils sont pris en compte afin de calculer les caractéristiques réelles de fonctionnement d’un compresseur. La différence avec le modèle adiabatique apparaît très importante pour les faibles puissances compresseur. Ceci correspond aux faibles charges moteur.

  2. Statistical model calculations with a double-humped fission barrier GIVAB computer code

    International Nuclear Information System (INIS)

    Delagrange, H.; Gilat, J.

    1979-01-01

    Neutron and gamma emission probabilities and fission probabilities are computed, taking into account the special feature of the actinide fission barriers with two maxima. Spectra and cross sections are directly deduced from these probabilities. Populations of both wells are followed step by step. For each initial E and J, decay rates are computed and normalized in order to obtain the de-excitation probabilities imposed by the two-humped fission barrier

  3. The statistical model calculation of prompt neutron spectra from spontaneous fission of {sup 244}Cm and {sup 246}Cm

    Energy Technology Data Exchange (ETDEWEB)

    Gerasimenko, B.F. [V.G. Khlopin Radium Inst., Saint Peterburg (Russian Federation)

    1997-03-01

    The calculations of integral spectra of prompt neutrons of spontaneous fission of {sup 244}Cm and {sup 246}Cm were carried out. The calculations were done by the Statistical Computer Code Complex SCOFIN applying the Hauser-Feschbach method as applied to the description of the de-excitation of excited fission fragments by means of neutron emission. The emission of dipole gamma-quanta from these fragments was considered as a competing process. The average excitation energy of a fragment was calculated by two-spheroidal model of tangent fragments. The density of levels in an excited fragment was calculated by the Fermi-gas model. The quite satisfactory agreement was reached between theoretical and experimental results obtained in frames of Project measurements. The calculated values of average multiplicities of neutron number were 2,746 for {sup 244}Cm and 2,927 for {sup 246}Cm that was in a good accordance with published experimental figures. (author)

  4. Monte Carlo code Serpent calculation of the parameters of the stationary nuclear fission wave

    Directory of Open Access Journals (Sweden)

    V. M. Khotyayintsev

    2017-12-01

    Full Text Available n this work, propagation of the stationary nuclear fission wave was simulated for series of fixed power values using Monte Carlo code Serpent. The wave moved in the axial direction in 5 m long cylindrical core of fast reactor with pure 238U raw fuel. Stationary wave mode arises some period later after the wave ignition and lasts sufficiently long to determine kef with high enough accuracy. The velocity characteristic of the reactor was determined as the dependence of the wave velocity on the neutron multiplication factor. As we have recently shown within a one-group diffusion description, the velocity characteristic is two-valued due to the effect of concentration mechanisms, while thermal feedback affects it only quantitatively. The shape and parameters of the velocity characteristic critically affect feasibility of the reactor design since stationary wave solutions of the lower branch are unstable and do not correspond to any real waves in self-regulated reactor, like CANDLE. In this work calculations were performed without taking into account thermal feedback. They confirm that theoretical dependence correctly describes the shape of the velocity characteristic calculated using the results of the Serpent modeling.

  5. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  6. Calculation of the nuclear fission data based on the framework of the QMD + SDM

    International Nuclear Information System (INIS)

    Rong Jian; Iwamoto, O.; Fukahori, T.

    2002-01-01

    The quantum molecular dynamics (QMD), statistical decay model (SDM) and the statistical fission theory were used to analyze the mass distribution of the fission products, the prompt fission neutron spectrum (x(E)) and the prompt fission neutron multiplicities (ν-bar pf (E)) caused by the intermediate energy nucleon-induced fission. The semi-empirical formula of energy level density parameter used in the statistical process was also studied. Very few adjustable parameters were included in the present method. By some physical analysis, it can be thought that the present results are reasonable. The x(E) and ν-bar pf (E) can be obtained in the intermediate energy region by the present method

  7. Calculation of the distribution of the escaping from a fissionable sample neutrons number when introducing one fission neutron in that sample

    International Nuclear Information System (INIS)

    Dorlet, J.

    1991-01-01

    A describing algorithm furnishes the probabilities of having exactly N escaping neutrons in the descent of one fission neutron, using the punctual reactor model. Calculations can be performed even for N-values greater than 1000. Numerical results show that discrete neutrons counting is unfit to obtain the N mean value, even for very far subcritical devices. That mean value is still very used in the existing theoretical studies, because its obvious correlation with the device effective multiplication coefficient. For that reason modelling coincidence neutrons counting is not suitable using statistical moments approach, but only using probabilities it selves [fr

  8. Calculation of dose distribution for 252Cf fission neutron source in tissue equivalent phantoms using Monte Carlo method

    International Nuclear Information System (INIS)

    Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong

    2001-01-01

    Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  10. Measurement of the {sup 240}Pu concentration of a plutonium sample by the spontaneous fission method (1960); Mesure de la teneur d'un plutonium en {sup 240}Pu par la methode des fissions spontanees (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Caizergues, R; Clouet d' Orval, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    It is very important to know the plutonium-240 content of the plutonium formed in piles. The method of measurement described here consists in counting the number of spontaneous fissions produced in a known quantity of plutonium. This measurement is carried out in a multiple-plate ionisation chamber, the plutonium being deposited on the plates. The disintegration constant of plutonium-240 by spontaneous fission being known the plutonium-240 content in the sample can be calculated. (author) [French] La connaissance de la teneur en plutonium-240 du plutonium forme dans les piles est une question importante. La methode de mesure presentee ici consiste a compter le nombre de fissions spontanees issues d'une quantite connue de plutonium. Cette mesure est effectuee dans une chambre d'ionisation a plateaux multiples sur lesquels est depose le plutonium. La connaissance de la constante de desintegration du plutonium-240 par fissions spontanees permet de calculer la teneur en plutonium-240. (auteur)

  11. Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

    The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

  12. Shielding calculation of a hot cell for the processing of fission products

    International Nuclear Information System (INIS)

    Rocha, A.C.S. da; Pina, J.L.S. de; Silva, J.J.G. da.

    1986-12-01

    A dose rate estimation is made for an operator of a lead wall, fission products processing hot cell, in a distance of 50 cm from the emission source, at Brazilian Institute of Nuclear Engineering (IEN). (L.C.J.A.)

  13. Consultants’ Meeting on Recommended Input Parameters for Fission Cross-Section Calculations. Summary Report

    International Nuclear Information System (INIS)

    Capote Noy, Roberto; Simakov, Stanislav; Goriely, Stephane; Hilaire, Stephane; Iwamoto, Osamu; Kawano, Toshihiko; Koning, Arjan

    2014-12-01

    A Consultants’ Meeting on “Recommended Input Parameters for Fission Cross-Section Calculations” was held at IAEA Headquarters, Vienna, Austria to define the scope, deliverables and appropriate work programme of a possible Coordinated Research Project (CRP) on the subject. Presentations are available online at https://www-nds.iaea.org/indexmeeting-crp/CM-RIPL-fission/. A new CRP was endorsed to recommend a comprehensive set of fission input parameters needed for the modelling of fission cross sections. Special attention will be given to the modelling of photon and nucleon induced reactions on actinides with emphasis on incident energies below 30 MeV. The goals and detailed deliverables of the planned CRP were proposed. A Hauser-Feshbach code intercomparison was recommended. (author)

  14. LOFC fission product release and circulating activity calculations for gas-cooled reactors

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

    1977-01-01

    The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

  15. Application of Higher Order Fission Matrix for Real Variance Estimation in McCARD Monte Carlo Eigenvalue Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2015-05-15

    In a Monte Carlo (MC) eigenvalue calculation, it is well known that the apparent variance of a local tally such as pin power differs from the real variance considerably. The MC method in eigenvalue calculations uses a power iteration method. In the power iteration method, the fission matrix (FM) and fission source density (FSD) are used as the operator and the solution. The FM is useful to estimate a variance and covariance because the FM can be calculated by a few cycle calculations even at inactive cycle. Recently, S. Carney have implemented the higher order fission matrix (HOFM) capabilities into the MCNP6 MC code in order to apply to extend the perturbation theory to second order. In this study, the HOFM capability by the Hotelling deflation method was implemented into McCARD and used to predict the behavior of a real and apparent SD ratio. In the simple 1D slab problems, the Endo's theoretical model predicts well the real to apparent SD ratio. It was noted that the Endo's theoretical model with the McCARD higher mode FS solutions by the HOFM yields much better the real to apparent SD ratio than that with the analytic solutions. In the near future, the application for a high dominance ratio problem such as BEAVRS benchmark will be conducted.

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

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R.

    1963-05-15

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

  17. Stablité des Eléments Trans-ferminums à Haut Spin : Mesure de la barrière de fission de 254No

    OpenAIRE

    Henning , Gregoire

    2012-01-01

    Super heavy nuclei provide opportunities to study nuclear structure near three simultaneous limits: in charge Z, spin I and excitation energy E∗. These nuclei exist only because of a fission barrier, created by shell effects. It is therefore important to determine the fission barrier and its spin dependence Bf(I), which gives information on the shell energy Eshell(I). Theoretical calculations predict different fission barrier heights from Bf(I = 0) = 6.8 MeV for a macro-microscopic model to 8...

  18. Statistical model calculation of fission isomer excitation functions in (n,n') and (n,γ) reactions

    International Nuclear Information System (INIS)

    Chatterjee, A.; Athougies, A.L.; Mehta, M.K.

    1977-01-01

    A statistical model developed by Britt and others (1971, 1973) to analyze isomer excitation functions in spallation type reactions like (α,2n) has been adopted in fission isomer calculations for (n,n') and (n,γ) reactions. Calculations done for 235 U(n,n')sup(238m)U and 235 U(n,γ)sup(236m)U reactions have been compared with experimental measurements. A listing of the computer program ISOMER using FORTRAN IV to calculate the isomer to prompt ratios is given. (M.G.B.)

  19. Study of fission product {gamma} spectra in the band 2-500 keV; Etude du spectre {gamma} des produits de fission dans la bande 0-500 keV

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    In the study of the {gamma} spectrum of uranium fission products, particular attention has been given in this note to the part of the spectrum ranging between 0 and 500 keV after a given pile operating programme and the evolution of this spectrum with time after a pile shutdown has been followed. The study be related to the fission products which appear in the pile as a whole or on those produced in a uranium sample assumed to have been placed in the pile. The latter case has been envisaged here. The spectrum determination is based partly on theory and partly on experiment. The pile operating conditions are different in the two cases, which widens the range of validity of the spectra traced here. (author) [French] Dans l'etude du spectre {gamma} des produits de fission de l'uranium, on s'est plus particulierement interesse dans la presente note a determiner la partie du spectre qui s'etend entre 0 et 500 keV, au bout d'un fonctionnement donne de pile, et a suivre l'evolution de ce spectre dans le temps apres un arret de pile. L'etude peut porter sur les produits de fission apparus dans toute la pile ou sur ceux apparus dans un echantillon d'uranium suppose place en pile. C'est ce dernier cas que nous avons envisage. La determination du spectre s'appuie sur une partie theorique et sur une partie experimentale. Les fonctionnements de pile choisis sont differents dans les deux cas, ce qui permet d'etendre la gamme de validite des spectres traces ici. (auteur)

  20. Study of fission product {gamma} spectra in the band 2-500 keV; Etude du spectre {gamma} des produits de fission dans la bande 0-500 keV

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    In the study of the {gamma} spectrum of uranium fission products, particular attention has been given in this note to the part of the spectrum ranging between 0 and 500 keV after a given pile operating programme and the evolution of this spectrum with time after a pile shutdown has been followed. The study be related to the fission products which appear in the pile as a whole or on those produced in a uranium sample assumed to have been placed in the pile. The latter case has been envisaged here. The spectrum determination is based partly on theory and partly on experiment. The pile operating conditions are different in the two cases, which widens the range of validity of the spectra traced here. (author) [French] Dans l'etude du spectre {gamma} des produits de fission de l'uranium, on s'est plus particulierement interesse dans la presente note a determiner la partie du spectre qui s'etend entre 0 et 500 keV, au bout d'un fonctionnement donne de pile, et a suivre l'evolution de ce spectre dans le temps apres un arret de pile. L'etude peut porter sur les produits de fission apparus dans toute la pile ou sur ceux apparus dans un echantillon d'uranium suppose place en pile. C'est ce dernier cas que nous avons envisage. La determination du spectre s'appuie sur une partie theorique et sur une partie experimentale. Les fonctionnements de pile choisis sont differents dans les deux cas, ce qui permet d'etendre la gamme de validite des spectres traces ici. (auteur)

  1. Contribution to the study of the desorption of fission gases formed in irradiated uranium oxide; Contribution a l'etude de la desorption des gaz de fission formes dans l'oxyde d'uranium irradie

    Energy Technology Data Exchange (ETDEWEB)

    Berry, J -L; Darras, R; Roger, B [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    The release of {sup 133}Xe from irradiated UO{sub 2} has been studied in the temperature range 1300 to 1900 deg C, using various monocrystalline or sintered samples. Up to 1600 deg C, this release is proportional to the square root of the time and thus occurs essentially by diffusion. The apparent diffusion constant D' decreases and the activation energy of the corresponding process increases as the integrated neutron flux received by the fuel increases. As the density of the sintered samples decreases however, the activation energy of the release also decreases, so that the difference between D' values for sintered samples of different densities decreases as the temperature rises. Finally, above 1600 deg C, the fission gas release phenomenon is governed by UO{sub 2} evaporation, and all the different types of oxide studied have similar behaviors, characterized by poor retention of these gases. (authors) [French] La desorption du xenon 133 forme dans le bioxyde d'uranium irradie a ete etudiee dans l'intervalle de 1300 a 1900 C, a l'aide de differents echantillons monocristallins ou frittes. Jusqu'a 1600 C, elle s'effectue proportionnellement a la racine carree du temps, donc essentiellement par diffusion. La pseudo-constante de desorption D' decroit et l'energie d'activation du processus correspondant croit lorsque le flux de neutrons integre recu par le combustible augmente. Cependant, lorsque la densite des frittes diminue, l'energie d'activation de desorption diminue egalement, de sorte que l'ecart entre les valeurs de D' relatives a des frittes de densites differentes se restreint lorsque, la temperature s'eleve. Finalement, au-dessus de 1600 C, l'evaporation de l'UO{sub 2} regit le phenomene de liberation des gaz de fission, et toutes les qualites d'oxyde etudiees presentent alors des comportements voisins a cet egard, caracterises par une mediocre retention de ces gaz. (auteurs)

  2. CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors

    International Nuclear Information System (INIS)

    England, T.R.; Gorrell, T.C.; Hightower, J.H.

    2001-01-01

    1 - Description of problem or function: CINDER is a four-group, one- point depletion and fission product program based on the evaluation of a general analytical solution of nuclides coupled in any linear sequence of radioactive decays and neutron absorptions in a specified neutron flux spectrum. The desired depletion and fission product chains and all physical data are specified by the problem originator. The program computes individual nuclide number densities, activities, nine energy-group disintegration rates, and macroscopic and barns/fission poisons at each time-step as well as selected summaries of these data. 2 - Method of solution: Time-dependent variations in nuclide cross sections and neutron fluxes are approximated by a user-specified sequential set of values which are considered constant during the duration of the user-specified associated time-increments. When a nuclide concentration is independent of the concentration of any of its progeny, it is possible to resolve the couplings so as to obtain nuclides fed by a single parent. These chains are referred to as linear. 3 - Restrictions on the complexity of the problem: The program is limited to 500 total nuclides formed in up to 240 chains of 20 or fewer nuclides each. Up to 10 nuclides may act as fission product sources, contributing to power, and as many as 99 time-steps of arbitrary length are permitted. All stable nuclides must have a cross section if zero power time-increments are anticipated

  3. Choice of initial conditions in dynamical calculations of distributions of nuclear fission fragments

    International Nuclear Information System (INIS)

    Kosenko, G.I.

    1993-01-01

    The distribution function in the coordinates and momenta for a fissioning system traversing a barrier is determined in terms of Langevin fluctuation-dissipation dynamics. It is shown that this distribution is best described by the Kramers distribution. The equilibrium distribution can be used as the initial condition, provided that the system is in the overdamping regime. 28 refs., 5 figs., 3 tabs

  4. Coordinate-dependent mass and the validity of the WKB approximation in fission barrier penetration calculations

    International Nuclear Information System (INIS)

    Kumar, K.

    1990-03-01

    It is shown that the quasi-classical condition for the validity of the WKB approximation is satisfied although the mass associated with the spontaneous fission of 240 Pu varies by a factor of 12. A general numerical estimate of the allowed variations in such a mass is also given. (author)

  5. Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

    International Nuclear Information System (INIS)

    Walker, J.R.; de Vaal, J.W.; Arimescu, V.I.; McGrady, T.G.; Wong, C.

    1992-04-01

    A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

  6. Calculated model of radioactive fission and corrosion product accumulation and distribution in a fast reactor sodium coolant circuit

    International Nuclear Information System (INIS)

    Kizin, V.D.; Konyashov, V.V.

    1987-01-01

    A simple calculation procedure of radioactive products accumulation and distribution in a primary circuit has been developed on the basis of experimental investigations at the BOR-60 reactor. Common knowledge on the impurity products transfer at the liquid-solid and liquid-gas phase boundary is taken. Use is made of the typical in reactor physics relationships for the description of the products transition to the equipment surfaces, of fission products release, metal corrosion and others. Satisfactory agreement of the calculation data with the experimental ones has been obtained. (orig.)

  7. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  8. SOURCES-3A: A code for calculating (α, n), spontaneous fission, and delayed neutron sources and spectra

    International Nuclear Information System (INIS)

    Perry, R.T.; Wilson, W.B.; Charlton, W.S.

    1998-04-01

    In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an

  9. Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-07-01

    1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

  10. Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

  11. CALCUL DU SPECTRE DE REFLEXION DU MULTICOUCHE Ni/C DANS LE DOMAINE DES RAYONS X

    Directory of Open Access Journals (Sweden)

    A MEDDOUR

    2000-12-01

    Full Text Available Le pouvoir réflecteur d’un dioptre quelconque dans le domaine des rayons X est trop faible, mais il est toujours possible de choisir des systèmes pouvant présenter un pic de réflexion d’intensité importante autour d’une incidence caractéristique du matériau. Ce dernier est un multicouche, composé de deux couches déposées en sandwich.                 Nous avons élaboré un programme qui permet de calculer la réflexion d’un tel matériau en suivant la méthode d’Abelès dans laquelle une couche mince est représentée par une matrice carrée contenant toutes les informations nécessaires pour le calcul de la réflexion. Ce programme tient compte aussi des rugosités aux interfaces du multicouche, vue leur importante influence sur l’intensité du pic apparaissant sur le spectre de réflexion.                 L’application du programme au multicouche Ni/C a montré  l’existence d’un pic centré autour de 31.32°. Son intensité est sensible au nombre de périodes dans le multicouche, aux épaisseurs des couches minces de Ni et de C et à la taille des rugosités des interfaces Ni/C et C/Ni.

  12. Chemical Production using Fission Fragments; Emploi des fragments de fission dans la production de substances chimiques; Ispol'zovanie produkto v raspada v khimicheskom proizvodstve; Empleo de los fragmentos de fision en la industria quimica

    Energy Technology Data Exchange (ETDEWEB)

    Dawson, J K; Moseley, F [AERE, Harwell (United Kingdom)

    1960-07-15

    Some reactor design considerations of the use of fission recoil fragment energy for the production of chemicals of industrial importance have been discussed previously in a paper given at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy [A/Conf. 15/P.76]. The present paper summarizes more recent progress made on this topic at AERE, Harwell. The range-energy relationship for fission fragments is discussed in the context of the choice of fuel system for a chemical production reactor, and the experimental observation of a variation of chemical effect along the length of a fission fragment track is described for the irradiation of nitrogen-oxygen mixtures. Recent results are given on the effect of fission fragments on carbon monoxide-hydrogen gas mixtures and on water vapour. No system investigated to date shows any outstanding promise for large-scale chemical production. (author) [French] En ce qui concerne la construction de reacteurs, certaines considerations relatives a l'emploi de l'energie des fragments de fission de recul pour la production de certaines substances chimiques d'importance industrielle ont deja ete examinees dans un memoire soumis a la deuxieme Conference internationale sur l'utilisation de l'energie atomique a des fins pacifiques [A/Conf. 15/PP. 76]. Le present memoire donne un apercu des progres accomplis depuis lors dans ce domaine par 1'Atomic Energy Research Establishment a Harwell. Les auteurs etudient la relation entre le parcours et l'energie pour des fragments de fission a propos du choix du combustible pour un reacteur destine a la production de substances chimiques; ils decrivent aussi une variation d'effet chimique observee le long de la trajectoire d'un fragment de fission pendant l'irradiation de melanges azote-oxygene. Les auteurs fournissent les resultats de recherches recentes relatives a l'effet des fragments de fission sur des melanges oxyde de carbone-hydrogen e et sur la vapeur d

  13. The Technology and Applications of Large Fission Product Beta Sources; Technologie et applications des grandes sources beta de fission; Tekhnologiya i primenenie krupnykh istochnikov beta-izluchenij, ispuskaemykh produktami deleniya; Tecnologia y utilizacion de los productos de fision como fuentes de irradiacion beta de elevada intensidad

    Energy Technology Data Exchange (ETDEWEB)

    Silverman, Joseph [Radiation Applications Incorporated (United States)

    1960-07-15

    test the validity of the calculations. 4. An engineering cost analysis indicates that fission product beta power can be competitive with alternative sources of radiation. (author) [French] Si les emetteurs beta ne sont pas entres en ligne de compte en tant que sources de rayonnement de grandes dimensions, c'est que dans le passe les procedes sous irradiation presentant un interet pour l'industrie etaient bases sur l'utilisation de particules a grand pouvoir de penetration; de la l'importance considerable que l'on accordait aux rayons gamma et aux electrons artificiellement acceleres. Il y a environ quatre ans, il devint evident que de vastes perspectives s'ouvraient aux applications pratiques des rayonnements beta dans le domaine du traitement par irradiation superficielle, par exemple, la modification par copolymerisation greffee de la surface des moulages en matiere plastique et la pasteurisation superficielle des aliments. En effet, l'irradiation en profondeur occasionne dans ces cas des pertes et peut meme presenter des dangers. D'autre part, il y a deux autres domaines dans lesquels l'utilisation des electrons artificiellement acceleres ne semble pas indiquee; ce sont les syntheses chimiques amorcees par irradiation dans des cuves sous pression et certaines categories de reactions en chaine provoquees par des radicaux libres pour lesquels le taux de production par kilowatt decroit en fonction de la racine carree de l'intensite d'irradiation. Dans tous ces domaines, les sources beta de grandes dimensions presentent des avantages potentiels evidents; puisque la fission nucleaire permet d'en obtenir en quantites importantes, elles meritent que leurs- possibilites d'utilisation fassent l'objet d'un examen approfondi. Compte tenu de ces considerations, on a procede - sous l'egide de la Commission de l'energie atomique - a une etude des applications'et de la technologie des sources beta de fission. Cette etude a permis d'aboutir aux conclusions suivantes: 1. Des

  14. Irradiation effects and behaviour of fission products in zirconia and spinel; Effets d'irradiation et comportement des produits de fission dans la zircone et le spinelle

    Energy Technology Data Exchange (ETDEWEB)

    Gentils, A

    2003-10-01

    Crystalline oxides, such as zirconia (ZrO{sub 2}) and spinel (MgAl{sub 2}O{sub 4}), are promising inert matrices for the transmutation of plutonium and minor actinides. This work deals with the study of the physico-chemical properties of these matrices, more specifically their behaviour under irradiation and their capacity to retain fission products. Irradiations at low energy and incorporation of stable analogs of fission products (Cs, I, Xe) into yttria-stabilized zirconia and magnesium-aluminate spinel single crystals were performed by using the ion implanter IRMA (CSNSM-Orsay). Irradiations at high energy were made on several heavy ion accelerators (GANIL-Caen, ISL-Berlin, HIL-Warsaw). The damage induced by irradiation and the release of fission products were monitored by in situ Rutherford Backscattering Spectrometry experiments. Transmission electron microscopy was also used in order to determine the nature of the damage induced by irradiation. The results show that irradiation of ZrO{sub 2} and MgAl{sub 2}O{sub 4} with heavy ions (about hundred keV and about hundred MeV) induces a huge structural damage in crystalline matrices. Total disorder (amorphization) is however never reached in zirconia, contrary to what is observed in the case of spinel. The results also emphasize the essential role played by the concentration of implanted species on their retention capacity. A dramatic release of fission products was observed when the concentration exceeds a threshold of a few atomic percent. Irradiation of implanted samples with medium-energy noble-gas ions leads to an enhancement of the fission product release. The exfoliation of spinel crystals implanted at high concentration of Cs ions is observed after a thermal treatment at high temperature. (author)

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

    International Nuclear Information System (INIS)

    Misfeldt, I.B.

    1987-01-01

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

  16. Using a zero-variance scheme to accelerate the fission source convergence in a Monte Carlo calculation

    International Nuclear Information System (INIS)

    Christoforou, S.; Hoogenboom, J. E.

    2009-01-01

    We have used Boltzmann entropy in order to test whether a zero-variance based scheme can speed up the fission source convergence in a Monte Carlo calculation. It is shown that the choice of the initial source distribution significantly influences the evolution of the source, even leading to cases where the source does not converge at all throughout the calculation. The results from a loosely coupled system based on the NEA/OECD source convergence benchmarks indicate that, when using a biasing scheme such as the one we have developed, there can be significant improvement in the convergence, up to 3 times faster, which coupled with an figure of merit improvement of 1.5 leads to more efficient calculations. (authors)

  17. Contribution to the study of the diffusion of fission products in uranium; Contribution a l'etude de la diffusion des produits de fission dans l'uranium

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-10-15

    In this work we have developed a simple method for determining the diffusion constants and the solid-state solubilities of a metal which is volatile and only slightly soluble in second metal. This method has been applied to the behaviour of certain fission products in {gamma} uranium: strontium, barium, lanthanum, samarium and cerium. This work has made it possible to show the effect of the atomic radius of the solute on the diffusion constants. (author) [French] Dans ce travail nous avons mis au point une methode simple permettant de determiner les constantes de diffusion ainsi que les solubilites a l'etat solide d'un metal volatil et peu soluble dans un autre. Cette methode a ete appliquee au comportement de certains produits de fission dans l'uranium {gamma}: strontium, baryum, lanthane, samarium et cerium. Cette etude a permis de mettre en evidence le role du rayon atomique du solute sur les constantes de diffusion. (auteur)

  18. Determination of the neutron-induced fission cross section of {sup 242}Pu; Bestimmung des neutroneninduzierten Spaltquerschnitts von {sup 242}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Koegler, Toni Joerg

    2016-04-26

    Neutron induced fission cross sections of actinides like the Pu-isotopes are of relevance for the development of nuclear transmutation technologies. For {sup 242}Pu, current uncertainties are of around 21%. Sensitivity studies show that the total uncertainty has to be reduced to below 5% to allow for reliable neutron physics simulations. This challenging task was performed at the neutron time-of-flight facility of the new German National Center for High Power Radiation Sources at HZDR, Dresden. Within the TRAKULA project, thin, large and homogeneous deposits of {sup 235}U and {sup 242}Pu have been produced successfully. Using two consecutively placed fission chambers allowed the determination of the neutron induced fission cross section of {sup 242}Pu relative to {sup 235}U. The areal density of the Plutonium targets was calculated using the measured spontaneous fission rate. Experimental results of the fast neutron induced fission of {sup 242}Pu acquired at nELBE will be presented and compared to recent experiments and evaluated data. Corrections addressing the neutron scattering are discussed by using results of different neutron transport simulations (Geant 4, MCNP 6 and FLUKA).

  19. Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

    Energy Technology Data Exchange (ETDEWEB)

    Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

    2013-03-15

    The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

  20. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide; Volatilite du ruthenium au cours des operations de vitrification des produits de fission. 1. partie distillation des solutions nitriques calcination des concentrats 2. partie fixation sur un tube d'acier decomposition du peroxyde

    Energy Technology Data Exchange (ETDEWEB)

    Ortins de Bettencourt, A; Jouan, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO{sub 4}, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO{sub 4} marked with {sup 106}Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO{sub 4} is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO{sub 4} {yields} RuO{sub 2} + O{sub 2}. To do this, we have tried to introduce gaseous RuO{sub 4} into a container placed in an electric oven, and to follow the reaction by {gamma} counting. (author) [French] Au cours de la vitrification des solutions de produits de

  1. Volatility of ruthenium during vitrification operations on fission products. part 1. nitric solutions distillation concentrates calcination. part 2. fixation on a steel tube. decomposition of the peroxide; Volatilite du ruthenium au cours des operations de vitrification des produits de fission. 1. partie distillation des solutions nitriques calcination des concentrats 2. partie fixation sur un tube d'acier decomposition du peroxyde

    Energy Technology Data Exchange (ETDEWEB)

    Ortins de Bettencourt, A.; Jouan, A. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-07-01

    During vitrification of fission product solutions, a high percentage of the ruthenium initially present in these solutions in the form of nitrosyl-ruthenium nitrates is volatilized with the production of the peroxide which itself is decomposed to ruthenium dioxide. The aim of this work has been to study the volatility of the ruthenium during the vitrification processes. During the distillation of the nitric solutions, we have studied in particular the influence on the volatility of the temperature , of the chemical form in which the ruthenium is introduced, of the bubbling of a gas through the solution, of the nitric concentration and of the nitrate concentration. During the calcination, we have observed the influence of the temperature, of the time, of the flow rate and of the nature of the carrier gas, as well as of the action of the ruthenium bi-oxide and the iron oxide on the volatility of the ruthenium. Part 2. This report deals with the study of the thermal decomposition of ruthenium peroxide, RuO{sub 4}, and its deposition on steel tubing. After a brief bibliographic review of the various properties of this substance, a study is made, in the first part, of its deposition on a steel tube. In order to do this, we pass a gas current containing RuO{sub 4} marked with {sup 106}Ru through a stainless steel tube subjected to a temperature gradient which decreases in the direction of the gas flow. The temperature at which RuO{sub 4} is deposited or is fixed on the tube is determined and the influence of the flow rate on this deposit is studied. In the second part, an attempt has been made to study by a static method the kinetics of the decomposition reaction of ruthenium peroxide to give the dioxide: RuO{sub 4} {yields} RuO{sub 2} + O{sub 2}. To do this, we have tried to introduce gaseous RuO{sub 4} into a container placed in an electric oven, and to follow the reaction by {gamma} counting. (author) [French] Au cours de la vitrification des solutions de produits de

  2. Ternary fission

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Since its discovery in 1946, light (charged) particle accompanied fission (ternary fission) has been extensively studied, for spontaneous as well as for induced fission reactions. The reason for this interest was twofold: the ternary particles being emitted in space and time close to the scission point were expected to supply information on the scission point configuration and the ternary fission process was an important source of helium, tritium, and hydrogen production in nuclear reactors, for which data were requested by the nuclear industry. Significant experimental progress has been realized with the advent of high-resolution detectors, powerful multiparameter data acquisition systems, and intense neutron and photon beams. As far as theory is concerned, the trajectory calculations (in which scission point parameters are deduced from the experimental observations) have been very much improved. An attempt was made to explain ternary particle emission in terms of a Plateau-Rayleigh hydrodynamical instability of a relatively long cylindrical neck or cylindrical nucleus. New results have also been obtained on the so-called open-quotes trueclose quotes ternary fission (fission in three about-equal fragments). The spontaneous emission of charged particles has also clearly been demonstrated in recent years. This chapter discusses the main characteristics of ternary fission, theoretical models, light particle emission probabilities, the dependence of the emission probabilities on experimental variables, light particle energy distributions, light particle angular distributions, correlations between light particle accompanied fission observables, open-quotes trueclose quotes ternary fission, and spontaneous emission of heavy ions. 143 refs., 18 figs., 8 tabs

  3. Quelques considérations sur l’évolution des normes de calcul des poteaux avec la section mixte acier-béton

    Directory of Open Access Journals (Sweden)

    Nicolae Chira

    2010-03-01

    Full Text Available Depuis près d’un siècle, le système de construction basé sur des portiques en acier ou mixtes acier béton est devenu l’un des types les plus utilisés dans le domaine du génie civil. Plusieurs générations d’ingénieurs se sont préoccupées du développement des méthodes de calcul et des technologies de fabrication relatives à ces structures. En vue d’un dimensionnement optimal des structures, les ingénieurs sont tenus de trouver un compromis entre les exigences structurales de résistance, rigidité et ductilité d’une part, et les objectifs d’utilisation et de fonction relevant d’exigences architecturales d’autre part. Cette article fait une comparaison entre différents méthodes de dimensionnement des poteaux mixtes acier béton, en tenant compte des plusieurs paramètres.

  4. 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

  5. Half-lives for proton emission, alpha decay, cluster radioactivity, and cold fission processes calculated in a unified theoretical framework

    International Nuclear Information System (INIS)

    Duarte, S.B.; Tavares, O.A.P.; Guzman, F.; Dimarco, A.; Garcia, F.; Goncalves, M.

    2002-01-01

    Half-life values of spontaneous nuclear decay processes are presented in the framework of the Effective Liquid Drop Model (ELDM) using the combination of varying mass asymmetry shape description for the mass transfer with Werner-Wheeler's inertia coefficient V MAS /WW. The calculated half-lives of ground-state to ground-state transitions for the proton emission, alpha decay, cluster radioactivity, and cold fission processes are compared with experimental data. Results have shown that the ELDM is a very efficient model to describe these different decay processes in a same, unified theoretical framework. A Table listing the predicted half-life values, τ c is presented for all possible cases of spontaneous nuclear break-up such that -7.30 10 τ c [S] 10 (τ/τ c ) > -17.0, where τ is the total half-life of the parent nucleus. (author)

  6. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  7. Fission theory and actinide fission data

    Energy Technology Data Exchange (ETDEWEB)

    Michaudon, A.

    1975-06-01

    The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and γ rays emitted in the resonances of {sup 235}U and {sup 239}Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and γ-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n,γf) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for {sup 240}Pu.

  8. Transmutation of Tc-99 and I-129 in fission reactors. A calculational study

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1995-03-01

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MW e CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MW e ), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MW e PWR). When all guide tubes of an UO 2 fuelled PWR are filled with Tc-99 with density of 5 g cm -3 , the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

  9. Forecasting the Quantity and Activity of Fission Products in France in Future Years in the Light of Atomic Energy Development; Quantite et Activite des Produits de Fission Obtenus en France dans les Annees a Venir Compte Tenu du Developpement de l'Energie Atomique; 041a 041e 041b 0414 ; Cantidad y Actividad de los Productos de Fision que se Obtendran en Francia en los Anos Venideros, Habida Cuenta del Desarrollo de la Energia Atomica

    Energy Technology Data Exchange (ETDEWEB)

    Guirlet, J.; Lavie, J. M. [Commissariat a l' Energie Atomique, Saclay (France)

    1960-07-01

    With the aid of Wigner and Way's formula it is possible to predict theoretically the activity of the complex mixture of fission products originating in a reactor. Account is taken in the study of prospective nuclear power production in France up to 1975. It is assumed that the uranium stays in the reactor for periods of three and of six months. It is also possible to determine the activity of a given fission product and to calculate its decay time. The substance selected was strontium for an activation period of three months. Each group of curves shows total activity at any moment and the activity of particular fission products over a given period. (author) [French] En utilisant la formule de Wigner et Way, il est possible de prevoir theoriquement l'activite du melange complexe de produits de fission provenant d'une pile. L'etude a ete faite en tenant compte de previsions en ce qui concerne en France la production d'electricite d'origine atomique jusqu'en 1975. On a suppose des temps de sejour en pile de l'uranium de trois mois et de six mois. Il est possible egalement de trouver l'activite d'un produit de fission particulier, et de donner sa decroissance. Le corps choisi est le strontium pour un temps d'activation de trois mois. Chaque ensemble de courbes donne a tout instant l'activite totale, et l'activite propre des produits de fission correspondant a une periode donnee. (author) [Spanish] Empleando la formula de Wigner y Way se puede calcular teoricamente la actividad de la mezcla compleja de productos de fision procedentes de una pila. Para esta memoria se han tenido en cuenta las previsiones de produccion de electricidad nucleoelectrica en Francia hasta 1975. Los autores suponen que el uranio permanece tres o seis meses en la pila y han conseguido determinar la actividad de un producto de fision determinado, asi como su periodo de desintegracion. Han escogido el estroncio, con un tiempo de activacion de 3 meses. Cada conjunto de curvas da la actividad total

  10. Calculation of pole profiles for the design of particle accelerators magnets. Equivalent current and conformal representation methods; Calcul de profils polaires destines a des aimants d'accelerateurs de particules. Methode des courants et methode des transformations conformes

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-04-01

    These two methods allow the determination of the shape of the poles in magnets, for a given field distribution in the air-gap. First method: The principle of the method consists to create the desired law of field by means of current sheets in which one can adjust the density given in a polynomial form. For the right distribution of these currents, the equipotential corresponding to the magnetic potential of the excitation coils is calculated. The pole profile of the H or C magnet identified with this equipotential line will finally take the place of the distribution of the current sheets used in the calculation. Steel permeability is assumed to be infinite and Foucault current effects are neglected in the case of variable fields. Second method: It consists to find a conformal representation that maps the pole profile plane upon the upper half of another plane where the equipotentials are two half straight lines, and where the field problems are easier to solve. Steel permeability is also considered to be infinite and the coils far from the pole faces. This known method has been applied to be compared with the first one. (author) [French] Ces deux methodes consistent a determiner la forme des pieces polaires d'aimants pour une distribution de champ determinee a l'avance dans l'entrefer. Premiere methode: Le principe de la methode consiste a creer la loi de champ desiree par l'intermediaire de nappes de courant dont on peut ajuster la densite exprimee sous une forme polynominale. Pour une distribution convenable de ces courants, on calcule l'equipotentielle correspondant au potentiel magnetique des bobines d'excitation. Le profil polaire d'un aimant en H ou C identifie a l'equipotentielle se substitue finalement a la repartition des nappes de courant utilisee dans la methode de calcul. La permeabilite de l'acier est supposee infinie et les courants de Foucault sont negliges dans le cas des champs variables. Seconde methode: Elle consiste a trouver une transformation

  11. Calculation of fission product behavior in a multiple reactor barriers in case of an accident

    International Nuclear Information System (INIS)

    Ezzedin, A. A.; Dakhil, A. S.; Elbaden, S. E.

    2012-12-01

    Radiation protection of the population in case of a reactor accident utilizes reference levels which are based on doses values. Therefore, adequate provisions for effective and timely dose assessment for population in case of accidents at nuclear power plant (NPP) are important. Developing the background for such provisions is the objective of this study. In particular, an exponential model has been developed and utilized to calculate the release rate of the most volatile gaseous materials from different reactor barriers. Calculation has been performed for noble gases (1 33X e, 1 35X e, 1 38X e, 8 5K r, 8 7K r, 8 8K r) and the halogens(1'3 1I , 1 32I , 1 33I , 1'3 4I , 1 35I ). The effective dose rate equivalent is calculations in the nearly stage of a reactor accident. Calculations are performed using the MCNP-4C code. The results are comparable with the final analysis report which utilizes different codes. Results of our calculation shows no excessive dose in populated regions and it is recommended to use secondary containment barrier for highly reduction of the release rate to the environment. (Author)

  12. Production and study of fission fragments, from Lohengrin to Alto; Production et etude des fragments de fission, de Lohengrin a Alto

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, F

    2005-06-15

    The study of nuclei far from stability is constitutive of the history of nuclear physics at its very beginning and has been making considerable great strides since then. The study of these nuclei give the opportunity to reach new information on the nuclear structure and thus to measure the solidity of our knowledge on nuclear matter and its validity when it is pushed to its limits. The reaction selected for the production of exotic nuclei in the framework of the PARRNe program is the fission of uranium 238. The nuclei produced have an intermediate mass and are very rich in neutrons. The technique to recover them in order to accelerate them is the thick target method called also the Isol technique. The installation of the ancient Lep injector at the Tandem line in Orsay (IPN) is expected to increase by a factor 100 the production rate of exotic nuclei in the PARRNe program, it is the Alto project. The work presented here concerns studies carried out at the Lohengrin spectrometer installed at the ILL in Grenoble, and at the Tandem installation in Orsay. This document is divided into 4 parts: 1) in flight techniques at Lohengrin, 2) the Isol technique, 3) magic numbers in the domain N=50, and 4) the Alto project.

  13. FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

    International Nuclear Information System (INIS)

    Stamatelatos, M.G.; England, T.R.

    1977-05-01

    FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

  14. Ternary fission

    Indian Academy of Sciences (India)

    the energy minimization of all possible ternary breakups of a heavy radioactive nucleus. Further, within the TCM we have analysed the competition between different geometries as well as different positioning of the fragments. Also, an attempt was made to calculate the mass distribution of ternary fission process within the ...

  15. Development of calculation method for one-dimensional kinetic analysis in fission reactors, including feedback effects

    International Nuclear Information System (INIS)

    Paixao, S.B.; Marzo, M.A.S.; Alvim, A.C.M.

    1986-01-01

    The calculation method used in WIGLE code is studied. Because of the non availability of such a praiseworthy solution, expounding the method minutely has been tried. This developed method has been applied for the solution of the one-dimensional, two-group, diffusion equations in slab, axial analysis, including non-boiling heat transfer, accountig for feedback. A steady-state program (CITER-1D), written in FORTRAN 4, has been implemented, providing excellent results, ratifying the developed work quality. (Author) [pt

  16. Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

    Directory of Open Access Journals (Sweden)

    Minato Futoshi

    2016-01-01

    Full Text Available Nuclear β-decay and delayed neutron (DN emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA and the Hauser-Feshbach statistical model (HFSM. In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

  17. Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Allelein, Hans-Josef

    2014-01-01

    The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

  18. Calculation of multidimensional potential energy surfaces for even-even transuranium nuclei: systematic investigation of the triaxiality effect on the fission barrier

    Science.gov (United States)

    Chai, Qing-Zhen; Zhao, Wei-Juan; Liu, Min-Liang; Wang, Hua-Lei

    2018-05-01

    Static fission barriers for 95 even-even transuranium nuclei with charge number Z = 94–118 have been systematically investigated by means of pairing self-consistent Woods-Saxon-Strutinsky calculations using the potential energy surface approach in multidimensional (β 2, γ, β 4) deformation space. Taking the heavier 252Cf nucleus (with the available fission barrier from experiment) as an example, the formation of the fission barrier and the influence of macroscopic, shell and pairing correction energies on it are analyzed. The results of the present calculated β 2 values and barrier heights are compared with previous calculations and available experiments. The role of triaxiality in the region of the first saddle is discussed. It is found that the second fission barrier is also considerably affected by the triaxial deformation degree of freedom in some nuclei (e.g., the Z=112–118 isotopes). Based on the potential energy curves, general trends of the evolution of the fission barrier heights and widths as a function of the nucleon numbers are investigated. In addition, the effects of Woods-Saxon potential parameter modifications (e.g., the strength of the spin-orbit coupling and the nuclear surface diffuseness) on the fission barrier are briefly discussed. Supported by National Natural Science Foundation of China (11675148, 11505157), the Project of Youth Backbone Teachers of Colleges and Universities of Henan Province (2017GGJS008), the Foundation and Advanced Technology Research Program of Henan Province (162300410222), the Outstanding Young Talent Research Fund of Zhengzhou University (1521317002) and the Physics Research and Development Program of Zhengzhou University (32410017)

  19. A rigorous nonorthogonal configuration interaction approach for the calculation of electronic couplings between diabatic states applied to singlet fission

    NARCIS (Netherlands)

    Wibowo, Meilani; Broer, Ria; Havenith, Remco W. A.

    2017-01-01

    For the design of efficient singlet fission chromophores, knowledge of the factors that govern the singlet fission rate is important. This rate is approximately proportional to the electronic coupling between the lowest (diabatic) spin singlet state that is populated following photoexcitation state

  20. FIFRELIN - TRIPOLI-4® coupling for Monte Carlo simulations with a fission model. Application to shielding calculations

    Science.gov (United States)

    Petit, Odile; Jouanne, Cédric; Litaize, Olivier; Serot, Olivier; Chebboubi, Abdelhazize; Pénéliau, Yannick

    2017-09-01

    TRIPOLI-4® Monte Carlo transport code and FIFRELIN fission model have been coupled by means of external files so that neutron transport can take into account fission distributions (multiplicities and spectra) that are not averaged, as is the case when using evaluated nuclear data libraries. Spectral effects on responses in shielding configurations with fission sampling are then expected. In the present paper, the principle of this coupling is detailed and a comparison between TRIPOLI-4® fission distributions at the emission of fission neutrons is presented when using JEFF-3.1.1 evaluated data or FIFRELIN data generated either through a n/g-uncoupled mode or through a n/g-coupled mode. Finally, an application to a modified version of the ASPIS benchmark is performed and the impact of using FIFRELIN data on neutron transport is analyzed. Differences noticed on average reaction rates on the surfaces closest to the fission source are mainly due to the average prompt fission spectrum. Moreover, when working with the same average spectrum, a complementary analysis based on non-average reaction rates still shows significant differences that point out the real impact of using a fission model in neutron transport simulations.

  1. Migration under irradiation of the I, Cs fission products in SiC; Migration sous irradiation des produits de fission I, Cs dans SiC

    Energy Technology Data Exchange (ETDEWEB)

    Benyagoub, A. [Paris-11 Univ., 91 - Orsay (France). Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse; Centre Interdisciplinaire de Recherche Ions Lasers (CIRIL), 14 - Caen (France)

    2007-07-01

    This work is a part of the CEA-CNRS program on the behaviour of materials under irradiation. Its aim is to contribute to scientifically base the predicting methods of the behaviour of ceramic materials under irradiation in using the important development of calculation means. In particular, its role has been 1)to develop the basic knowledge and the interactions physics and 2)to elaborate calculation models at relevant scales. The studied topics are until now, the damage mechanisms, the diffusion under irradiation, the micro-structural evolutions and the incidences on the mechanical properties. (O.M.)

  2. Etude structurale et propriétés des verres peralumineux de conditionnement des produits de fission et actinides mineurs"

    OpenAIRE

    Gasnier , Estelle

    2013-01-01

    In this work, peraluminous glasses (lack of alkaline and alkaline earth ions regarding aluminum) are under study to assess the potentiality of these matrices to confine fission products and minor actinides (FPA) at higher rate than current R7T7 glass (18,5 wt % FPA). The first part of this work aims at studying the physical and chemical properties of complex peraluminous glasses containing increasing FPA rate (18.5 to 32 wt %) to compare them with the specifications. The very low crystallizat...

  3. Nondestructive analysis of the RA fuel burnup, Calculation of the gamma activity ratio of fission products in the fuel - program QU0C1

    International Nuclear Information System (INIS)

    Bulovic, V.F.

    1973-01-01

    The γ radiation of RA reactor fuel element was measured under precisely defined measuring conditions. The spectrum was analysed by spectrometer with semiconductor Ge(Li) detector. The gamma counting rate in the fuel spectrum is defined as a function of fission product activity, gamma energy and yield, fuel thickness and additional absorbers, dimensions of the gamma collimator. Activity ratio of two fission products is defined as a function of counting rate peaks and part of the mentioned quantities. Four options for calculating the activities for fission products are discussed. Three of them are covered by the QU0C1 code written in FORTRAN for the CDC 3600 computer. The code is included in this report [sr

  4. SPARC-90: A code for calculating fission product capture in suppression pools

    International Nuclear Information System (INIS)

    Owczarski, P.C.; Burk, K.W.

    1991-10-01

    This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs

  5. Half-lives for proton emission, alpha decay, cluster radioactivity, and cold fission processes calculated in a unified theoretical framework

    Energy Technology Data Exchange (ETDEWEB)

    Duarte, S.B.; Tavares, O.A.P.; Guzman, F.; Dimarco, A. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Garcia, F. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Universidade Estadual de Santa Cruz, Ilheus, BA (Brazil). Dept. de Ciencias Exatas e Tecnologicas; Rodriguez, O. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Instituto Superior de Ciencias e Tecnologia Nucleares, La Habana (Cuba); Goncalves, M. [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil)

    2002-01-01

    Half-life values of spontaneous nuclear decay processes are presented in the framework of the Effective Liquid Drop Model (ELDM) using the combination of varying mass asymmetry shape description for the mass transfer with Werner-Wheeler's inertia coefficient V{sub MAS}/WW. The calculated half-lives of ground-state to ground-state transitions for the proton emission, alpha decay, cluster radioactivity, and cold fission processes are compared with experimental data. Results have shown that the ELDM is a very efficient model to describe these different decay processes in a same, unified theoretical framework. A Table listing the predicted half-life values, {tau}{sub c} is presented for all possible cases of spontaneous nuclear break-up such that -7.30 <{approx_equal} log{sub 10} {tau}{sub c} [S] <{approx_equal} 27.50 and log {sub 10}({tau}/{tau}{sub c}) > -17.0, where {tau} is the total half-life of the parent nucleus. (author)

  6. A method of calculating fission gas diffusion from UO{sub 2} fuel and its application to the X-2-f loop test

    Energy Technology Data Exchange (ETDEWEB)

    Booth, A H

    1957-09-15

    A method for calculating the fraction of the rare gas fission products that diffuses out of a UO{sub 2} fuel element under conditions In a reactor is outlined, The method is based on the values of the diffusion constant found in laboratory experiments, as described In CRDC-718, and assumes that these remain unaltered during the period that the fuel is in the reactor, The method has been applied to two types of oxide in the X-2-f loop test of 1956 and the results compared with the amounts of fission gas found by analysis of the gases collected in sheath puncture experiments, as described in CRDC-719. The calculated values depend heavily on the estimated temperatures In the fuel. They are in close agreement with the experimental values provided that, in calculating the temperature, certain assumptions are made regarding the thermal expansion of the oxide cylinder. (author)

  7. CACA-2: revised version of CACA-a heavy isotope and fission-product concentration calculational code for experimental irradiation capsules

    International Nuclear Information System (INIS)

    Allen, E.J.

    1976-02-01

    A computer program is described which calculates nuclide concentration histories, power or neutron flux histories, burnups, and fission-product birthrates for fueled experimental capsules subjected to neutron irradiations. Seventeen heavy nuclides in the chain from 232 Th to 242 Pu and a user-specified number of fission products are treated. A fourth-order Runge-Kutta calculational method solves the differential equations for nuclide concentrations as a function of time. For a particular problem, a user-specified number of fuel regions may be treated. A fuel region is described by volume, length, and specific irradiation history. A number of initial fuel compositions may be specified for each fuel region. The irradiation history for each fuel region can be divided into time intervals, and a constant power density or a time-dependent neutron flux is specified for each time interval. Also, an independent cross-section set may be selected for each time interval in each irradiation history. The fission-product birthrates for the first composition of each fuel region are summed to give the total fission-product birthrates for the problem

  8. Calculation characterization of spent fuel hazard related to partitioning and transmutation of minor actinides and fission products

    International Nuclear Information System (INIS)

    Gerasimov, A. S.; Bergelson, B. R.; Tikhomirov, G.V.; Volovik, A.I. . E-mail of corresponding author: geras@itep.ru; Gerasimov, A.S.)

    2005-01-01

    Radiotoxicity is one of important characteristics of radwaste hazard. Radiotoxicity of actinides and fission products from spent fuel of VVER-1000 reactor for processes of burnup, long-term storage, and transmutation is discussed. (author)

  9. Treatment of solutions of fission products - Separation of caesium-137; Traitement des solutions de produits de fission - Separation du cesium-137; Obrabotka rastvorov produktov deleniya - Razdelenie tseziya-137; Tratamiento de soluciones de productos de fision - Separacion del cesio-137

    Energy Technology Data Exchange (ETDEWEB)

    Stiennon-Bovy, R [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium); Gvion, R [Commission Israelienne de l' Energie Atomique, Tel-Aviv (Israel)

    1962-01-15

    For the industrial recovery of caesium-137 from solutions of fission products, the authors utilized the analytical method for determination of caesium by dipicrylamine, adapting it to use on an industrial scale and to the high level of the activities encountered. The process recommended makes it possible both to isolate caesium as a chloride and to recover the precipitation reagent, in one and the same operation. A basic method is suggested. The authors studied the effect of radiation on dipicrylamine and its compounds, this effect proving to be practically nil for solid compounds and negligible for their solutions. The entrainment of caesium by ammonia ion was also studied. The advantages of the proposed process are : high decontamination of the caesium, simple operation and free recycling fo the reagent, high yield for caesium recovery and for dipicrylamine, considerable concentration of caesium activity, operation at room temperature and possibility of continuous operation. By this process caesium can be recovered before certain fission products are eliminated. (author) [French] Pour la recuperation industrielle du cesium-137 des solutions de produits de fission, les auteurs ont adapte le procede analytique de dosage du cesium par la dipicrylamine a l'echelle industrielle et au niveau eleve des activites traitees. Le procede preconise permet en une seule operation d'isoler le cesium sous forme de chlorure et de recuperer aisement le reactif de precipitation; un schema de principe est propose. Les auteurs ont etudie l'effet des radiations sur la dipicrylamine et ses composes; cet effet s'avere pratiquement nul pour les composes solides et negligeable pour leurs solutions. L'entrainement du cesium par l'ion ammoniaque est etudie. Les avantages du procede propose sont la decontamination elevee du cesium, la simplicite du mode operatoire avec recyclage aise du reactif, le rendement eleve pour la recuperation du cesium et pour la dipicrylamine, une concentration

  10. Release of Fission Products from UC-ZrC Fuel Inserts; Degagement des produits de fission liberes dans des noyaux combustibles UC-ZrC; Vydelenie produktov deleniya iz topliv UC - ZrC; Liberacion de productos de fision por pastillas de combustible de UC-ZrC

    Energy Technology Data Exchange (ETDEWEB)

    Barth, F.; Von der Decken, C. B.; Schifferstein, K. [Brown Boveri/Krupp Reaktorbau G.M.B.H., Duesseldorf (Germany); Clauss, A.; Reichel, H.; Rygaert, J.; Ruston, W. R. [Societe d' Etudes de Recherches et d' Applications pour l' Industrie (S.E.R.A.I.), Brussels (Belgium)

    1963-11-15

    The release of fission products from pellets of the approximate composition UC + 20 ZrC has been investigated in a loop during irradiation. The purpose was to check the possibility to use these pellets as inserts for the spherical fuel elements of the high temperature reactor presently being built at Juelich by Brown-Boveri/Krupp Reaktorbau G.m.b.H. The tests were carried out on free pellets and on pellets enclosed in graphite capsules. Some pellets had a 1- to 2-mm thick border zone consisting of pure zirconium carbide. The samples were placed in electric resistance furnaces reaching a maximum temperature of 1600{sup o}C. The neutron fluxes during the irradiation were 3 to 4 x 10{sup 10} n/cm{sup 2}s. The loop permitted the measurement of the release of the rare gas isotopes Kr{sup 85m}, Kr{sup 87}, Kr{sup 88}, Xe{sup 133} and Xe{sup 135} during irradiation and the indirect determination of I{sup 133} and I{sup 135} after shut-down of the reactor. The minimum release of I{sup 131}, Sr{sup 89}, Ba{sup 140} and Ce{sup 141} was radiochemically determined after withdrawal of the specimen. It was also possible to measure the recoil fraction {epsilon}{sub R} alone when the fuel insert was suspended within a furnace of larger diameter. The measured values of {epsilon}{sub R} were only slightly higher than the calculated ones. At higher sample temperatures the diffusion fraction {epsilon}{sub D} could be calculated by subtraction of the recoil fraction from the total release. From {epsilon}{sub D} it was possible to calculate D/r{sub 0}{sup 2}s{sup -1} using the approximation equations of Fliigge and Zimens where D = diffusion constant and r{sub 0} = mathematically defined mean particle radius. The logarithmic plot of these values against 1/T yielded the activation energies (T = sample temperature in {sup o}K). The activation energies for the diffusion of krypton and xenon did not show significant differences. There were likewise no big differences in the activation

  11. Amount and activity of fission products which will be obtainable in France in the immediate future taking into account the development of atomic energy; Quantite et activite des produits de fission abtenus en France dans les annees a venir compte tenu du developpement de l'energie atomique

    Energy Technology Data Exchange (ETDEWEB)

    Guirlet, J; Lavie, J M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    By using the Wigner and Way formula, the activity of the complex mixture of fission products produced in a pile may be estimated theoretically in advance. This study was carried out on the basis of forecasts, in the case of France for the production of electricity of atomic origin up to the year 1975. The uranium was assumed to be in the pile during periods of three months and six months. It is also possible to find the activity of a particular fission product and to give its decay rate. The element chosen is strontium for a three months' activation period. Each set of curves gives at any moment the total activity accumulated, and the characteristic activity of the fission products corresponding to a given half-life. (author) [French] En utilisant la formule de Wigner et Way, il est possible de prevoir theoriquement l'activite du melange complexe de produits de fission provenant d'une pile. L'etude a ete faite en tenant compte de previsions, en ce qui concerne la France, de la production d'electricite d'origine atomique jusqu'en 1975. On a suppose des temps de sejour en pile de l'uranium de trois mois et de six mois. Il est possible egalement de trouver l'activite d'un produit de fission particulier, et de donner sa decroissance. Le corps choisi est le strontium pour un temps d'activation de trois mois. Chaque ensemble de courbes donne a tout instant l'activite totale accumulee, et l'activite propre des produits de fission correspondant a une periode donnee. (auteur)

  12. EURISOL-DS Multi-MWatt Hg Target: Neutron flux and fission rate calculations for the MAFF configuration

    CERN Document Server

    Romanets, Y; Vaz, P; Herrera-Martinez, A; Kadi, Y; Kharoua, C; Lettry, J; Lindroos, M

    The EURISOL (The EURopean Isotope Separation On-Line Radioactive Ion Beam) project aims at producing high intensity radioactive ion beams produced by neutron induced fission on a fissile target (235U) surrounding a liquid mercury converter. A proton beam of 1 GeV and 4 MW impinges on the Hg converter generating by spallation reactions high neutron fluxes. In this work the state-of-the-art Monte Carlo codes MCNPX and FLUKA were used to assess the neutronics performance of the system which geometry, inspired from the MAFF concept, allows a versatile manipulation of the fission targets. The objective of the study was to optimize the geometry of the system and the materials used in the fuel and reflector elements of the system, in order to achieve the highest possible fission rate.

  13. Cross-Sections for Low-Energy Neutron-Induced Fission; Sections Efficaces de Fission pour des Neutrons de Faible Energie; 0421 0415 0427 0415 041d 0418 042f 0414 0415 041b 0415 041d 0418 042f , 0412 042b 0417 0412 0410 041d 041d 041e 0413 041e 041d 0415 0419 0422 0420 041e 041d 0410 041c 0418 041d 0418 0417 K 041e 0419 042d 041d 0415 0420 0413 0418 0418 ; Secciones Eficaces en la Fision Inducida por Neutrones de Baja Energia

    Energy Technology Data Exchange (ETDEWEB)

    Rae, E. R. [Atomic Energy Research Establishment, Harwell, Didcot, Berks. (United Kingdom)

    1965-07-15

    The cross-sections of the heavy nuclei for neutron-induced fission are of fundamental importance to the technology of nuclear energy production. The manner in which these cross-sections vary with the neutron energy and with the mass and charge of the target nuclei also provides much information on the structure of heavy nuclei. In the case of a thermally fissile target nucleus, as the neutron energy is increased from thermal the cross-section first exhibits an inverse velocity dependence, followed by a region in which sharp resonance peaks appear, and finally a continuum region where the cross-section exhibits relatively smooth steps and breaks. All these phenomena can be explained in principle, but certain features of the data have proved very difficult to explain quantitatively. Recent cross-section measurements, stimulated by the needs of reactor technology, have concentrated on improving the energy resolution and accuracy of the data on fuel materials, and this has led to a more detailed study of the resonance region, which is of considerable interest for reactor Doppler effect calculations. More accurate measurements at rather higher energies have established the existence of appreciable fission cross-sections below the so-called fission threshold in certain cases. Careful measurements of this nature have in turn stimulated interest in the interpretation of the cross-sections in terms of nuclear models on a firm quantitative basis. This paper outlines the main features of neutron-induced-fission cross-sections and their interpretation. Some emphasis is placed on recent improvements in the quality of the measurements and in attempts at quantitative interpretation of certain aspects of the data. (author) [French] Les sections efficaces de fission des noyaux lourds ont une importance essentielle pour la technologie de la production d'energie d'origine nucleaire. La maniere dont ces sections efficaces varient Selon l'energie des neutrons et Selon la masse et la

  14. Fission product removal by containment spray - influence of the distance between the drops on the aerosol collection efficiency; Influence de la densite spatiale des gouttes d'aspersion sur l'efficacite de collecte des produits de fission

    Energy Technology Data Exchange (ETDEWEB)

    Gauchet, N

    2000-07-01

    This work is within the framework of the studies that are conducted at the IPSN concerning the loss of coolant in a nuclear reactor. During this kind of accident, a spray system in the reactor containment induces the scrubbing of fission products in the atmosphere, and allows the decrease of their concentration in the containment. Our objective is to study the influence of the distance between the drops of their aerosol collection efficiency. This is not taken into account in the existing models. We stimulate the various aerosol collection mechanisms with one free falling drop using computational fluid dynamics codes. The mechanisms are: deposition by brownian diffusion, impaction and interception of the particles by the drop, and collection of particles in the presence of steam condensation at the surface of the drop. These phenomena are studied for drops ranging in diameter from 100 to 700 micrometers, falling in a saturated air-steam mixture whose temperature varies between 20 and 140 degrees Celsius, and total pressure varies between 1 and 5 bars. We validate these models with results available in the literature and with experimental results CARAIDAS. Then we apply these models to the case of three aligned drops, which constitutes a case for which the proximity of the drops has a strong influence on the collection of aerosols. While varying the distance between the drop from 5 to 25 drop diameters, we can highlight the modification of the collection efficiency of each mechanism related to the disturbance of the velocity and concentration fields in the vicinity of the drops. We note that the strongest variations of efficiency observed are in the field of impaction, and that the steam condensation at the surface of the drops limits the decrease. (authors)

  15. Fission level densities

    International Nuclear Information System (INIS)

    Maslov, V.M.

    1998-01-01

    Fission level densities (or fissioning nucleus level densities at fission saddle deformations) are required for statistical model calculations of actinide fission cross sections. Back-shifted Fermi-Gas Model, Constant Temperature Model and Generalized Superfluid Model (GSM) are widely used for the description of level densities at stable deformations. These models provide approximately identical level density description at excitations close to the neutron binding energy. It is at low excitation energies that they are discrepant, while this energy region is crucial for fission cross section calculations. A drawback of back-shifted Fermi gas model and traditional constant temperature model approaches is that it is difficult to include in a consistent way pair correlations, collective effects and shell effects. Pair, shell and collective properties of nucleus do not reduce just to the renormalization of level density parameter a, but influence the energy dependence of level densities. These effects turn out to be important because they seem to depend upon deformation of either equilibrium or saddle-point. These effects are easily introduced within GSM approach. Fission barriers are another key ingredients involved in the fission cross section calculations. Fission level density and barrier parameters are strongly interdependent. This is the reason for including fission barrier parameters along with the fission level densities in the Starter File. The recommended file is maslov.dat - fission barrier parameters. Recent version of actinide fission barrier data obtained in Obninsk (obninsk.dat) should only be considered as a guide for selection of initial parameters. These data are included in the Starter File, together with the fission barrier parameters recommended by CNDC (beijing.dat), for completeness. (author)

  16. Fast fission phenomena

    International Nuclear Information System (INIS)

    Gregoire, Christian.

    1982-03-01

    Experimental studies of fast fission phenomena are presented. The paper is divided into three parts. In the first part, problems associated with fast fission processes are examined in terms of interaction potentials and a dynamic model is presented in which highly elastic collisions, the formation of compound nuclei and fast fission appear naturally. In the second part, a description is given of the experimental methods employed, the observations made and the preliminary interpretation of measurements suggesting the occurence of fast fission processes. In the third part, our dynamic model is incorporated in a general theory of the dissipative processes studied. This theory enables fluctuations associated with collective variables to be calculated. It is applied to highly inelastic collisions, to fast fission and to the fission dynamics of compound nuclei (for which a schematic representation is given). It is with these calculations that the main results of the second part can be interpreted [fr

  17. Spontaneous fission of superheavy nuclei

    Indian Academy of Sciences (India)

    The fission-like configurations are used for the total deformation energy calculations. A ... oscillator potential for the two fission fragment regions reads as ... Beyond this limit, the contribution of more remote levels is negligible. Once the density ...

  18. Advanced burnup calculation code system in a subcritical state with continuous-energy Monte Carlo code for fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Ohta, Masayuki; Miyamaru, Hiroyuki; Murata, Isao

    2009-01-01

    The fusion-fission (FF) hybrid reactor is a promising energy source that is thought to act as a bridge between the existing fission reactor and the genuine fusion reactor in the future. The burnup calculation system that aims at precise burnup calculations of a subcritical system was developed for the detailed design of the FF hybrid reactor, and the system consists of MCNP, ORIGEN, and postprocess codes. In the present study, the calculation system was substantially modified to improve the calculation accuracy and at the same time the calculation speed as well. The reaction rate estimation can be carried out accurately with the present system that uses track-length (TL) data in the continuous-energy treatment. As for the speed-up of the reaction rate calculation, a new TL data bunching scheme was developed so that only necessary TL data are used as long as the accuracy of the point-wise nuclear data is conserved. With the present system, an example analysis result for our proposed FF hybrid reactor is described, showing that the computation time could really be saved with the same accuracy as before. (author)

  19. Detection of alpha particles and Cf-252 fission fragments with solid track detectors and surface barrier detector. Efficiency calculation

    International Nuclear Information System (INIS)

    Khouri, M.T.F.C.; Koskinas, M.F.; Andrade, C. de; Vilela, E.C.; Hinostroza, H.; Kaschiny, J.E.A.; Costa, M.S. da; Rizzo, P.; Santos, W.M.S.

    1990-01-01

    A technique for particle detection by using track solid detector and also types of revealing and result analysis are presented concerned to Cf-252 fission fragments detection. Measurements of alpha particles detection efficiency using Makrofol E and surface barrier detector are performed. (L.C.J.A.)

  20. Radio-chemical applications of columns of precipitated insoluble sulphates - contribution to the study of fission product solutions (1962); Applications en radiochimie de colonnes de precipites de sulfates insolubles - contribution a l'etude des solutions vieillies de produits de fission (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Sauvagnac, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Barrachina, M [Junta de la Energia Nuclear, Madrid (Spain)

    1962-07-01

    After describing the preparation of SrSO{sub 4} and BaS0{sub 4} columns, the authors study the following variables affecting fixation of Sr{sup 90}: - influence of pH, of the SO{sub 4}{sup --} concentration in the solution studied, of the column density, of the flow rate, of the use of the EDTA complex. A mechanism is proposed for the exchange and fixation. Once the column characteristics had been chosen, the fixation of rare-earths in the absence or presence of EDTA was studied. - it was observed that for nitrate or chloride concentrations of over 2 mg/ml, or in the presence of stronger than 1 N HNO{sub 3} or HCl, barium was fixed selectively. - these results were applied to radiochemical separations in fission product solutions: Ba{sup 140} and Sr{sup 90} separation, and that of the ceric earth group and the yttric earth group. - In order to establish a complete scheme for the radio-chemical separation in fission product solution, after at least one year ageing, the techniques of column separation and of solvent extraction using thenoyltrifluoroacetone and bis di-ethylhexylphosphoric acid are combined. (authors) [French] Apres avoir decrit la preparation des colonnes de SrS0{sub 4} et BaS0{sub 4}, on etudie les variables suivantes pour la fixation de {sup 90}Sr: - influence du pH, de la concentration en SO{sub 4}{sup --} dans la solution etudiee, de la densite de la colonne, de la vitesse de passage, de l'emploi du complexant EDTA. On propose une explication du mode d'echange et de fixation. Les caracteristiques de la colonne etant choisies, on a etudie la fixation des terres rares en presence ou non de EDTA. - on a d'autre part constate qu'en milieu chlorure ou nitrate, de concentration superieure a 2 mg/ml, ou en presence de HNO{sub 3} ou HCl superieur a 1 N, on fixait selectivement le baryum. - on a applique ces resultats a des separations radiochimiques dans des solutions de produits de fission: separation de {sup 140}Ba, {sup 90}Sr, separation du groupe des

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-05-15

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

  2. Delayed fission

    Energy Technology Data Exchange (ETDEWEB)

    Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-07-01

    Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

  3. Performance Analysis of Fission and Surface Source Iteration Method for Domain Decomposed Monte Carlo Whole-Core Calculation

    International Nuclear Information System (INIS)

    Jo, Yu Gwon; Oh, Yoo Min; Park, Hyang Kyu; Park, Kang Soon; Cho, Nam Zin

    2016-01-01

    In this paper, two issues in the FSS iteration method, i.e., the waiting time for surface source data and the variance biases in local tallies are investigated for the domain decomposed, 3-D continuous-energy whole-core calculation. The fission sources are provided as usual, while the surface sources are provided by banking MC particles crossing local domain boundaries. The surface sources serve as boundary conditions for nonoverlapping local problems, so that each local problem can be solved independently. In this paper, two issues in the FSS iteration are investigated. One is quantifying the waiting time of processors to receive surface source data. By using nonblocking communication, 'time penalty' to wait for the arrival of the surface source data is reduced. The other important issue is underestimation of the sample variance of the tally because of additional inter-iteration correlations in surface sources. From the numerical results on a 3-D whole-core test problem, it is observed that the time penalty is negligible in the FSS iteration method and that the real variances of both pin powers and assembly powers are estimated by the HB method. For those purposes, three cases; Case 1 (1 local domain), Case 2 (4 local domains), Case 3 (16 local domains) are tested. For both Cases 2 and 3, the time penalties for waiting are negligible compared to the source-tracking times. However, for finer divisions of local domains, the loss of parallel efficiency caused by the different number of sources for local domains in symmetric locations becomes larger due to the stochastic errors in source distributions. For all test cases, the HB method very well estimates the real variances of local tallies. However, it is also noted that the real variances of local tallies estimated by the HB method show slightly smaller than the real variances obtained from 30 independent batch runs and the deviations become larger for finer divisions of local domains. The batch size used for the HB

  4. Delayed Neutrons and Photoneutrons from Fission Products; Neutrons Retardes et Photoneutrons Emis par des Produits de Fission; 0417 0410 041f 0410 0417 0414 042b 0412 0410 042e 0429 0418 0415 041d 0415 0419 0422 0420 041e 041d 042b 0418 0424 041e 0422 041e 041d 0415 0419 0422 0420 041e 041d 042b 041f 0420 041e 0414 0423 041a 0422 041e 0412 0414 0415 041b 0415 041d 0418 042f ; Neutrones Retardados y Fotoneutrones de los Productos de Fision

    Energy Technology Data Exchange (ETDEWEB)

    Amiel, S. [Soreq Nuclear Research Centre, Atomic Energy Commission, Yavne (Israel)

    1965-07-15

    photoneutron emission after fission represents the composite decay of the corresponding gamma- ray-emifting nuclides in the mixture of fission product chains. The gross photoneutron decay curves from both D{sub 2}O and beryllium have been resolved into nine groups with half-lives ranging from a few seconds to {approx}13 d. Each of these groups represents a complex composite of nuclides. Derailed information on the constituents of these groups is important for reactors in which fission products fractionation may take place and affect relative group abundances. Such information exists only for the longer-lived fission products whose properties and yields are well known, and where experimental results are in agreement with calculated values based on fission mass and charge distributions and decay schemes. The characterization of individual nuclides in the more complex short-lived groups is indeed a formidable task, but this information is of considerable practical and theoretical importance, and warrants further extensive study. (author) [French] Neutrons retardes: La plupart des etudes sur les neutrons retardes emis par des produits de fission impliquent l'analyse du comportement cinetique des systemes qui subissent une reaction de fission en chaine, l'analyse de la desintegration totale des neutrons (se resolvant en six groupes avec des periodes d'environ 0,2, 0,5, 2, 6, 22 et 55 s) et cettaines mesures des spectres de neutrons - ayant une energie comprise entre 0,1 et 1.2 MeV, avec des pics dans la gamme de 0,2 a 0,5 MeV. Des separations rapides des halogenes produits par fission ont mis en evidence la presence de sept isotopes ({sup 87,88,89,90}Br et {sup 137,138,139}I) et le dosage des gaz rares a permis d'identifier Kr (1,5 s) et Rb (6 s), comme precurseurs de neutrons retardes. Environ 80% des rendements totaux en neutrons retardes doivent etre attribues aux precurseurs ainsi identifies. A la suite de recherches theoriques, quelques dizaines de ces nucleides ont ete

  5. Use of electron beams for the production of radioactive nuclei through photo-fission; Utilisation de faisceaux d'electrons pour la production des noyaux radioactifs par photo-fission

    Energy Technology Data Exchange (ETDEWEB)

    M' garrech, Slah

    2004-09-01

    The IPN (institute of nuclear physics) of Orsay decided to build a linear accelerator in order to produce an electron beam of 50 MeV energy and of 10 {mu}A average intensity. It is the ALTO project (Linear Accelerator near the Tandem of Orsay). This project will be dedicated to the production of the radioactive ions using the photo-fission process. The central topic of this thesis is the study of the beam dynamics of the ALTO facility. The first part presents studies concerning the injector. The simulations made with the simulation code PARMELA allowed the optimization of the characteristics of pre-buncher (dephasing HF, accelerating field peak...) to obtain a good bunching factor at the entrance of the buncher and at the entrance of the accelerating section according to the distance separating the two systems. The second part of this thesis is related to measurements of transverse emittance of the beam at the buncher exit. The three gradients method has been selected and the optical system used is a solenoid. The results obtained are in good agreement with former measurements. Finally a calculation of the beam line was carried out to optimize the transport of the beam to the PARRNe target without degrading its characteristics. The calculation codes that have been used are BETA and TRACE-WIN. (author)

  6. Calculations of fission rate distribution in the core of WWER-1000 mock-up on the LR-0 reactor using alternative methods and comparison with results of measurements

    International Nuclear Information System (INIS)

    Zaritskiy, S.; Kovalishin, A.; Tsvetkov, T.; Rypar, V.; Svadlenkova, M.

    2011-01-01

    General review of experimental and calculation researches on WWER-440 and WWER-1000 mock-ups on the reactor LR-0 was introduced on the twentieth Symposium AER. The experimental core fission rate distribution was obtained by means of gamma-scanning of the fuel pins - 140La single peak (1596 keV) measurements and wide energy range (approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were scanned in different experiments. The measurements were arranged in the middle of the fuel (the active part of pin). Pin-to-pin calculations of the WWER-1000 mock-up core fission rate distribution were performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental ones. The obtained results allow developing the benchmark for core calculations methodologies, evaluating and validating source reliability for the out-of-core (inside and outside pressure vessel) neutron transport calculations. (Authors)

  7. Calculation programme for the accidental transients in reactors of the gas-graphite type; Programme de calcul des transitoires accidentels des piles de la filiere graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    The study of the behaviour of the fuel during certain incidents or accidents in reactors is closely connected to the study of the changes in temperature. This document describes in the first part the main physical phenomena governing the kinetics of the accident. The aim is to know the temperatures at all points and at all times during the irregular regime which can follow the initial stable regime. In the second part an explanation is given of the numerical methods used. (authors) [French] L'etude du comportement du combustible lors de certains incidents ou accidents de pile est etroitement liee a l'etude de l'evolution des temperatures. Dans sa premiere partie, ce document decrit les phenomenes physiques principaux intervenant dans la cinetique de l'accident. Le but recherche est la connaissance des temperatures en tout point et a tout instant d'un regime varie, faisant suite a un regime initial stable. Dans la deuxieme partie les methodes numeriques employees sont explicitees. (auteurs)

  8. Fission product yields

    International Nuclear Information System (INIS)

    Valenta, V.; Hep, J.

    1978-01-01

    Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

  9. Relativistic Coulomb Fission

    Science.gov (United States)

    Norbury, John W.

    1992-01-01

    Nuclear fission reactions induced by the electromagnetic field of relativistic nuclei are studied for energies relevant to present and future relativistic heavy ion accelerators. Cross sections are calculated for U-238 and Pu-239 fission induced by C-12, Si-28, Au-197, and U-238 projectiles. It is found that some of the cross sections can exceed 10 b.

  10. 10 Méthode optimale de calcul des courants de court-circuit dans un ...

    African Journals Online (AJOL)

    Oresta

    1. Introduction. L'optimisation de la gestion d'un réseau de distribution électrique est liée à une meilleure coordination des protections dans le réseau. Un bon plan de protection dans le réseau, permet en effet de réduire considérablement les temps d'interruption suite aux défauts et contribue aussi à la réduction des pertes.

  11. MODELES ET CALCUL DES SYSTEMES DE SUSPENSION MAGNETIQUE PASSIVE -DEVELOPPEMENTS ET CALCULS ANALYTIQUES EN 2D ET 3D DES INTERACTIONS ENTRE AIMANTS PERMANENTS-

    OpenAIRE

    Allag , Hicham

    2010-01-01

    The passive magnetic suspensions operate using attractive or repulsive forces exerted between permanent magnets. After giving an overview of different possible configurations of magnetic suspensions, the calculation of interactions between permanent magnets was developed. The calculations are realised for parallelepipeds magnets, where the magnetization is represented by distributions of charges or poles (Coulombian approach). All interactions (energy, forces, torques, etc ....) was calculate...

  12. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    International Nuclear Information System (INIS)

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs

  13. Intermediate energy nuclear fission

    International Nuclear Information System (INIS)

    Hylten, G.

    1982-01-01

    Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)

  14. New calculation for the neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

    International Nuclear Information System (INIS)

    Mesa, J.; Deppman, A.; Likhachev, V.P.; Arruda-Neto, J.D.T.; Manso, M.V.; Garcia, C.E.; Rodriguez, O.; Guzman, F.; Garcia, F.

    2003-01-01

    The 233 Pa(n,f) cross section, a key ingredient for fast reactors and accelerators driven systems, was measured recently with relatively good accuracy [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. The results are at strong variance with accepted evaluations and an existing indirect experiment. This circumstance led us to perform a quite detailed and complete evaluation of the 233 Pa(n,f) cross section between 1.0 and 3.0 MeV, where use of our newly developed routines for the parametrization of the nuclear surface and the calculation of deformation parameters and level densities (including low-energy discrete levels) were made. The results show good quantitative and excellent qualitative agreement with the experimental direct data obtained by Tovesson et al. [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. Additionally, our methodology opens new possibilities for the analysis of subthreshold fission and above threshold second-chance fission for both 233 Pa and its decay product 233 U, as well as other strategically important fissionable nuclides

  15. Contribution to the study of the thermodynamic properties of solutions using their phase diagrams (1961); Contribution au calcul des proprietes thermodynamiques des solutions a partir des diagrammes de phases (1961)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-10-15

    The thermodynamic study of the behaviour of solutions is of great interest in applied metallurgical problems, since the use of physical phenomena (solubility or volatility, for example) makes it possible to effect chemical reactions which, would not take place if the products formed did not mix. It is interesting to be able to predict this behaviour, at least for binary systems, using a knowledge of the phase diagrams. After showing the theoretical impossibility of resolving this problem without further data, an attempt is made to show what can be calculated from a knowledge of the phase diagrams: on the one hand it is possible to study the coherence between different types of data (calorimetric or equilibrium); on the other hand can be calculated the 'model' parameters, whether they be empirical, or statistically derived, and their validity can be checked. (author) [French] L'etude thermodynamique du comportement des solutions presente un grand interet dans les problemes de metallurgie appliquee puisque l'emploi de phenomenes physiques (solubilite ou volatilisation par exemple) permet de realiser des reactions chimiques qui seraient irrealisables si les produits formes ne se melangeaient pas. Il semble tres interessant de pouvoir prevoir ce comportement, tout au moins dans le cas de systemes binaires, a partir de la connaissance des diagrammes de phases. Apres avoir montre l'impossibilite theorique de resoudre ce probleme sans autres donnees, on essaie de voir ce que permet de calculer la connaissance des diagrammes de phases: d'une part, on peut etudier la coherenc entre divers types de donnees (calorimetrique ou d'equilibre); d'autre-part, on peut calculer les parametres de 'modeles', qu'ils soient empiriques ou a base statistique, et verifier leur validite. (auteur)

  16. Calculation of control rods in rectangular reactor, and applications (1960); Calcul des barres de conteole dans un reacteur rectangulaire et applications (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Goshen, S; Pazy, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    The aim of this report is to find a method for estimating the anti-reactivity of control rods perpendicular to the axis in a cylindrical pile. The paper is divided into two parts. In the first is given a method of calculating control rods in a rectangular pile, similar to the Nordheim-Scalettar method for cylindrical piles. As an example the formulas are given for the theories of one and two neutron groups, the generalisation for several groups being evident. In the second part we find by a variation method a formula for estimating the Laplacian of a pile, which may be divided into parallelepipeds for which the Laplacian are given. Finally, this formula is used to calculate the anti-reactivity of rods perpendicular to the axis in a cylindrical pile. (author) [French] Le but de ce rapport est de trouver une methode pour estimer l'antireactivite des barres de controle perpendiculaires a l'axe dans pile cylindrique. Le rapport se divise en deux parties. Dans la premiere nous donnons une methode de calcul des barres de controle dans une pile rectangulaire, analogue a la methode de Nordheim-Scalettar pour les piles cylindriques. A titre d'exemple, nous donnons les formules de theories a un et deux groupes de neutrons, la generalisation pour plusieurs groupes est evidente. Dans la deuxieme partie, nous trouvons, par une methode de variation, une formule qui permet d'estimer le laplacien d'une pile, qui peut etre divisee en parallelepipedes dont les laplaciens sont donnes. Nous utilisons enfin, cette formule pour calculer l'antireactivite des barres perpendiculaires a l'axe dans une pile cylindrique. (auteur)

  17. Fission-Product Separation as a Final Solution to the Problem of Storing Highly Radioactive Waste; La Separation des Produits de Fission Comme Solution Definitive du Stockage des Dechets Fortement Radioactifs; 0412 042b 0414 ; La Separacion de los Productos de Fision Como Solucion Definitiva del Problema del Almacenamiento de los Desechos de Elevada Radiactividad

    Energy Technology Data Exchange (ETDEWEB)

    Raggenbass, A.; Lefevre, J. [Centre d' Etudes Nucleaires, Saclay (France)

    1963-02-15

    'isolement du cesium-137 et du strontium-90 avec un facteur de decontamination d'au moins 10{sup 4}, ce qui permettrait d'envisager le rejet, apres un traitement d'effluents classique, au bout d'une periode de 10 a 20 ans. Le but du memoire est de faire le point de cette question en l'annee 1962. Les progres realises dans les methodes de separation du cesium et du strontium permettent d'envisager l'obtention de ce facteur 10{sup 4} dans des conditions exploitables industriellement. Des methodes pleines de promesses sont: a) Pour le cesium-137, la precipitation par l'acide phosphotungstique, l'echange sur un sel d'ammonium d'un heteropolyacide, la fixation par un phosphate de zirconium. b) Pour le strontium-90, la precipitation d'oxalate, l'extraction par l'acide di-2-ethyl-hexylphosphorique, l'echange sur des sels varies, sulfate de strontium, etc. Ces progres chimiques sont renforces par les progres realises dans le traitement industriel des effluents et permettent de considerer comme techniquement possible la solution preconisee par Glueckauf. Parallelement, le developpement de l'utilisation de grandes quantites de produits de fission, particulierement dans les programmes SNAP, justifie des investissements qui seront au moins pairtiellement recuperes par la vente des produits fabriques. Il est certain que des sources refractaires, a resistance tres elevee a tous les agents d'agression exterieurs, constituent le mode de stockage le plus satisfaisant pour des produits aussi dangereux et d'aussi longue periode. L'auteur considere done, techniquement et psychologiquement, Inverted-Exclamation-Mark /utilisation de sources concentrees a gainage riche comme le moyen de stockage presentant le plus de garanties. Une discussion comparative de cette methode avec la methode d'incorporation dans les verres meriterait un examen complet et approfondi. (author) [Spanish] En la Conferencia de Ginebra de 1955, Glueckauf formplo la hipotesis de que los problemas planteados por las soluciones de

  18. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1992-01-01

    Classical dynamical calculations of the heavy ion induced fission processes have been performed for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus. As a result prescission lifetimes were obtained and compared with the experimental values. The comparison between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. (orig.)

  19. Nuclear fission and neutron-induced fission cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

    1981-01-01

    A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

  20. Development of an experimental device based on the digitalization of the signal and dedicated to the characterization of fission fragments and prompt neutrons; Developpement d'un dispositif experimental base sur la digitalisation des signaux et dedie a la caracterisation des fragments de fission et des neutrons prompts emis

    Energy Technology Data Exchange (ETDEWEB)

    Varapai, N

    2006-12-15

    The present work demonstrates the application of the digital technique for nuclear measurements. This new technique is based on the digitalization of the signals from the detectors and has several advantages. This technique allows us to extract the maximum amount of information contained in the signal shape. In the case of an ionization chamber this signal contains the necessary information on the particle kinetic energy, emission angle and mass. This method has been implemented for measurements of promptly emitted fission neutrons in coincidence with fission fragments from {sup 252}Cf(sf). A double Frisch-grid ionization chamber is used as fission fragment detector. The promptly emitted neutrons are detected by a NE213 liquid scintillation detector. This work displays how delicate analysis of the digitalized signals permitted us to infer the mass and kinetic energy distributions of the fission fragments as well as the neutron energy spectrum and multiplicity. The outline of this thesis is as follows: Chapter 2 gives an overview of the experimental tools used in this work. Chapter 3 explains the analysis procedure of the digitalized anode signal from an ionization chamber. Chapter 4 gives a detailed explanation of the analysis procedure of the digitalized signal from a neutron detector. In Chapter 5 the analysis procedure of the fission fragment events in coincidence with neutrons is given.

  1. Nuclear fission

    International Nuclear Information System (INIS)

    Kodama, T.

    1981-01-01

    The nuclear fission process is pedagogically reviewed from a macroscopic-microscopic point of view. The Droplet model is considered. The fission dynamics is discussed utilizing path integrals and semiclassical methods. (L.C.) [pt

  2. Fission in a Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Younes, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A three-year theory project was undertaken to study the fission process in extreme astrophysical environments, such as the crust of neutron stars. In the first part of the project, the effect of electron screening on the fission process was explored using a microscopic approach. For the first time, these calculations were carried out to the breaking point of the nucleus. In the second part of the project, the population of the fissioning nucleus was calculated within the same microscopic framework. These types of calculations are extremely computer-intensive and have seldom been applied to heavy deformed nuclei, such as fissioning actinides. The results, tools and methodologies produced in this work will be of interest to both the basic-science and nuclear-data communities.

  3. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    International Nuclear Information System (INIS)

    Vathonne, Emerson

    2014-01-01

    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  4. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    International Nuclear Information System (INIS)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C.

    2017-01-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S N ) formulation and the response matrix method was applied to solve the forward and the adjoint S N problems. Numerical results are given to verify the present. (author)

  5. Calculation of neutron interior source distribution within subcritical fission-chain reacting systems for a prescribed power density generation

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Leonardo R.C.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: lrcmoraes@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: ricardob@iprj.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pós-Graduação em Modelagem Computacional

    2017-07-01

    Accelerator Driven Systems (ADS) are sub-critical systems stabilized by stationary external sources of neutrons. A system is subcritical when the removal by absorption and leakage exceeds the production by fission and tends to shut down. On the other hand, any subcritical system can be stabilized by including time-independent external sources of neutrons. The goal of this work is to determine the intensity of uniform and isotropic sources of neutrons that must be added inside all fuel regions of a subcritical system so that it becomes stabilized, generating a prescribed distribution of electric power. A computer program has been developed in Java language to estimate the intensity of stationary sources of neutrons that must be included in the fuel regions to drive the subcritical system with a fixed power distribution prescribed by the user. The mathematical model used to achieve this goal was the energy multigroup, slab-geometry neutron transport equation in the discrete ordinates (S{sub N}) formulation and the response matrix method was applied to solve the forward and the adjoint S{sub N} problems. Numerical results are given to verify the present. (author)

  6. The Prompt Gamma-Ray, Prompt Electron and Prompt X-Ray Spectra Associated with Fission Fragments of Specific Mass; Spectres de Rayons Gamma Instantanes, d'Electrons Instantanes et de Rayons X Instantanes Associes a des Fragments de Fission de Masse Donnee; 041c 0413 041d 041e 0414 ; Espectros de Rayos Gamma. Electrones y Rayos X Inmediatos, Vinculados a Fragmentos de Fision de Masa Determinada

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, H. R.; Thompson, S. G.; Watson, R. L.; Kapoor, S. S.; Rasmussen, J. O. [Lawrence Radiation Laboratory, University of California, Berkeley, CA (United States)

    1965-07-15

    Well-defined prompt gamma-rays, prompt-conversion elections and prompt K-X-rays have been observed in coincidence with moving fission fragments of Cf{sup 252}. In a few cases, the masses and charges of the nuclei emitting the gamma-rays and conversion electrons have been identified. The gamma-ray, prompt-electron and prompt X-ray energies as well as the two fission fragments energies were measured with high-resolution solid-state detectors. The masses of the fragments were deduced from their energies, and the nuclear charges were determined by measuring the K-X - ray energies associated with different masses. The magnitude and sign of the Doppler shift in gamma-ray energy allowed assignment of the gamma-ray lines to single members of fragment pairs. The Doppler shift also provides an independent measure of the fragment velocity and hence the fragment mass after neutron emission. The results of the X-ray measurements are consistent with the view that the majority of the prompt X-rays emitted during the spontaneous fission of Cf{sup 252} are the result of internal conversion during the de-excitation of low-energy collective states of the primary fission fragments. Apart from the specific results discussed above, the most important consequence of these experiments has been the demonstration that it is possible to study the properties of individual fission fragments, as identified by their characteristic radiations, rather than studying the properties of an average fission fragment with an average mass and charge. The consequences of this advance in the technique of studying fission fragments ate being explored. (author) [French] Des rayons gamma instantanes, des electrons de conversions instantanes et des rayons X K instantanes bien definis ont ete observes en coincidences avec des fragments de fission de {sup 252}Cf en mouvement. Dans un petit nombre de cas' Les masses, et charges des noyaux emettant les rayons gamma et les electrons de conversion ont ete identifiees

  7. 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

  8. Theory of nuclear fission: a review

    International Nuclear Information System (INIS)

    Mosel, U.

    1976-01-01

    General properties of nuclear fission are reviewed and related to our present knowledge of fission theory. For this purpose the basic reasons for the shape of the fission barriers are discussed and their consequences compared with experimental results on barrier shapes and structures. Special emphasis is put on the asymmetry of the fission barriers and mass-distributions and its relation to the shells of the nascent fragment shells. Finally the problem of calculating fission cross sections is discussed

  9. A new statistical scission-point model fed with microscopic ingredients to predict fission fragments distributions; Developpement d'un nouveau modele de point de scission base sur des ingredients microscopiques

    Energy Technology Data Exchange (ETDEWEB)

    Heinrich, S

    2006-07-01

    Nucleus fission process is a very complex phenomenon and, even nowadays, no realistic models describing the overall process are available. The work presented here deals with a theoretical description of fission fragments distributions in mass, charge, energy and deformation. We have reconsidered and updated the B.D. Wilking Scission Point model. Our purpose was to test if this statistic model applied at the scission point and by introducing new results of modern microscopic calculations allows to describe quantitatively the fission fragments distributions. We calculate the surface energy available at the scission point as a function of the fragments deformations. This surface is obtained from a Hartree Fock Bogoliubov microscopic calculation which guarantee a realistic description of the potential dependence on the deformation for each fragment. The statistic balance is described by the level densities of the fragment. We have tried to avoid as much as possible the input of empirical parameters in the model. Our only parameter, the distance between each fragment at the scission point, is discussed by comparison with scission configuration obtained from full dynamical microscopic calculations. Also, the comparison between our results and experimental data is very satisfying and allow us to discuss the success and limitations of our approach. We finally proposed ideas to improve the model, in particular by applying dynamical corrections. (author)

  10. [Fission product yields of 60 fissioning reactions]. Final report

    International Nuclear Information System (INIS)

    Rider, B.F.

    1995-01-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  11. Measurement of the fission cross section induced by fast neutrons of the {sup 232}Th/{sup 233}U nuclei within the innovating fuel cycles framework; Mesure de la section efficace de fission induite par neutrons rapides des noyaux {sup 232}Th/{sup 233}U dans le cadre des cycles de combustible innovants

    Energy Technology Data Exchange (ETDEWEB)

    Grosjean, C

    2005-03-15

    The thorium-U{sup 233} fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (Th{sup 232}, Pa{sup 233} and U{sup 233}) at the level of those for the U-Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of Th{sup 232} and U{sup 233} has been measured from 1 to 7 MeV with a target accuracy lesser than 5 per cent. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n,p). The cross section of which is very well known in a large neutron energy domain ({approx} 0,5 % from 1 eV to 50 MeV) compared to the U{sup 235}(n,f) reaction. This technique has been applied for the first time to the Th{sup 232}(n,f) and U{sup 233}(n,f) cases. A Hauser-Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus U{sup 234} from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of U{sup 233}. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering U{sup 233}(n,n'), radiative capture U{sup 233}(n,{gamma}) and U{sup 233}(n,2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values have allowed us to fill the lack of experimental data for the major fissile nucleus of the thorium cycle. (author)

  12. 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)

  13. The design of coils for the production of high homogeneous fields; Calcul des bobinages pour la production de champs magnetiques tres homogenes

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    The discovery of type II superconductors has considerably increased the possibilities of air-core coils, in particular with regard to the production of high homogeneous fields. The design of such magnets,calls for elaborate calculations which, in practise, can only be carried out on computers. The present report describes a complete set of programs for the calculation, in the case of cylindrical systems, of the magnetic field components at any point, the lines of flux, the forces, the self and mutual inductances, as well as the design of compensated coils for the production of high homogeneous fields. These programs have been employed for the calculation of two magnets which are described in detail. (author) [French] L'interet des bobines sans fer s'est considerablement accru depuis l'apparition recente des supraconducteurs de la deuxieme espece, en particulier pour la realisation d'aimants a champ tres homogene. Le calcul de tels bobinages fait appel a des methodes complexes dont l'execution pratique necessite l'emploi de machines a calculer. Le present rapport decrit un ensemble de programmes permettant de calculer, dans le cas de systemes de revolution de structure quelconque, le champ dans tout l'espace, les lignes de force du champ, les efforts electromagnetiques, les selfs et mutuelles, et de determiner des enroulements de compensation destines a uniformiser le champ. Ces programmes ont servi au calcul de deux aimants particuliers dont les caracteristiques detaillees sont fournies a titre d'application. (auteur)

  14. Fission in Empire-II version 2.19 beta1, Lodi

    International Nuclear Information System (INIS)

    Sin, M.

    2003-01-01

    This is a description of the fission model implemented presently in EMPIRE-II. This package offers two ways to calculate the fission probability selected by parameters in the optional input. Fission barriers, fission transmission coefficients, fission cross sections and fission files are calculated

  15. Integral parameters for the Godiva benchmark calculated by using theoretical and adjusted fission spectra of 235U

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1987-05-01

    The theoretical and adjusted Watt spectrum representations for 235 U are used as weighting functions to calculate K eff and θ f 28 /θ f 25 for the benchmark Godiva. The results obtained show that the values of K eff and θ f 28 /θ f 25 are not affected by spectrum form change. (author) [pt

  16. LEAF: a computer program to calculate fission product release from a reactor containment building for arbitrary radioactive decay chains

    International Nuclear Information System (INIS)

    Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.

    1976-10-01

    The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building

  17. Sur deux mémoires de d'Alembert l'un concernant le calcul des probabilités, l'autre l'inoculation

    CERN Document Server

    Diderot, Denis

    2015-01-01

    Extrait : ""M. d'Alembert vient de publier ses Opuscules mathématiques. Il y a dans ce recueil deux mémoires qu'il n'est pas impossible de réduire à la langue ordinaire de la raison. L'un a pour objet le calcul des probabilités ; calcul dont l'application a tant d'importance et d'étendu. C'est proprement la science physico-mathématique de la vie. L'autre traite des avantages ou désavantages de l'inoculation.""

  18. 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

  19. Metabolism of fission products in man: Marshallese experience; Metabolisme des produits de fission radioactifs chez l'homme: donnees recueillies aux iles Marshall; Metabolizm radioaktivnykh produktov deleniya v organizme cheloveka. (dannye obsledovaniya zhite- lej marshal'skikh ostrovov); Metabolismo de productos de fision radiactivos en el hombre: datos obtenidos en los habitantes de las islas Marshall

    Energy Technology Data Exchange (ETDEWEB)

    Cohn, S. H. [Brookhaven National Laboratory, Upton, Long Island, NY (United States)

    1963-02-15

    The medical study of the Marshallese accidentally exposed to local fall-out in 1954 is unique in that; along with he Japanese fishermen study, it provides the only data existing on the metabolism of mixed fission products in a human population. Early diagnosis of the internal radioactive contamination was made by radiochemical analysis of the excreta of the exposed people and by radiochemical analysis of the tissues of animals simultaneously exposed. Initially, Sr{sup 89}, Ba{sup 140}, I{sup 131} and its shorter-lived daughters and a number of rare-earth elements contributed the major portion of the internal radiation dose. After a year, the principal radioisotopes were Sr{sup 90}, Cs{sup 137} and Zn{sup 65}. Subsequently these radionuclides and more recently, Co{sup 60}, have been measured periodically. Since 1958, the {gamma}-spectra of a number of Marshallese have been obtained with a portable whole-body counter. The report discusses the findings of these studies for the past eight years. The results of an early attempt o alter the rate of removal of the mixed fission products in the Marshallese with calcium disodium EDTA are presented. The metabolism of die radionuclides and their relationship to levels present in the environment is also discussed. (author) [French] L'enquete medicale a laquelle ont ete soumis les habitants des iles Marshall qui, en 1954, ont ete exposes accidentellement a des retombees locales presente un caractere unique en ce sens que, tout comme l'enquete sur les pecheurs japonais, elle fournit les seules donnees que l'on possede sur le metabolisme d'un melange de produits de fission chez l'homme. Un premier diagnostic de la contamination radioactive interne a ete fait par analyse radiochimique des excreta des sujets exposes et par analyse radiochimique des tissus des animaux exposes simultanement. Au debut, {sup 89}Sr, {sup 140}Ba, {sup 131}I et leurs produits de filiation de courte periode,ainsi qu'un certain nombre de terres rares

  20. Analyse des erreurs dans les calculs sur ordinateurs Error Analysis in Computing

    Directory of Open Access Journals (Sweden)

    Vignes J.

    2006-11-01

    Full Text Available La méthode présentée ici permet d'évaluer l'erreur sur les résultats d'algorithmes, erreurs dues à l'arithmétique à précision limitée de la machines L'idée de base de cette méthode est qu'à un algorithme donné fournissant un résultat algébrique unique r, correspond en informatique un ensemble R de résultats numériques qui sont tous représentatifs de résultat exact r. La méthode de permutation-perturbation que nous présentons ici permet d'obtenir les éléments de R. La perturbation agit sur les données et résultats de chaque opération élémentaire. La permutation agit sur l'ordre d'exécution des opérations. Une étude statistique des éléments de R permet d'estimer l'erreur commise. Dans la pratique, il suffit de 2 ou 3 éléments de R pour connaître cette erreur. This paper describes a new method for evaluating the error in the results of computation of an algorithm. The basic idea underlying the method is that while in algebra a given algorithm provides a single result r, this same algorithm carried out on a computer provides a set R of numerical results that are ail representative of the exact algebraic result r. The permutation-perturbation method described here can be used to obtain the elements of R. The perturbation acts on the data and results of each elementary operation, and the permutation acts on the order in which operations are carried out. A statistical analysis of the elements of R is performed to determine the error committed. In practice, 2 to 4 R elements are sufficient for determining the error.

  1. A method to calculate fission-fragment yields Y(Z,N) versus proton and neutron number in the Brownian shape-motion model. Application to calculations of U and Pu charge yields

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, Peter [Los Alamos National Laboratory, Theoretical Division, Los Alamos, NM (United States); Ichikawa, Takatoshi [Kyoto University, Yukawa Institute for Theoretical Physics, Kyoto (Japan)

    2015-12-15

    We propose a method to calculate the two-dimensional (2D) fission-fragment yield Y(Z,N) versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use the Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment Q{sub 2}), neck d, left nascent fragment spheroidal deformation ε{sub f1}, right nascent fragment deformation ε{sub f2} and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalized potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of Z and N of the compound system and its shape, including the asymmetry of the shape. We outline here how to generalize the model from the ''compound-system'' model to a model where the emerging fragment proton and neutron numbers also enter, over and above the compound system composition. (orig.)

  2. Post-scission fission theory: Neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1997-01-01

    A survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N (E) and the average prompt neutron multiplicity bar ν p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the recent models. In particular, the dependencies of N (E) and bar ν p upon the fissioning nucleus and its excitation energy are treated. Recent work in the calculation of the prompt fission neutron spectrum matrix N (E, E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of our ability to calculate these observables with confidence, the direction of future theoretical efforts, and limitations to current (and future) approaches

  3. Mass and Charge Distribution in Low-Energy Fission; Repartition des Masses et des Charges dans la Fission a Basse energie; 0420 0410 0421 041f 0420 0415 0414 0415 041b 0415 041d 0418 0415 041c 0410 0421 0421 042b 0418 0417 0410 0420 042f 0414 0410 041f 0420 0418 0414 0415 041b 0415 041d 0418 0418 , 0412 042b 0417 0412 0410 041d 041d 041e 041c 0427 0410 0421 0422 0418 0426 0410 041c 0418 041c 0410 041b 041e 0419 042d 041d 0415 0420 0413 0418 0418 ; Distribucion de Masas y de Cargas en la Fision a Bajas Energias

    Energy Technology Data Exchange (ETDEWEB)

    Wahl, A. C. [Washington University, St. Louis, MO (United States)

    1965-07-15

    repartition des masses et des charges dans la fission de {sup 235}U par neutrons thermiques et la compare ensuite a la repartition correspondante dans les autres processus de fission a basse energie. Les questions a examiner en ce qui concerne la repartition des masses dans la fission comprennent les positions des sommets et de la vallee et la structure fine de la distribution des produits de fission Selon la masse pour des couches fermees, ainsi que les changements de position qui se produisent lorsque le noyau subissant la fission et l'energie d'excitation ne sont pas les memes. L'auteur examine egalement la repartition des masses dans le cas de la fission ternaire. Au sujet de la fission binaire et de la fission ternaire, il fait des observations touchant la repartition des masses entre les fragments primaires (avant 'evaporation' des neutrons) et les produits de fission (apres 'evaporation' des neutrons). L'auteur etudie la repartition des charges d'apres la dispersion des charges entre les produits de fission ayant le meme nombre de masse, et la variation de Zp Selon la masse, la 'charge la plus probable' (non entiere) pour une.masse donnee. Bien que des renseignements directs sur la repartition des charges soient limites aux produits de fission, l'auteur presente des estimations de la repartition des charges pour les fragments de fission primaires. D'apres des donnees et des estimations relatives a la repartition des masses et des charges pour un procede de fission, on peut determiner les rendements primaires en tous produits ou fragments de fission. Bien que l'estimation de nombreux rendements primaires soit tout a fait approximative en raison surtout de l'absence de donnees sur la repartition des charges, notamment pour les produits de fission dont le rendement est faible, l'auteur presente des estimations des rendements primaires pour illustrer la necessite d'une nouvelle experimentation et ses possibilites pratiques de realisation. Les processus de fission - autres

  4. Experimental validation of calculated capture rate for nucleus involved in fuel cycle; Validation experimentale du calcul du taux de capture des noyaux intervenant dans le cycle du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Benslimane-Bouland, A

    1997-09-01

    The framework of this study was the evaluation of the nuclear data requirements for Actinides and Fission Products applied to current nuclear reactors as well as future applications. This last item includes extended irradiation campaigns, 100 % Mixed Oxide fuel, transmutation or even incineration. The first part of this study presents different types of integral measurements which are available for capture rate measurements, as well as the methods used for reactor core calculation route design and nuclear data library validation. The second section concerns the analysis of three specific irradiation experiments. The results have shown the extent of the current knowledge on nuclear data as well as the associated uncertainties. The third and last section shows both the coherency between all the results, and the statistical method applied for nuclear data library adjustment. A relevant application of this method has demonstrated that only specifically chosen integral experiments can be of use for the validation of nuclear data libraries. The conclusion is reached that even if co-ordinated efforts between reactor and nuclear physicists have made possible a huge improvement in the knowledge of capture cross sections of the main nuclei such as uranium and plutonium, some improvements are currently necessary for the minor actinides (Np, Am and Cm). Both integral and differential measurements are recommended to improve the knowledge of minor actinide cross sections. As far as integral experiments are concerned, a set of criteria to be followed during the experimental conception have been defined in order to both reduce the number of required calculation approximations, and to increase as much as possible the maximum amount of extracted information. (author)

  5. Irradiation effects and diffusion of fission products (cesium and iodine) in silicon carbide; Effets d'irradiation et diffusion des produits de fission (cesium et iode) dans le carbure de silicium

    Energy Technology Data Exchange (ETDEWEB)

    Audren, A

    2007-03-15

    Silicon carbide is envisaged as a cladding material for the nuclear fuel in the fourth generation reactors. The aim of this work is to study the capacity to retain fission products and the structure evolution of this material under the combined effects of temperature and irradiation. The low energy ion implantations and the incorporation of stable analogues of fission products (Cs and I) in single crystalline 6H-SiC samples were performed by using the ion implanter or the accelerator of the CSNSM. The high energy heavy ion irradiations were made at GANIL. The evolution of the implanted ion profiles and the crystal structure were studied by RBS and Channeling. Complementary information were obtained by using the UV-visible absorption spectroscopy. The low energy ion implantations at room temperature induce a fast structural damage in the crystal. On the other hand, it is possible to attain a small disorder rate in the crystal during implantation by increasing the implantation temperature (600 C). The high energy heavy ion irradiations do not damage the SiC crystals. On the contrary, they cause an annealing of the disorder created by the low energy implantations. The implanted ions (I) do not diffuse during low or high energy ion irradiations at room temperature and at 600 C. However, a diffusion of Cs ions was observed during a post-implantation annealing at 1300 C. At this temperature, the crystal which had an extended amorphous layer starts to recover a single-crystal structure. (author)

  6. Study of the retention of fission products by a few common minerals. Application to the treatment of medium activity effluents (1962); Etude de la retention des produits de fission par quelques mineraux usuels. Application aux traitements d'effluents de moyenne activite specifique (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Auchapt, J M [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

    1962-07-01

    The conditions in which strontium is fixed on calcite (the object of the Geneva report P/395 - USA - 1958) are more closely studied and the work is extended to five fission products present in the effluents, and to 17 common rocks and minerals. Although as it turns out this fixation is not suitable as a method of treating the S.T.E. effluents (i.e. those from the Effluent Treatment plant at Marcoule), the study shows that all the crystals considered are strongly contaminated by simple contact with the effluents. (author) [French] Les conditions de fixation du strontium sur la calcite (objet du rapport de Geneve P/395 - USA - 1958) sont approfondies et l'etude est etendue aux cinq produits de fission presents dans les effluents et a 17 roches et minerais courants. Bien qu'en definitive cette fixation se revele inutilisable comme procede de traitement d'effluents S.T.E. (Station de Traitement des Effluents, Marcoule), l'etude montre que tous les cristaux consideres se contaminent fortement par simple contact avec les effluents. (auteur)

  7. Développement de la méthode de Monte Carlo pour le calcul des ...

    African Journals Online (AJOL)

    In this paper; we show the interest of the heterostructures initially, then the need for using a numerical method and in particular that of Monte Carlo, to calculate electric transport in the semiconductors. We justify also the composition of our ternary semiconductor AlxGa1-xAs. Afterwards; we give the principle and the ...

  8. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Science.gov (United States)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    -Z mini-plate fuel model was developed. The Y-Z model divides each fuel plate into 30 equal intervals in both the Y and Z directions. The MCNP-calculated results and the detailed Y-Z fission power mapping were used to help design the AFIP fuel test assembly to demonstrate that the AFIP test assembly thermal-hydraulic limits will not exceed the ATR safety limits.

  9. ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

  10. Coulomb fission and transfer fission at heavy ion collisions

    International Nuclear Information System (INIS)

    Himmele, G.

    1981-01-01

    In the present thesis the first direct evidence of nuclear fission after inelastic scattering of heavy ions (sup(183,184)W, 152 Sm → 238 U; 184 W → 232 Th; 184 W, 232 Th → 248 Cm) is reported. Experiments which were performed at the UNILAC of the Gesellschaft fuer Schwerionenforschung in Darmstadt show the observed heavy ion induced fission possesses significant properties of the Coulomb fission. The observed dependence of the fission probability for inelastic scattering on the projectile charge proves that the nuclear fission is mediated by the electromagnetic interaction between heavy ions. This result suggests moreover a multiple Coulomb-excitation preceding the fission. Model calculations give a first indication, that the Coulomb fission proceeds mainly from the higher β phonons. In the irradiation with 184 W the fission probability of 232 Th is for all incident energies about 40% smaller that at 238 U. The target dependence of the Coulomb fission however doesn't allow, to give quantitative statements about the position and B(E2)-values of higher lying β phonons. (orig./HSI) [de

  11. Accounting Systems for Heavy Water and Fissionable Materials; Comptabilite de l'Eau Lourde et des Matieres Fissiles; Sistema ucheta tyazheloj vody i delyashchikhsya materialov; Sistemas de Contabilidad para el Agua Pesada y los Materiales Fisionables

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, G. W.; Reid, H. B.; Jenkinson, W. G. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1966-02-15

    Detailed accounting and reporting procedures used by Atomic Energy of Canada Limited (AECL) for maintaining adequate records and control of heavy water supplies and stocks of fissionable materials are described, along with the duties and responsibilities of those administering the system. An appraisal is made of these procedures with respect to their adaptability for use in rapidly expanding research and power programmes. In particular the use of electronic data processing equipment is evaluated. A senior management committee is responsible for ensuring that there is a proper system for recording, reporting and controlling fissionable materials. The Production Planning and Control Branch (Pp and C B) of the Operations Division at the Chalk River Nuclear Laboratories (CRNL) is responsible to the committee for keeping the over-all records and for the general administration of the system. The duties involved are detailed in the report. The system for fissionable materials is segregated into several accountability units 15 of which are allocated to AECL departments and the others to Canadian industries and research organizations. A control ledger is kept by PP and CB for each of the units; however, the units are responsible for preparing detailed accounts of all material under their jurisdiction. The basic recording procedures covering the movement Of materials between units, the changing of forms within units, the handling of gains and losses, and disposals, are outlined in the report. The transfer of this data to IBM cards, the ultimate processing through an IBM 1401 computer and the preparation of reports for management approval are described. The heavy-water accounting system based on the same principles as used for the fissionable materials is explained. In this case the control ledger lists the pounds of D{sub 2}O allocated to each of the 15 accountability units. Again the basic recording methods and the use of a computer system are outlined. (author) [French

  12. Tip model of cold fission

    International Nuclear Information System (INIS)

    Goennenwein, F.; Boersig, B.

    1991-01-01

    Cold fission is defined to be the limiting case of nuclear fission where virtually all of the available energy is converted into the total kinetic energy of the fragments. The fragments have, therefore, to be born in or at least close to their respective ground states. Starting from the viewpoint that cold fission corresponds to most compact scission configurations, energy constraints have been exploited to calculate minimum tip distances between the two nascent fragments in binary fission. Crucial input parameters to this tip model of cold fission are the ground-state deformations of fragment nuclei. It is shown that the minimum tip distances being compatible with energy conservation vary strongly with both the mass and charge fragmentation of the fission prone nucleus. The tip distances refer to nuclei with equivalent sharp surfaces. In keeping with the size of the surface width of leptodermous nuclei, only configurations where the tip distances are smaller than a few fm may be considered as valid scission configurations. From a comparison with experimental data on cold fission this critical tip distance appears to be 3.0 fm for the model parameters chosen. Whenever the model calculation yields tip distances being smaller than the critical value, a necessary condition for attaining cold fission is considered to be fulfilled. It is shown that this criterion allows to understand in fair agreement with experiment which mass fragmentations are susceptible to lead to cold fission and which fragment-charge divisions are the most favored in each isobaric mass chain. Being based merely on energy arguments, the model cannot aim at predicting fragment yields in cold fission. However, the tip model proposed appears well suited to delineate the phase space where cold fission phenomena may come into sight. (orig.)

  13. Predictions of steam generator soiling - experiment, calculation, numerical simulation; Vorhersage des Ansatzverhaltens in Dampferzeugern - Experiment, Berechnung und numerische Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, W.; Hildebrand, V.; Ullrich, B.; Szilagyi, J.

    2000-07-01

    Power plant reliability depends to a large extent on steam generator operation. Investigations were therefore carried out for early detection of combustion chamber soiling at different operating conditions. Apart from CFD numerical simulations of the gaseous phase and particle tracks, both in the original and coupled with thermochemical calculations, emphasis was given to validations with measured values in the original plant, in combustion chamber models, and in experimental facilities. Three different fuels with different composition and behaviour were investigated, i.e. oil shales, mixed coal, and coal pretreated with sorption agents. Apart from the influence of the temperature and reaction conditions (oxygen concentration) in the vicinity of the particle, also different times of residue were investigated. [German] Die Betriebssicherheit von Kraftanlagen wird massgeblich von der Arbeitsweise der Dampferzeuger bestimmt. Ziel dieses Forschungskomplexes ist es, fruehzeitig Aussagen zum Ansatzverhalten im Feuerraum bei verschiedenen Betriebsbedingungen zu treffen. Neben der numerischen Simulation der Gasphase und der Partikelbahnen mit CFD-Codes im Original und der Kopplung mit thermochemischen Berechnungen stehen vor allem Validierungen mit Messwerten am Original, an Feuerraummodellen und Versuchsanlagen im Vordergrund der Arbeiten. Die hier vorgestellten Untersuchungen erfolgten an drei unterschiedlichen Brennstoffen, die sich sowohl in der Zusammensetzung als auch im Einsatz stark voneinander unterscheiden. Dies sind: Oelschiefer, Mischkohle und vorbehandelte Kohle mit Sorbent. Dabei werden neben den Einfluessen der Temperatur und der Reaktionsbedingung (Sauerstoffgehalt) in der Umgebung des Partikels auch verschiedene Verweilzeiten untersucht. (orig./AKF)

  14. Use of an oscillation technique to measure effective cross-sections of fissionable samples in critical assemblies; Mesure des sections efficaces effectives d'echantillons fissiles par une methode d'oscillation dans les-assemblages critiques

    Energy Technology Data Exchange (ETDEWEB)

    Tretiakoff, O; Vidal, R; Carre, J C; Robin, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors describe the technique used to measure the effective absorption and neutron-yield cross-sections of a fissionable sample. These two values are determined by analysing the signals due to the variation in reactivity (over-all signal) and the local perturbation in the flux (local signal) produced by the oscillating sample. These signals are standardized by means of a set of samples containing quantities of fissionable material ({sup 235}U) and an absorber, boron, which are well known. The measurements are made for different neutron spectra characterized by lattice parameters which constitute the central zone within which the sample moves. This technique is used to study the effective cross-sections of uranium-plutonium alloys for different heavy-water and graphite lattices in the MINERVE and MARIUS critical assemblies. The same experiments are carried out on fuel samples of different irradiations in order to determine the evolution of effective cross-sections as a function of the spectrum and the irradiations. (authors) [French] On decrit la methode utilisee pour mesurer les sections efficaces effectives d'absorption et de production de neutrons d'un echantillon fissile. Ces deux grandeurs sont determinees en analysant les signaux dus a la variation de reactivite (signal global) et a la perturbation locale de flux (signal local) produits par l'echantillon oscillant. Ces signaux sont etalonnes a l'aide d'un jeu d'echantillons dont les teneurs en materiau fissile ({sup 235}U) et en absorbeur (bore) sont bien connues. Les mesures sont realisees pour differents spectres de neutrons caracterises par les parametres du reseau constituant la zone centrale a l'interieur de laquelle se deplace l'echantillon. A l'aide de cette methode on etudie les sections efficaces effectives d'alliage uranium-plutonium pour differents reseaux a eau lourde et a graphite dans les assemblages crtiques MINERVE et MARIUS. Les memes experiences sont effectuees sur des echantillons de

  15. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

    The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

  16. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Weissfloch, R.

    The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

  17. Reconstruction et identification des électrons dans l'expérience ATLAS Participation à la mise en place d'un Tier 2 de la grille de calcul

    CERN Document Server

    Derue, Frederic

    2008-01-01

    L'origine de la masse des particules élémentaires est liée au mécanisme de brisure de la symétrie électrofaible. Son étude sera l'un des enjeux majeurs de l'expérience Atlas auprès du Large Hadron Collider, au Cern à partir de 2008. Dans la plupart des cas, les recherches seront limitées par notre connaissance des performances du détecteur, telles que la précision avec laquelle l'énergie des particules est reconstruite ou l'efficacité avec laquelle elles sont identifiées. Ce mémoire d'habilitation présente un travail portant sur la reconstruction des électrons dans Atlas avec des données simulées et des données prises durant le test en faisceau combiné qui s'est déroulé en 2004. L'analyse des données d'Atlas nécessite l'utilisation de ressources de calcul et de stockage importantes qui a impliqué le développement d'une grille de calcul mondiale dont un des noeuds est développé au laboratoire. Le manuscrit présente aussi l'effort effectué au LPNHE Paris pour la mise en place d'...

  18. Determination of the fission barrier height in fission of heavy radioactive beams induced by the (d,p)-transfer

    CERN Multimedia

    A theoretical framework is described, allowing to determine the fission barrier height using the observed cross sections of fission induced by the (d,p)-transfer with accuracy, which is not achievable in another type of low-energy fission of neutron-deficient nuclei, the $\\beta$-delayed fission. The primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission of...

  19. Study of the variation with the energy of the fission cross-sections of {sup 233}U, {sup 235}U, {sup 239}Pu for the fast neutrons; Etude de la variation avec l'energie des sections efficaces de fission de {sup 233}U, {sup 235}U, {sup 239}Pu pour les neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Szteinsznaider, D; Naggiar, V; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This measurements have been done while taking the value of the fission cross-sections of {sup 238}U as reference. The neutrons are produced by the reaction {sup 7}Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, cross-section constant between 150 and 2000 keV, for {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, cross-section constant between 700 and 1000 keV, for {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, for neutrons of 850 keV. (authors) [French] Ces mesures ont ete effectuees en prenant la valeur de la section efficace de fission de {sup 238}U comme reference. Les neutrons sont produits par la reaction {sup 7}Li(p,n) au generateur Van de Graaff de Saclay. Le domaine explore s'etend de quelques dizaines de kev a 2000 kev. Nous trouvons: pour {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, section efficace constante entre 150 et 2000 kev. pour {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, section efficace constante entre 700 et 1000 kev. pour {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, pour des neutrons de 850 kev. (auteurs)

  20. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1991-04-01

    Classical dynamical calculations of the heavy ion induced fission process for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus have been performed. As a result prescission lifetimes were obtained and compared with the experimental values. The agreement between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. Somewhat bigger calculated times than the experimental ones in case of the C+Lu reaction at 16 MeV/nucleon may be a signal on the energy range applicability of the one-body dissipation model. (author)

  1. Fission modelling with FIFRELIN

    International Nuclear Information System (INIS)

    Litaize, Olivier; Serot, Olivier; Berge, Leonie

    2015-01-01

    The nuclear fission process gives rise to the formation of fission fragments and emission of particles (n,γ, e - ). The particle emission from fragments can be prompt and delayed. We present here the methods used in the FIFRELIN code, which simulates the prompt component of the de-excitation process. The methods are based on phenomenological models associated with macroscopic and/or microscopic ingredients. Input data can be provided by experiment as well as by theory. The fission fragment de-excitation can be performed within Weisskopf (uncoupled neutron and gamma emission) or a Hauser-Feshbach (coupled neutron/gamma emission) statistical theory. We usually consider five free parameters that cannot be provided by theory or experiments in order to describe the initial distributions required by the code. In a first step this set of parameters is chosen to reproduce a very limited set of target observables. In a second step we can increase the statistics to predict all other fission observables such as prompt neutron, gamma and conversion electron spectra but also their distributions as a function of any kind of parameters such as, for instance, the neutron, gamma and electron number distributions, the average prompt neutron multiplicity as a function of fission fragment mass, charge or kinetic energy, and so on. Several results related to different fissioning systems are presented in this work. The goal in the next decade will be i) to replace some macroscopic ingredients or phenomenological models by microscopic calculations when available and reliable, ii) to be a support for experimentalists in the design of detection systems or in the prediction of necessary beam time or count rates with associated statistics when measuring fragments and emitted particle in coincidence iii) extend the model to be able to run a calculation when no experimental input data are available, iv) account for multiple chance fission and gamma emission before fission, v) account for the

  2. Problems to be Expected in Disposing of Fission Products from a Possible Nuclear Power Programme; Problemes Eventuels d'Elimination des Produits de Fission dans les Futures Centrales Nucleaires; 0412 0415 0420 041e 042f 0414 ; Problemas que Puede Plantear la Evacuacion de Productos de Fision Resultantes de la Ejecucion de un Programa de Produccion de Energia de Origen Nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, G. K. [Cenral Technical Services, Engineering and Development Group, United Kingdom Atomic Energy Authority, Risley (United Kingdom)

    1960-07-01

    typical British nuclear power programme covering the next few decades is examined. The types of reactor to be erected and the types of fuel to be initially used in them are well appreciated but of course the author's conclusions will necessarily become less certain with the passage of time. The chemical processes which must be applied to irradiated fuels will depend on the type of fuel, its burn-up etc., and so will change during the period under review. Consideration is therefore given to the quantities of fission products and heavy isotopes which may be produced, the forms in which they may leave the chemical processing streams and the methods available for handling them safely, either by storage or by disposal. (author) [French] L'auteur examine un programme d'energie nucleaire caracteristique pour la Grande-Bretagne et s'etendant sur les prochaines decennies. Il donne une evaluation fort judicieuse des types de reacteurs dont la construction est envisagee et des genres de combustibles qu'il est question d'y utiliser pendant la periode initiale; toutefois, ses conclusions auront avec le temps de moins en moins de valeur. En effet, le traitement chimique auquel on devra soumettre le combustible irradie depend du genre de ce dernier, de son taux de combustion, etc.; il subira inevitablement des modifications au cours de la periode consideree. L'auteur tient donc compte des quantites de produits fissiles et d'isotopes lourds qu'il sera possible de produire, des formes sous lesquelles ils pourront se presenter a la fin du traitement chimique et des methodes existantes permettant d'assurer leur manipulation sans danger en vue de leur stockage ou de leur elimination. (author) [Spanish] Esta memoria estudia un programa britanico caracteristico de produccion de energia nuclear que cubre las proximas decadas. Se sabe con bastante certeza que tipos de reactores se construiran y el tipo de combustible que se empleara en ellos en un principio, pero estos conocimientos son

  3. 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

  4. Fission fragment angular momentum

    International Nuclear Information System (INIS)

    Frenne, D. De

    1991-01-01

    Most of the energy released in fission is converted into translational kinetic energy of the fragments. The remaining excitation energy will be distributed among neutrons and gammas. An important parameter characterizing the scission configuration is the primary angular momentum of the nascent fragments. Neutron emission is not expected to decrease the spin of the fragments by more than one unit of angular momentum and is as such of less importance in the determination of the initial fragment spins. Gamma emission is a suitable tool in studying initial fragment spins because the emission time, number, energy, and multipolarity of the gammas strongly depend on the value of the primary angular momentum. The main conclusions of experiments on gamma emission were that the initial angular momentum of the fragments is large compared to the ground state spin and oriented perpendicular to the fission axis. Most of the recent information concerning initial fragment spin distributions comes from the measurement of isomeric ratios for isomeric pairs produced in fission. Although in nearly every mass chain isomers are known, only a small number are suitable for initial fission fragment spin studies. Yield and half-life considerations strongly limit the number of candidates. This has the advantage that the behavior of a specific isomeric pair can be investigated for a number of fissioning systems at different excitation energies of the fragments and fissioning nuclei. Because most of the recent information on primary angular momenta comes from measurements of isomeric ratios, the global deexcitation process of the fragments and the calculation of the initial fragment spin distribution from measured isomeric ratios are discussed here. The most important results on primary angular momentum determinations are reviewed and some theoretical approaches are given. 45 refs., 7 figs., 2 tabs

  5. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Witold Nazarewicz

    2003-01-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process

  6. Theoretical Description of the Fission Process

    Energy Technology Data Exchange (ETDEWEB)

    Witold Nazarewicz

    2003-07-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process.

  7. Nuclear fission with a Langevin equation

    International Nuclear Information System (INIS)

    Boilley, D.; Suraud, E.; Abe, Yasuhisa

    1992-01-01

    A microscopically derived Langevin equation is applied to thermally induced nuclear fission. An important memory effect is pointed out and discussed. A strong friction coefficient, estimated from microscopic quantities, tends to decrease the stationary limit of the fission rate and to increase the transient time. The calculations are performed with a collective mass depending on the collective variable and with a constant mass. Fission rates calculated at different temperatures are shown and compared with previous available results. (author) 23 refs.; 7 figs

  8. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

    This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

  9. (d,p)-transfer induced fission of heavy radioactive beams

    CERN Document Server

    Veselsky, Martin

    2012-01-01

    (d,p)-transfer induced fission is proposed as a tool to study low energy fission of exotic heavy nuclei. Primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission offers a possibility for systematic study the low energy fission of heavy exotic nuclei at the ISOLDE.

  10. Guide du calcul en mécanique valider le comportement des systèmes techniques

    CERN Document Server

    Spenlé, Daniel

    2012-01-01

    Ce guide couvre l’essentiel des programmes de la Première STI2D aux sections de BTS pour les compétences liées à la mécanique. Pour prendre en compte les nouveautés pédagogiques, cette nouvelle édition évolue vers des systèmes « grand public » (robotique humanoïde, cycle trainer, power-ball…) ou des systèmes liés au développement durable(éolienne, scooter électrique…). Points forts - Un guide qui tient compte des orientations du nouveau programme de STI2D : - systèmes vibratoires ; - accéléromètres, effet gyroscopique ; - développement de l’énergétique. - Les disciplines Génie mécanique/Génie électrique sont décloisonnées: - introduction de dispositifs pluri-techniques ; - présentation d’interfaces liant la partie opérative à la partie commande ; - introduction de la conversion de l’énergie électrique en énergie mécanique et réciproquement, étude comportementale, point de fonctionnement, définition des grandeurs associées. - Le caractère expérimental, ...

  11. Delayed β ray spectrum of 235U fission fragments

    International Nuclear Information System (INIS)

    Pascholati, P.R.

    1973-01-01

    The time-dependent electron spectra of fission fragments from the thermal-neutron-induced fission of 235 U are calculated. The Gross theory of nuclear beta decay is used to obtain the decay constant and individual electron spectra. The mean energy per fission carried by the electrons and the number of electrons per fission are also calculated. Comparison of these calculated spectra to experimental ones shows good agreements. (Author) [pt

  12. Fission meter

    Science.gov (United States)

    Rowland, Mark S [Alamo, CA; Snyderman, Neal J [Berkeley, CA

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  13. Fission-Fragment Angular, Energy, and Mass Division Correlations for the U{sup 234} (d, Pf) Reaction; Correlation des Angles, des energies et des Masses pour les Fragments de Fission Produits par la Reaction {sup 234}U(d, pf); K 041e 0420 0420 0415 041b 042f 0426 0418 042f 041c 0415 0416 0414 0423 0420 0410 0421 041f 0420 0415 0414 0415 041b 0415 041d 0418 0415 041c 041e 0421 K 041e 041b K 041e 0412 0414 0415 041b 0415 041d 0418 042f 041f 041e 0423 0413 041b 0423 , 041f 041e 042d 041d 0415 0420 0413 0418 0418 0418 041f 041e 041c 0410 0421 0421 0415 0412 0420 0415 0417 0423 041b 042c 0422 0410 0422 0415 0420 0415 0410 K 0426 0418 0419 (d, pf) 0423 0420 0410 041d 0410 -234; Correlaciones Entre la Distribucion Angular, la Energia y la Division Masica de los Fragmentos en la Reaccion {sup 234}U (d, pf)

    Energy Technology Data Exchange (ETDEWEB)

    Vandenbosch, R. [University of Washington, Seattle, WA (United States); Unik, J. P.; Huizenga, J. R. [Argonne National Laboratory, Argonne, IL (United States)

    1965-07-15

    'un analyseur a trois parametres ils ont enregistre simultanement les energies cinetiques des deux fragments de fission et l'energie du proton pour chaque coincidence. L'energie d'excitation pour laquelle la fission se produit est definie par l'energie cinetique du proton libere. Les variations de l'anisotropie angulaire Selon l'energie d'excitation ont une structure beaucoup plus nette que celle que Lamphere a obtenue pour le meme moyau, en bombardant {sup 234}U par des neutrons rapides. Les auteurs ont observe au moins huit voies de fission au point d'etranglement dans la region des energies comprises entre le seuil et 2 MeV au-dessus du seuil. Ils ont calcule (calculs de type Nilsson) des energies de particules isolees pour des noyaux deformes dans le cas de deformations plus grandes qui dorment une description pius precise de la configuration au point d'etranglement. Les etats d'une particule isolee identifies par Lamphere sont en accord avec ceux dont on a trouve par le calcul qu'ils sont voisins de la surface de Fermi pour des deformations raisonnables au point d'etranglement. Ces experiences avaient primitivement pour but de rechercher s'il n'y avait pas une correlation entre l'asymetrie de masse et l'anisotropie angulaire. Les rendements Selon la masse obtenus a partir des energies de fragments en correlation ne font apparaitre aucune variation de l'anisotropie Selon le rapport des masses, contrairement a ce que se passe dans les experiences ou l'on ne fixe pas l'energie d'excitation a laquelle la fission a lieu (experiences ou l'on observe effectivement une variation de l'anisotropie Selon le rapport des masses). On ne saurait donc affirmer, d'apres les mesures de l'anisotropie, que les proprietes du point d'etranglement influencent la repartition finale des masses. La moyenne de l'energie cinetique totale liberee au cours de la fission varie dans une proportion inferieure a 0,5% pour les differentes voies a l'etranglement observees. Les auteurs ont aussi etudie la

  14. Contribution to the study of the behaviour, in the urban environment, during the runoff of rainwater, of the fission products emitted during a nuclear accident; Contribution a l`etude du devenir, en milieu urbain, pendant le ruissellement des eaux pluviales, des produits de fission emis en cas d`accident nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Pioch, M

    1993-05-24

    In the context of research into the environmental consequences of a serious accident occurring on a pressurized water reactor, this paper concerns the experimental study of behaviour of five fission products (caesium, strontium, iodine, ruthenium and tellurium) in the urban environment under the action of rainwater. Stable or radioactive multiple-element aerosols were produced. Their physicochemical characteristics and their solubility in rainwater were studied. Caesium and rubidium forms solutions totally and quickly, while strontium is partially soluble (approximately 50 %) and iodine is only slightly soluble. The behaviour of fission products on five urban surfaces was then studied. Batch experiments showed that the retention of dissolved forms of radioelements varied according to the material. The reactions involved are ion exchange reactions. The presence of certain ions in water (in particular NH{sub 4}{sup +}) increase the desorption of radioelements. Using a laboratory rainfall simulator, the re-entrainment of fission products by rainwater was examined. Two modes of deposition and two intensities of rainfall were simulated. The desorption of radioelements is greater after wet deposition and remobilization is reduced by an increase in intensity of rainfall. An addition of NH{sub 4}{sup +} in water is especially effective in the case of wet depositions. Suggestions are made in order to improve experimental protocols and continue the research. (author). 75 refs., 51 figs., 69 tabs., 14 appends.

  15. Comparison of the Economic Aspects of the Treatment and Storage of Fission Products from Installations Processing Irradiated Natural Uranium; Aspects Economiques Compares du Traitement et du Stockage des Produits de Fission Issus des Usines de Traitement de l'Uranium Naturel Irradie; Sravnitel'nye ehkonomicheskie aspekty obrabotki i khraneniya produktov deleniya, poluchaemykh na zavodakh po pererabotke obluchennogo prirodnogo urana; Aspectos Economicos Comparados del Tratamiento y del Almacenamiento de los Productos de Fision que Salen de las Plantas de Tratamiento de Uranio Natural Irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Thiriet, L.; Lesur, P.; Giraud, R. [Commissariat a l' Energie Atomique, Paris (France); Wanlin, J. [Societe Saint-Gobain Techniques Nouvelles, Paris (France)

    1966-02-15

    The paper summarizes the main policies adopted for the management of fission product effluents and briefly describes the essential technical features of the stainless steel storage installations (in acid medium) used in France, giving a breakdown of the costs involved. The dependence of these costs on the activity of the solutions stored and on their heat release is shown. The second part of the paper discusses the cost of storage in terms of feasible tank size. The optimum size for such tanks and the rate at which they are placed in service are determined in relation to the characteristics of the fission products to be stored and to the respective capacities of the installations for processing irradiated natural uranium. It is shown that although storage costs depend on the assumptions made regarding the useful life of installations, rates of interest and operating costs, optimum policy decisions (as to size and rate of entry into service) are independent of these assumptions, being determined solely by the variation in the cost of tanks, according to size. These are the factors which enter into the optimized cost of storing these effluents indefinitely. In the third part, this method of indefinite storage is compared with other possibilities of fission product management, e.g. vitrification and separation of Cs and Sr. The paper discusses the economics of the various methods and summarizes some possible long-term solutions. (author) [French] Apres avoir rappele les principales politiques de gestion le plus souvent envisagees pour les produits de fission, le memoire decrit sommairement les caracteristiques techniques essentielles des installations de stockage en acier inoxydable (en milieu acide) utilisees en France et la structure des couts de ces installations. U montre la sensibilite de ces couts a l'activite des solutions stockees et a leur degagement de chaleur. Dans une deuxieme partie, les auteurs examinent leur cout de stockage, selon la taille des

  16. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's commission on the accident at Three Mile Island

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.

    1980-03-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  17. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Wilson, W.B.

    1979-10-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis.

  18. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.

    1979-10-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  19. Heat treatments of irradiated uranium oxide in a pressurised water reactor (P.W.R.): swelling and fission gas release; Traitements thermiques de l`oxyde d`uranium irradie dans un reacteur a eau pressurisee (R.E.P.): gonflement et relachement des gaz de fission

    Energy Technology Data Exchange (ETDEWEB)

    Zacharie, I

    1997-03-27

    In order to keep pressurised water reactors at a top level of safety, it is necessary to understand the chemical and mechanical interaction between the cladding and the fuel pellet due to a temperature increase during a rapid change in reactor. In this process, the swelling of uranium oxide plays an important role. It comes from a bubble precipitation of fission gases which are released when they are in contact with the outside. Therefore, the aim of this thesis consists in acquiring a better understanding of the mechanisms which come into play. Uranium oxide samples, from a two cycles irradiated fuel, first have been thermal treated between 1000 deg C and 1700 deg C for 5 minutes to ten hours. The gas release amount related to time has been measured for each treatment. The comparison of the experimental results with a numerical model has proved satisfactory: it seems that the gases release, after the formation of intergranular tunnels, is controlled by the diffusion phenomena. Afterwards, the swelling was measured on the samples. The microscopic examination shows that the bubbles are located in the grain boundaries and have a lenticular shape. The swelling can be explained by the bubbles coalescence and a model was developed based on this observation. An equation allows to calculate the intergranular swelling in function of time and temperature. The study gives the opportunity to predict the fission gases behaviour during a fuel temperature increase. (author) 56 refs.

  20. 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

  1. Device for measuring fission product density

    International Nuclear Information System (INIS)

    Kaneda, Mitsunori.

    1980-01-01

    Purpose: To determine the fission product density of xenon or the like and enable measurement of real time of fission product density in a reactor by calculating the disintegration and annihilation of the fission product on the basis of neutron detected output. Constitution: The neutron flux in a reactor is detected by a detector, and applied to first and second density calculators. Second fission product density signal of xenon or the like outputted from first device is again inputted to the device to form an annihilation signal due to disintegration to determine the present density of the second fission product of xenon or the like corresponding to the decrease of the neutron due to the poison of xeron or the like. Similarly, second device determines the first fission product density of iodine or the like. (Sekiya, K.)

  2. Ideological Fission

    DEFF Research Database (Denmark)

    Christiansen, Steen Ledet

    ; it is a materialisation of an ideological fission which attempts to excise certain ideological constructions, yet paradoxically casting them in a form that is recognizable and familiar. The monstrous metonomy which is used shows us glimpses of a horrid being, intended to vilify the attack on New York City. However......, it is a being which is reminiscent of earlier monsters - from Godzilla to The Blob. It is evident that the Cloverfield monster is a paradoxical construction which attempts to articulate fear and loathing about terrorism, but ends up trapped in an ideological dead-end maze, unable to do anything other than...

  3. Cumulative fission yield of Ce-148 produced by thermal-neutron fission of U-235

    International Nuclear Information System (INIS)

    Hasan, A.A.

    1984-12-01

    Cumulative fission yield of 148 cesium isotopes and some other fission products produced by thermal-neutron fission of 235 uranium is determined by Germanium/Lithium spectroscopic methods. The measuremets were done at Tsing-Hua open pool reactor using 3 to 4 mg of 93.15% enriched 235 uranium samples. Gamma rays are assigned to the responsible fission products by matching gamma rays energies and half lives. Fission rate is calculated by fission track method. Cumulative fission yields of 148 cesium, 90 krypton, 130 iodine, 144 lanthanum, 89 krypton, 136 xenon, 137 xenon and 140 cesium are calculated. This values are compared with previously predicted values and showed good agreement. 21 Ref

  4. Poison 1 - a programme for calculation of reactivity transients due to fission product poisoning and its application in continuous determination of xenon and samarium poisoning in reactor KS-150

    International Nuclear Information System (INIS)

    Rana, S.B.

    1973-12-01

    The report contains a user's description of the 3-dimensional programme POISON 1 for calculating reactivity transients due to fission-product poisoning during changes of reactor power. The chapter dealing with Xe poisoning contains a description of Xe tables, the method of operational determination of Xe poisoning, use of Xe transients for calibrating control rods and means of shutting down the reactor without being overriden by Xe poisoning. Sm poisoning is determined continuously on the basis of the power diagram of reactor operation. In conclusion a possibility of using the programme in a process computer in combination with self-powered detectors as local power sensors is indicated. (author)

  5. In Vivo and Air Dosimetry of Fission-Spectrum Neutrons; Dosimetrie In Vivo et dans l'Air du Spectre des Neutrons de Fission; Dozimetriya v vozdukhe i dozimetriya In Vivo nejtronov spektra deleniya; Dosimetria In Vivo y en el Aire de Neutrones de un Espectro de Fision

    Energy Technology Data Exchange (ETDEWEB)

    Mobley, T. S.; Engel, R. E.; Godden, W. R. [Kirtland Airforce Base, New Mexico (United States); Penikas, V. T. [AFIT, Wright-Patterson Air Force Base, Ohio, with Duty Station at University of Rochester, School of Medicine and Dentistry, Rochester, NY (United States)

    1964-03-15

    internal package varied with burst temperature rise, distance from the reactor, orientation of the foils with respect to the reactor, and the thickness and composition of the attenuating tissues. (author) [French] En vue d'evaluer la dose profonde du fait de l'exposition au spectre des neutrons de fission, les auteurs ont mis au point une methode de dosimetrie in vivo sur le mouton. On a utilise pour ces etudes un reacteur a flux puise semblable a l'appareil GODIVA II du Laboratoire scientifique de Los Alamos. En regime puise, le reacteur donne une impulsion d'une largeur de 50 {mu}s a la moitie du maximum d'amplitude pour une elevation de temperature du combustible d'environ 100 Degree-Sign C. Une insertion sous-cutanee contient une feuille d'or recouverte de cadmium, une feuille d'or nue et trois dosimetres en verre thermoluminescent places dans un etui de lithium et de plomb double de verre. On a place les insertions dans le tissu sous-cutane des deux cotes du corps pour detecter la dose totale de neutrons thermiques-gamma, les neutrons etant delivres sur les flancs. Des insertions abdominales sont preparees a peu pres de la meme facon que les insertions sous-cutanees; elles sont constituees d'une boule de bore contenant des feuilles a seuil de fission (Pu, Np, Ue) recouvertes de cadmium, d'un comprime de soufre et de trois dosimetres identiques a ceux des insertions sous-cutanees. On les a placees dans l'abdomen des sujets 18 h avant l'exposition, en les fixant au pilier anterieur du rumen. Les auteurs decrivent un procede de rumenotomie. On a determine la dose de neutrons au moyen de la feuille a seuil. On a utilise les feuilles d'or et les feuilles d'or recouvertes de cadmium pour determiner le flux integre de neutrons thermiques, le plutonium-239, le neptumum-237, l'uranium-238 et le soufre-32 pour determiner le flux integre de neutrons rapides, et les batonnets de verre radiophotoluminescents pour determiner la dose gamma. Les moutons ont subi des expositions

  6. Méthodes de calcul pour la conception des systèmes de protection cathodique des structures longilignes Computing Methods for Designing Cathodic Protection Systemes for Elongate Stuctures

    Directory of Open Access Journals (Sweden)

    Roche M.

    2006-11-01

    Full Text Available Les différentes structures longilignes qu'utilise l'industrie des hydrocarbures sont, dans la plupart des cas, soumises à un système de protection cathodique par anodes sacrificielles ou par courant imposé. La conception de ces systèmes doit être basée sur l'étude de la variation du potentiel et de l'intensité le long de la structure causée par la chute ohmique. La méthode classique de calcul résoud couramment le cas des structures longilignes à diamètre constant traversant un terrain dont la résistivité est considérée comme constante sur toute la longueur. Dans le cas où la constitution de la structure varie, comme celui des casings de puits de forage, ou quand celle-ci traverse plusieurs types de terrain, le problème se complique. Nous proposons une méthode générale permettant de traiter rapidement tout problème de ce type, le nombre de tronçons n'étant pas limité. Cette méthode fait appel à des notions de facteur de réflexion et de résistance équivalente déjà exposées dans la littérature mais dont l'usage ne semble pas s'être répandu. The different elongate structures used by the hydrocarbon industry are, in most cases, subjected ta a cathodic protection system consisting of sacrificial anodes or an impressed current. Desings of such systems must be boséd on an analysis of variations in the potential and intensity along the structure as the result of the ohm drop. The conventional computing method commonly solves cases of elongote structures with a constant diameter, running through ground whose resistivity is considéred to be constant over the entire length. When the~nake-up of the structure varies, as is the case for borehole casings, or when it goes through several types of formations, the problem gets more complicated. We propose a general method for quickly dealing with any problem of this type, with no limit ta the number of lengths involved. This method makes use of reflection factor and

  7. Mica fission detectors

    International Nuclear Information System (INIS)

    Wong, C.; Anderson, J.D.; Hansen, L.; Lehn, A.V.; Williamson, M.A.

    1977-01-01

    The present development status of the mica fission detectors is summarized. It is concluded that the techniques have been refined and developed to a state such that the mica fission counters are a reliable and reproducible detector for fission events

  8. Use of proportional counter in X-Ray spectrometry between 5 and 100 keV. Application to the detection of fission products and to the determination of absolute X-Ray disintegration rates; Utilisation d'un compteur proportionnel en spectrometrie de rayons X entre 5 et 100 keV. Application a la detection des produits de fission et a la determination des taux d'emission X absolus

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    The measurement of electromagnetic radiations is difficult in the energy range 20-100 keV. We made suitable for this purpose a regular proportional counter, modifying both the nature and pressure of the gaseous mixture filling the detector volume. We selected the CPEN-SAIP counter, which is able to withstand such modifications. In the energy range considered, the counter is to be standardized with radioactive sources. Such standards were selected according to their disintegration schemes. We thus defined the conditions of use (resolution, yield) of the CPN counter, filled with an argon-methane mixture under a pressure of about 3 bars, in the energy range 5-100 keV. With such an equipment, we were able to measure the absolute disintegration rate for the X-rays of 133 Ba and 75 Se, then to perform the study of a mixed fission products sample. In the same way, we used xenon-based gaseous mixtures, in order to improve the detector yield; in the later case, we carefully examined the limitations introduced by the presence of many parasite rays emitted by the gas. We thus displayed in addition to the leakage peak, the fluorescence ray of the gas, whose origin is difficult to explain. (author) [French] La mesure des rayonnements electromagnetiques est difficile dans une bande d'energie comprise entre 20 et 100 keV environ. Nous avons rendu utilisable, a cet effet, un compteur proportionnel existant en modifiant la nature et la pression du melange gazeux constituant le volume detecteur. De par sa construction, le compteur CPXN, fabrique par les Etablissements S.A.I.P., rend ces experiences possibles. C'est pourquoi nous l'avons choisi. Dans le domaine d'energie consideree l'etalonnage du compteur necessite l'emploi de sources radioactives. Pour les preparer, nous sommes amenes a choisir les radio-nucleides utilisables par l'etude de leur schema de desintegration. Ainsi, nous avons defini les conditions d'utilisation (resolution, efficacite) du compteur CPNX rempli d

  9. Improved perturbative calculations in field theory; Calculation of the mass spectrum and constraints on the supersymmetric standard model; Calculs perturbatifs variationnellement ameliores en theorie des champs; Calcul du spectre et contraintes sur le modele supersymetrique standard

    Energy Technology Data Exchange (ETDEWEB)

    Kneur, J.L

    2006-06-15

    This document is divided into 2 parts. The first part describes a particular re-summation technique of perturbative series that can give a non-perturbative results in some cases. We detail some applications in field theory and in condensed matter like the calculation of the effective temperature of Bose-Einstein condensates. The second part deals with the minimal supersymmetric standard model. We present an accurate calculation of the mass spectrum of supersymmetric particles, a calculation of the relic density of supersymmetric black matter, and the constraints that we can infer from models.

  10. Proportioning of {sup 79}Se and {sup 126}Sn long life radionuclides in the fission products solutions coming from spent fuels processing; Dosage des radionucleides a vie longue {sup 79}Se et {sup 126}Sn dans les solutions de produits de fission issues du traitement des combustibles nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Comte, J

    2001-11-01

    The determination of radionuclides present in waste resulting from the nuclear fuel reprocessing is a request from the regulatory authorities to ensure an optimal management of the storage sites. Long-lived radionuclides (T{sub 1/2} > 30 years) are particularly concerned owing to the fact that their impact must be considered for the long term. Safety studies have established a list of long-lived radionuclides (LLRN) whose quantification is essential for the management of the disposal site. Among these, several are pure {beta} emitters, present at low concentration levels in complex matrices. Their determination, by radiochemical method or mass spectrometry, involves selective chemical separations from the others {beta}/{gamma} emitters and from the measurement interfering elements. The work undertaken in this thesis relates to the development of analytical methods for the determination of two long-lived radionuclides: selenium 79 and tin 126, in acid solutions of fission products present in nuclear fuel reprocessing plant. For selenium 79, a {beta} emitter with a half live estimated to be 10{sup 6} years, the bibliography describes different chemical separation methods including precipitation, liquid-liquid extraction and chromatography on ionic resins. After optimisation on a synthetic solution, two of these techniques, precipitation by potassium iodine and separation with ion exchange resins were applied to a genuine solution of fission products at Cogema La Hague. The results showed that only the ion exchange method allows us to obtain a solution sufficiently decontaminated (FD{beta}{gamma} = 250) with a significant selenium recovery yield (85%). This separation allows the measurement of the {sup 79}Se by electrothermal vaporization coupled with inductively coupled plasma mass spectrometry (ETV-ICP/MS), after transfer of the samples to CEA/Cadarache. The concentration of {sup 79}Se measured is 0,42 mg/L in the solution of fission products with an isotopic ratio

  11. A stochastic approach to fission

    International Nuclear Information System (INIS)

    Boilley, D.; Suraud, E.; Abe, Yasuhisa

    1992-01-01

    A microscopically derived Langevin equation is applied to thermally induced nuclear fission. An important memory effect is pointed out. A strong friction coefficient, calculated from microscopic quantities, tends to decrease the stationary limit of the fission rate and to increase the transient time. Fission was described as a diffusion over a barrier of a collective variable, and a Langevin Equation (LE) was used to study the phenomenon. A study of the stationary flow over the saddle point with a Fokker-Planck Equation (FPE), equivalent to the LE was used to give formula for the stationary fission rate (or reaction rate for the chemistry applications). More recently, a complete study of the fission process was performed numerically with both FPE and LE. A long transient time, that could allow more pre-scission neutrons to evaporate, was pointed out. The derivation of this new LE is recalled, followed by the description of the memory dependence and by the effect of a large friction coefficient on the fission rate. (author) 6 refs., 3 figs

  12. Neutronics of Laser Fission-Fusion Systems

    International Nuclear Information System (INIS)

    Velarde, G.

    1976-01-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-∞). (Author) 14 refs

  13. Neutronics of Laser Fission-Fusion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G

    1976-07-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-{infinity}). (Author) 14 refs.

  14. Calculation of proper values {lambda}{sub mn}(c) of the spheroidal equation; Calcul des valeurs propres {lambda}{sub mn}(c) de l'equation spheroidale

    Energy Technology Data Exchange (ETDEWEB)

    Dufour, J M [CEA Limeil Valenton, 94 - Villeneuve-Saint-Georges (France)

    1969-05-01

    The aim of this report is to find, with a fair accuracy, a proper value {lambda}{sub mn}(c) for the spheroidal differential equation: d/dz[(1-z{sup 2})du/dz]+[ {lambda} - c{sup 2}z{sup 2} - m{sup 2}/(1-z{sup 2})]u = 0 obtained by the separation of the three variables of the wave equation: {delta}{sup 2}u + k{sup 2}u = 0 with rotational elongated or flattened ellipsoidal coordinates. The program drawn up calculates {lambda}{sub mn}(c) for any values of (mnc) chosen in the zones 0 {<=} | m | {<=} 10, a whole number; |m| {<=} n {<=} 20, n a whole number; 0 {<=} |c | {<=} 30; previous work has covered a smaller field of values. The function to be solved by the approximation method of the Newton-Raphson type, and the initial value, are chosen so as to converge towards the required solution. (author) [French] L'objet de ce rapport est de rechercher avec une tres bonne approximation, une valeur propre {lambda}{sub mn}(c) de l'equation differentielle spheroidale: d/dz[(1-z{sup 2})du/dz]+[ {lambda} - c{sup 2}z{sup 2} - m{sup 2}/1-z{sup 2}]u = 0 obtenue par separation des 3 variables de l'equation des ondes: {delta}{sup 2}u + k{sup 2}u = 0 en coordonnees des ellipsoides de revolution allonges ou aplatis. Le programme etabli calcule {lambda}{sub mn}(c) quel que soit le jeu (mnc) choisi dans le domaine 0 {<=} | m | {<=} 10 entier; |m| {<=} n {<=} 20, n entier; 0 {<=} |c | {<=} 30; alors que les etudes precedentes portaient sur un domaine plus restreint. La fonction a resoudre par la methode d'approximation du type NEWTON-RAPHSON et la valeur initiale, sont choisies de facon a converger vers la solution desiree. (auteur)

  15. Study by photo-ionization of some simple poly-atomic molecules and calculation of the Franck-Condon factors; Etude par photo-ionization de quelques molecules poly-atomiques simples et calcul des facteurs de Franck-Condon

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-02-01

    The photo-ionization yield curves for C{sub 2}H{sub 2}, C{sub 2}D{sub 2}, C{sub 2}H{sub 4} and C{sub 2}H{sub 2}D{sub 2} have been determined using a mass spectrometer coupled with an U.V. monochromator. Besides exhibiting a stair case structure near threshold due to the excitation of vibrational levels in the ion ground state, all the curves have broad maxima corresponding to auto-ionization phenomena. The ionization potentials of these molecules have been measured, together with the appearance potentials of the main ion-fragments. The excitation probabilities for the vibrational levels during ionization, or Franck-Condon factors, have been calculated for C{sub 2}H{sub 2} and C{sub 2}D{sub 2} using the method developed by Sharp and Rosenstock. Good agreement is generally obtained between the calculated values and those obtained experimentally from the photo-ionization yield curves. The preceding calculation method is then extended to the case where the electronic transition occurs with changes in the geometrical structure of the molecule (in particular, changes of symmetry). The Franck-Condon factors have been determined for NH{sub 3} (symmetry changes) and for H{sub 2}O (changes in the equilibrium angle). Calculations show that there is generally considerable excitation of the combination bands. (author) [French] Les courbes de rendement de photoionisation pour C{sub 2}H{sub 2}, C{sub 2}D{sub 2}, C{sub 2}H{sub 4} et C{sub 2}H{sub 2}D{sub 2} determinees a l'aide d'un spectrometre de masse auquel etait couple un monochromateur U.V. En plus d'une structure en escalier au voisinage du seuil, due a l'excitation de niveaux vibrationnels dans l'ion a l'etat fondamental, toutes les courbes presentent des maxima tres aplatis correspondant a des phenomenes d'auto-ionisation. Les potentiels d'ionisation de ces molecules ont ete mesures ainsi que les potentiels d'apparition des principaux ions fragments. Les probabilites d'excitation de niveaux de vibration au cours de l

  16. Calculation and Evaluation of Fission Yields and Capture Cross Sections Leading to the Production of Therapeutic Radionuclide by Means of Nuclear Reactors

    International Nuclear Information System (INIS)

    Sublet, J.C.

    2009-01-01

    Much progress has been made in nuclear medicine that involves the use of radionuclides for both diagnosis and therapy. Because of this qualitative and quantitative growth, the adoption of a set of established radionuclides for various applications, the methods of nuclide production need to be addressed and consideration given to other, emerging radionuclides that are judged to be developing in importance. The methods involved are characterized by the transmutation of isotopes by neutron-induced reactions and decays. Therefore, newly evaluated cross sections, fission yields and decay characteristics of relevance to the reactor production of those therapeutic radionuclides have been reviewed. Considerations of the decay schemes of all the nuclides involved are also included. (author)

  17. Monte carlo sampling of fission multiplicity.

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J. S. (John S.)

    2004-01-01

    Two new methods have been developed for fission multiplicity modeling in Monte Carlo calculations. The traditional method of sampling neutron multiplicity from fission is to sample the number of neutrons above or below the average. For example, if there are 2.7 neutrons per fission, three would be chosen 70% of the time and two would be chosen 30% of the time. For many applications, particularly {sup 3}He coincidence counting, a better estimate of the true number of neutrons per fission is required. Generally, this number is estimated by sampling a Gaussian distribution about the average. However, because the tail of the Gaussian distribution is negative and negative neutrons cannot be produced, a slight positive bias can be found in the average value. For criticality calculations, the result of rejecting the negative neutrons is an increase in k{sub eff} of 0.1% in some cases. For spontaneous fission, where the average number of neutrons emitted from fission is low, the error also can be unacceptably large. If the Gaussian width approaches the average number of fissions, 10% too many fission neutrons are produced by not treating the negative Gaussian tail adequately. The first method to treat the Gaussian tail is to determine a correction offset, which then is subtracted from all sampled values of the number of neutrons produced. This offset depends on the average value for any given fission at any energy and must be computed efficiently at each fission from the non-integrable error function. The second method is to determine a corrected zero point so that all neutrons sampled between zero and the corrected zero point are killed to compensate for the negative Gaussian tail bias. Again, the zero point must be computed efficiently at each fission. Both methods give excellent results with a negligible computing time penalty. It is now possible to include the full effects of fission multiplicity without the negative Gaussian tail bias.

  18. Nuclear fission and reactions

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The nuclear fission research programs are designed to elucidate basic features of the fission process. Specifically, (1) factors determining how nucleons of a fissioning nucleus are distributed between two fission fragments, (2) factors determining kinetic energy and excitation energies of fragments, and (3) factors controlling fission lifetimes. To these ends, fission studies are reported for several heavy elements and include investigations of spontaneous and neutron-induced fission, heavy ion reactions, and high energy proton reactions. The status of theoretical research is also discussed. (U.S.)

  19. The Cyrano program. 1 - description and operation of an irradiation device 'Cyrano'. 2 - results of the experiments Cyrano 1 and 2 (study of the EL 4 first-bach pencil); measurement of the thermal conductivity integral for UO{sub 2} sintered up to 2300 deg C; evolution of fission gases at constant power; Programme Cyrano. 1 - description et exploitation d'un dispositif d'irradiation ''cyrano''. 2 - resultats des experiences cyrano 1 et 2 (etude du crayon EL4 1. jeu). Mesure de l'integrale de conductibilite thermique d'UO{sub 2} fritte jusqu'a 2300 deg C. Evolution des gaz de fission a puissance constante

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

    Two rods of the type 'EL 4 first score' have been irradiated in 'Cyrano' capsules which are suited for continuous measurement of the nuclear power evolved and equipped with thermocouples. The variations of the integral of conductivity of sintered 96 per cent theoretical dense UO{sub 2} has been established until 2300 deg. C; these variations are coherent with those previously measured out-of-pile. The released fission gases have been recovered at different times during the irradiation: the kinetics of release of stable gases is such that, in the experimental conditions (T{sub surface} = 610 deg. C, {integral}{sub T{sub s}}{sup T{sub c}} kdT = 34.7 W cm{sup -1}) the fraction of released gases is still widely increasing after 12 days of irradiation. Numerous observations have been made on concentrations of rare gases locally present in the irradiated fuel. (authors) [French] L'irradiation de deux crayons combustibles type EL 4, 1er jeu, a ete menee a bien dans des capsules 'Cyrano' equipees de dispositifs de mesure continue de la puissance nucleaire et de plusieurs reperes de temperatures; la variation de l'integrale de conductibilite thermique de l'oxyde d'uranium fritte (96 pour cent d. th.) a ete tracee jusqu'a 2300 deg. C; la courbe representative de ces variations est coherente avec celle obtenue precedemment hors pile. Les gaz de fission apparus hors du combustible ont ete extraits du crayon a plusieurs reprises pendant l'experience: la cinetique d'accumulation des gaz stables est telle que dans les conditions etudiees (T{sub surface} = 610 deg. C, {integral}{sub T{sub s}}{sup T{sub c}} kdT = 34.7 W cm{sup -1}) la fraction des gaz degages est encore largement croissante apres 12 jours d'irradiation. De nombreuses observations ont ete recueillies sur la nature et la concentration des gaz rares presents en differents points du combustible irradie. (auteurs)

  20. Baby fission chambers; Etude de chambres a fission miniatures

    Energy Technology Data Exchange (ETDEWEB)

    Guery, U; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

  1. Baby fission chambers; Etude de chambres a fission miniatures

    Energy Technology Data Exchange (ETDEWEB)

    Guery, U.; Tachon, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

  2. Measurements of fission yields

    International Nuclear Information System (INIS)

    Denschlag, H.O.

    2000-01-01

    After some historical introductory remarks on the discovery of nuclear fission and early fission yield determinations, the present status of knowledge on fission yields is briefly reviewed. Practical and fundamental reasons motivating the pursuit of fission yield measurements in the coming century are pointed out. Recent results and novel techniques are described that promise to provide new interesting insights into the fission process during the next century. (author)

  3. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

    Research progress is reported on nuclear chemistry; topics considered include: recoil range and kinetic energy distribution in the thermal neutron ftssion of /sup 245/Cm; mass distribution and recoil range measurements in the reactor neutron-induced fission of /sup 232/U; fission yields in the thermal neutron fission of /sup 241/PU highly asymmetric binary fission of uranium induced by reactor neutrons; and nuclear charge distribution in low energy fission. ( DHM)

  4. Contribution to research on the metabolism of fission product. Studies on the physico-chemical state and the metabolic fate of radio-cerium solution; Contribution a l'etude du metabolisme des produits de fission. Recherches sur l'etat physico-chimique et le devenir metabolique des solutions de radiocerium

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    This paper describes a study of the physico-chemical state of radio-cerium in dilute solutions on the tracer scale, as a function of the pH of the solution. The way in which this radioelement is transported in the blood is studied in vitro and in vivo, with reference to the ionic or colloidal state of the radio-cerium used. The distribution of cerium amongst the various components of the blood is studied by a new method of blood fractionation and by paper electrophoresis. Evidence of a cerium globulin connection is shown in the case of ionic cerium. A study of the initial distribution of radio-cerium in rats, after intravenous administration of ionic or colloidal solutions, shows considerable differences according to the physico-chemical state of the cerium injected. (author) [French] Ce travail presente une etude de l'etat physico-chimique du radiocerium en solutions diluees a l'echelle des indicateurs, en fonction du pH de la solution. En fonction de l'etat ionique ou colloidal du radiocerium utilise, le mode de transport de ce radioelement dans le sang est etudie in vitro et in vivo. La distribution du cerium entre les differents constituants du sang est etudiee par une methode nouvelle de fractionnement du sang et par electrophorese sur papier. L'existence d'une liaison cerium-globuline est mise en evidence dans le cas du cerium ionique. L'etude de la distribution initiale du radiocerium, chez le rat apres administration par voie veineuse de solutions ionique ou colloidale, montre des differences importantes en fonction de l'etat physico-chimique du cerium injecte. (auteur)

  5. The fast fission effect in a cylindrical fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Carlvik, I; Pershagen, B

    1959-06-15

    A new formula for the fast fission factor is derived, which takes proper account to fast capture. The fission neutron spectrum is divided into two groups with constant fission cross section in one group and zero fission cross section in the other. The average total, elastic, inelastic and capture cross sections in the two groups are calculated. Different assumptions regarding anisotropic and inelastic scattering are investigated. The effects of backscattering from the moderator and fast fission in neighbouring fuel elements are pointed out. Formulas for the fast fission ratio and for the fast conversion ratio are derived. The calculated fast fission ratios are compared with experimental values. Curves are given for the fast fission factor in uranium metal and uranium oxide.

  6. Muon induced fission and fission track dating of minerals

    International Nuclear Information System (INIS)

    Marques, A.

    1988-01-01

    The effects of muon induced fission on geological dating of samples by the fission track method are evaluated for the case of muscovite minerals. It is found a small but significant effect, greater for the longer ages. Since calculations are developped under the hypothesis of constant atmosphere and primary cosmic ray flux it is suggested that any discrepancy found in ages of very old material that cannot be accounted for by well known environmental influences, be taken as an indication of variation on either the atmospheric stopping power or the intensity of cosmic radiation along the ages. (author) [pt

  7. Solidification of residual fission-product solutions; laboratory studies; Solidification de solutions residuaires concentrees de produits de fission: etudes de laboratoire

    Energy Technology Data Exchange (ETDEWEB)

    Boniaud, R; Cohen, P; Sombret, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1961-07-01

    This paper describes the results obtained, at laboratory scale, during the study of the incorporation of fission products into glasses and synthetic micas. The rate of leaching of fission products from the glass and their volatility during firing were measured. A hot cell was built to complete these results. (author) [French] Ce rapport resume le resultat des etudes faites en laboratoire (activite de l'ordre du millicurie) sur l'incorporation des produits de fission dans des verres et micas synthetiques. On a mesure le taux de lixiviation des produits de fission et leur volatilisation au cours de la cuisson. Une cellule chaude a ete installee pour completer ces resultats au moyen d'essais realises avec une activite superieure. (auteur)

  8. Prompt fission neutron spectra of n + 235U above the (n, nf) fission threshold

    International Nuclear Information System (INIS)

    Shu Nengchuan; Chen Yongjing; Liu Tingjin; Jia Min

    2015-01-01

    Calculations of prompt fission neutron spectra (PFNS) from the 235 U(n, f) reaction were performed with a semi-empirical method for En = 7.0 and 14.7 MeV neutron energies. The total PFNS were obtained as a superposition of (n, xnf) pre-fission neutron spectra and post-fission spectra of neutrons which were evaporated from fission fragments, and these two kinds of spectra were taken as an expression of the evaporation spectrum. The contributions of (n, xnf) fission neutron spectra on the calculated PFNS were discussed. The results show that emission of one or two neutrons in the (n, nf) or (n, 2nf) reactions influences the PFNS shape, and the neutron spectra of the (n, xnf) fission-channel are soft compared with the neutron spectra of the (n, f) fission channel. In addition, analysis of the multiple-chance fission component showed that second-chance fission dominates the PFNS with an incident neutron energy of 14.7 MeV whereas first-chance fission dominates the 7 MeV case. (authors)

  9. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Wahl, A.C.

    2002-01-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z F = 90 thru 98, mass number A F = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

  10. Theory of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1989-01-01

    Following a summary of the observables in neutron emission in fission, a brief history is given of theoretical representations of the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity /bar /nu///sub p/. This is followed by descriptions, together with examples, of modern approaches to the calculation of these quantities including recent advancements. Emphasis will be placed upon the predictability and accuracy of the modern approaches. In particular, the dependence of N(E) and /bar /nu///sub p/ on the fissioning nucleus and its excitation energy will be discussed, as will the effects of and competition between first-, second- and third-chance fission in circumstances of high excitation energy. Finally, properties of neutron-rich (fission-fragment) nuclei are discussed that must be better known to calculate N(E) and /bar /nu///sub p/ with higher accuracy than is currently possible. 17 refs., 11 figs

  11. Systematics of Fission-Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    A.C. Wahl

    2002-05-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

  12. Fission - track age of the Marjalahti Pallasite

    International Nuclear Information System (INIS)

    Bondar, Yu.V.; Perelygin, V.P.

    2006-01-01

    Full text: Investigation of fossil charged-particle tracks in various mineral phases of extraterrestrial samples is a powerful method for research the early stages of the solar system. Over geological time, meteorites crystals have accumulated a record of tracks produced by heavily charged energetic particles from both internal (spontaneous fission of 238U and some other extinct isotopes) and external sources (galactic cosmic rays with Z>20). The fortunate fact that meteorite grains can accumulate latent and very long-lived tracks since soon after the end of nucleosynthesis in the solar nebula enables one to decode their radiation history and to detect any thermal events in the meteorite cosmic history by revealing these tracks through suitable etching procedures. Only a few minerals in meteorites (mainly phosphates) contain small amount of uranium; the fact that 238 U undergoes fission with fission-decay constant λ f ∼ 8.2x10 -17 yr -1 allows one to use this isotope as a chronometer. By measuring the U concentration in the crystals (by reactor irradiation) and the density of the spontaneous-fission tracks it is relatively easy to calculate the 'fission-track age' if 238 U is the main source of fission tracks. However the fission-track dating of extraterrestrial samples compared with the terrestrial ones has some peculiar features due to presence of a number of other potential track sources except the spontaneous fission of 238 U, such as the spontaneous fission of presently extinct 244 Pu, heavy nuclei of cosmic rays and induced fission by cosmic ray primaries. Only tracks from the spontaneous fission of U and Pu are suitable for fission-track dating. The competing effects of these fissioning elements, whose half-lives differ by a factor of ∼50, form a basis for a fission-track chronology for samples older than ∼ 4.0 Gyr. Over small intervals in time (∼ few x10 8 yr ) the track density from spontaneous fission of 238 U is nearly constant. However, the

  13. Basic physics of the fission process. Chapter 2

    International Nuclear Information System (INIS)

    Michaudon, A.

    1981-01-01

    A general description of the fission process is given with special emphasis on those aspects which are necessary for the understanding of the measurements and calculations of neutron-induced fission cross-sections. Having considered the various phases of the process, some typical properties of the low-energy fission of actinide nuclei are presented and the more specific features of neutron induced fission are examined. (U.K.)

  14. Mass dependence of positive pion-induced fission

    International Nuclear Information System (INIS)

    Khan, H.A.; Khan, N.A.; Peterson, R.J.

    1991-01-01

    Fission cross sections for a range of targets have been measured by solid-state track detectors following 80 and 100 MeV π + bombardment. Fission probabilities have been inferred by comparison to computed reaction cross sections. Fission probabilities for heavy targets agree with those for other probes of comparable energy and with statistical calculations. Probabilities for lighter targets are much above those previously observed or computed. Ternary fission cross sections and multiplicities of light fragments have also been determined

  15. Finite fission chain length and symmetry around prompt-criticality

    International Nuclear Information System (INIS)

    Xie Qilin; Yin Yanpeng; Gao Hui; Huang Po; Fang Xiaoqiang

    2012-01-01

    Probability distribution of finite fission chain length was derived by assuming that all neutrons behave identically. Finite fission chain length was also calculated using a zero-dimension Monte-Carlo method based on point kinetics. Then symmetry of finite fission chain length probability distribution around prompt-criticality was deduced, which helps understanding the emission rate of delayed neutrons and initiation of fission chain in super-prompt-critical system. (authors)

  16. Calculation scheme for boiling water reactors cores; Methode de calcul des coeurs de reacteurs a eau bouillante par le systeme saphyr

    Energy Technology Data Exchange (ETDEWEB)

    Marsault, Ph [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SERSI), 13 - Saint-Paul-lez-Durance (France); Nicolas, A; Lenain, R; Richebois, E; Royer, E; Caruge, D [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France); Blaise, P [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPEX), 13 - Saint-Paul-lez-Durance (France); Gastaldi, B; Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs (DER/SPRC), 13 - Saint-Paul-lez-Durance (France)

    1999-07-01

    Boiling Water Reactors represent one third of the world's reactors. They are presently evolving towards greater simplification, allowing a reduction in the costs of operation, improved safety and a relative flexibility in their capacity to accommodate 100% MOX cores. The CEA, in a combined effort with its partners, the COGEMA and the EDF, would like to assess the interest of this reactor type, especially on this last point. A definition program and subsequent qualification of the calculation scheme have been undertaken. We are presenting here the specific features inherent in the calculation of these reactors, in comparison to PWRs, as well as the first results of the program. (authors)

  17. Fission products collecting devices

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1979-01-01

    Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

  18. Energy released in fission

    International Nuclear Information System (INIS)

    James, M.F.

    1969-05-01

    The effective energy released in and following the fission of U-235, Pu-239 and Pu-241 by thermal neutrons, and of U-238 by fission spectrum neutrons, is discussed. The recommended values are: U-235 ... 192.9 ± 0.5 MeV/fission; U-238 ... 193.9 ± 0.8 MeV/fission; Pu-239 ... 198.5 ± 0.8 MeV/fission; Pu-241 ... 200.3 ± 0.8 MeV/fission. These values include all contributions except from antineutrinos and very long-lived fission products. The detailed contributions are discussed, and inconsistencies in the experimental data are pointed out. In Appendix A, the contribution to the total useful energy release in a reactor from reactions other than fission are discussed briefly, and in Appendix B there is a discussion of the variations in effective energy from fission with incident neutron energy. (author)

  19. Conventional method for the calculation of the global energy cost of buildings; Methode conventionnelle de calcul du cout global energetique des batiments

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-05-01

    A working group driven by Electricite de France (EdF), Chauffage Fioul and Gaz de France (GdF) companies has been built with the sustain of several building engineering companies in order to clarify the use of the method of calculation of the global energy cost of buildings. This global cost is an economical decision help criterion among others. This press kit presents, first, the content of the method (input data, calculation of annual expenses, calculation of the global energy cost, display of results and limitations of the method). Then it fully describes the method and its appendixes necessary for its implementation: economical and financial context, general data of the project in progress, environmental data, occupation and comfort level, variants, investment cost of energy systems, investment cost for the structure linked with the energy system, investment cost for other invariant elements of the structure, calculation of consumptions (space heating, hot water, ventilation), maintenance costs (energy systems, structure), operation and exploitation costs, tariffs and consumption costs and taxes, actualized global cost, annualized global cost, comparison between variants. The method is applied to a council building of 23 flats taken as an example. (J.S.)

  20. Thorium-uranium fission radiography

    Science.gov (United States)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  1. HAC and fission reactors

    International Nuclear Information System (INIS)

    Fujiwara, I.; Moriyama, H.; Tachikawa, E.

    1984-01-01

    In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

  2. New fission-neutron-spectrum representation for ENDF

    International Nuclear Information System (INIS)

    Madland, D.G.

    1982-04-01

    A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

  3. What can be learnt from the channel analysis of the 232Th neutron fission cross section

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Jary, J.; Trochon, J.; Boldeman, J.W.; Musgrove, A.R. de L.

    1979-10-01

    Channel analyses of the neutron fission cross section of 232 Th have been made in two laboratories. The calculated fission cross sections and fission fragment anisotropies are compared with the experimental data. Despite some differences in the methods used, the conclusions on the physical aspects of the fission process are very similar

  4. Reactivity coefficients by perturbation theory; Calcul des coefficients de re activite par la theorie des perturbations; Koehffitsienty reaktivnosti, opredelennye pri pomoshchi teorii vozmushchenij; Determinacion, de coeficientes de reactividad con ayuda de la teoria de las perturbaciones

    Energy Technology Data Exchange (ETDEWEB)

    Webster, J W [International Atomic Energy Agency, Vienna (Austria)

    1962-03-15

    The development of the formulae of perturbation theory provides a good opportunity to use one of the principal devices of mathematical heuristics, i.e. proceeding by analogy from something that is simple to something that is more complicated. This paper: (a) Reviews the formulation of perturbation theory as a method of calculating reactivity coefficients; this consists mainly of developing the differential equation for the adjoint flux, as a continuous function of position and lethargy, by proceeding by analogy from the one-group differential equation for adjoint flux. (b) Presents an application of the two-group form of perturbation theory to a boiling-mercury-cooled fast-breeder reactor (MCBR). It is seen that the net Hg density coefficient of reactivity for the first-design-try for the MCBR is negative for some regions and positive for others. However, it is negative for regions of highest statistical weight and where the density change for a power change would be the greatest. The overall Hg density coefficient is thus negative, i.e. the void coefficient is positive-an unsafe condition. It can be easily seen from the two-group formulation what design changes had to be made to obtain a design which would have a negative void coefficient. It developed in subsequent investigations that there were such design changes that could be made and a design of the MCBR with a negative void coefficient was eventually achieved. (author) [French] La mise au points des formules de la theorie des perturbations fournit une bonne occasion d'employer en mathematiques une des principales regles de la methode heuristique, celle qui consiste a proceder par analogie du simple au complexe. L'auteur du memoire: a) Expose une methode permettant de calculer les coefficients de reactivite en utilisant les formules de la theorie des perturbations; en substance, cette methode consiste a developper l'equation differentielle relative au flux adjoint en fonction continue de la position et de la

  5. Etude des méthodes de calcul des pressions de contact dans les roulements à pistes int́egŕees des boîıtes de transmission aéronautiques

    OpenAIRE

    Zamponi , Laurent; Mermoz , Emmanuel; Linares , jean-marc

    2007-01-01

    International audience; Study of methodologies for contact pressure calculation in integrated raceways bearings for aeronautical gearbox. More and more designers use advanced calculation methodologies for the optimisation of mechanical part performance. This is even truer in the aeronautic industry, where the power weight ratio must be as low as possible. Some CAD softwares (CATIA, SOLIDWORKS, PRO ENGINEER, etc.) provide the possibility to do complex computation. However these kinds of tools ...

  6. Fission Research at IRMM

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2010-03-01

    Full Text Available Fission Research at JRC-IRMM has a longstanding tradition. The present paper is discussing recent investigations of fission fragment properties of 238 U(n,f, 234 U(n,f, prompt neutron emission in fission of 252 Cf(SF as well as the prompt fission neutron spectrum of 235 U(n,f and is presenting the most important results.

  7. 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

  8. JNDC nuclear data library of fission products

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

    1983-10-01

    The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

  9. Measurement of isotopic cross sections of the fission fragments produced in 500 AMeV {sup 208}Pb + p reaction; Etude de la production des fragments de fission issus de la reaction {sup 208}Pb + p a 500 AMeV

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Dominguez, B

    2003-03-01

    The aim of this work is the study of the fission fragments produced in the spallation reaction {sup 208}Pb + p at 500 AMeV. The fission fragments from Z=23 up to Z=59 have been detected and identified by using the inverse kinematics technique with the high-resolution spectrometer FRS. The production cross sections and the recoil velocities of 430 nuclei have been measured. The measured data have been compared with previous data. The isotopic distributions show a high precision. However, the absolute value of the fission cross section is higher than expected. From the experimental data the characteristics of the average fissioning system have been reconstructed (Z{sub fis}, A{sub fis}, E*{sub fis}). In addition, the number of post-fission neutrons emitted from the fission fragments, v{sub post}, has been determined by using a new method. The experimental data have been compared to the two-steps models describing the spallation reaction. The impact of the model parameters on the observables has been analysed and the reasons Leading to the observed differences between the codes are also presented. This analyse shows a good agreement with the INCL4+ABLA code. (author)

  10. Fission energy of uranium isotopes and transuranium elements

    International Nuclear Information System (INIS)

    Nemirovskij, P.Eh.; Manevich, L.G.

    1981-01-01

    A comparison is made between the prompt fission energy, Esub(pr), calculated from the mass and binding energy spectrum and the Esub(pr) value obtained from the experimental data on the kinetic energy of fragments, the energy of prompt neutrons and prompt γ-quanta. Basing on the data on β-decay chains of fission fragments, the energies of neutrinos, γ-quanta and β-electrons are obtained, which permits to calculate the actual energy released during fission. The calculations are performed for thermal neutron-induced fission, fast-neutron induced fission and for fission after bombardment with 14 MeV neutrons. The available experimental data on the fission fragment kinetic energy, prompt γ-quanta energy and fission neutron energy are presented. The comparison of the Esub(pr) values obtained experimentally for the thermal-neutron-induced fission with the calculated Esub(pr) value shows that for 233 U, 239 Pu, 241 Pu the agreement is rather favourable. For 235 U the agreement is within the error limits. As to the Esub(pr) values for the fast-neutron-induced fission, the agreement between the calculated and experimental data for all nuclides is quite good

  11. EXTENDCHAIN: a package of computer programs for calculating the buildup of heavy metals, fission products, and activation products in reactor fuel elements

    International Nuclear Information System (INIS)

    Robertson, M.W.

    1977-01-01

    Design of HTGR recycle and refabrication facilities requires a detailed knowledge of the concentrations of around 400 nuclides which are segregated into four different fuel particle types. The EXTENDCHAIN package of computer programs and the supporting input data files were created to provide an efficient method for calculating the 1600 different concentrations required. The EXTENDCHAIN code performs zero-dimensional nuclide burnup, decay, and activation calculations in nine energy groups for up to 108 nuclides per run. Preparation and handling of the input and output for the sixteen EXTENDCHAIN runs required to produce the desired data are the most time consuming tasks in the computation of the spent fuel element composition. The EXTENDCHAIN package of computer programs contains four codes to aid in the preparation and handling of these data. Most of the input data such as cross sections, decay constants, and the nuclide interconnection scheme will not change when calculating new cases. These data were developed for the life cycle of a typical HTGR and stored on archive tapes for future use. The fuel element composition for this typical HTGR life has been calculated and the results for an equilibrium recycle reload are presented. 12 figures, 7 tables

  12. Calcul des efforts de deuxième ordre à très haute fréquence sur des plates-formes à lignes tendues Computing High-Frequency Second Order Loads on Tension Leg Platforms

    Directory of Open Access Journals (Sweden)

    Chen X.

    2006-11-01

    Full Text Available Le problème considéré ici est celui de l'évaluation des efforts excitateurs de deuxième ordre (en mode somme, c'est-à-dire prenant place aux sommes deux à deux des fréquences de houle sur des plates-formes à lignes tendues. Ces efforts sont tenus pour responsables de comportements résonnants (en roulis, tangage et pilonnement observés lors d'essais en bassin et pourraient réduire sensiblement la durée de vie en fatigue des tendons. Des résultats sont tout d'abord présentés pour une structure simplifiée, consistant en 4 cylindres verticaux reposant sur le fond marin. L'intérêt de cette géométrie est que tous les calculs peuvent être menés à terme de façon quasi analytique. Les résultats obtenus permettent d'illustrer le haut degré d'interaction entre les colonnes, et la faible décroissance du potentiel de diffraction de deuxième ordre avec la profondeur. On présente ensuite des résultats pour une plate-forme réelle, celle de Snorre. Tension Leg Platforms (TLP's are now regarded as a promising technology for the development of deep offshore fields. As the water depth increases however, their natural periods of heave, roll and pitch tend to increase as well (roughly to the one-half power, and it is not clear yet what the maximum permissible values for these natural periods can be. For the Snorre TLP for instance, they are only about 2. 5 seconds, which seems to be sufficiently low since there is very limited free wave energy at such periods. Model tests, however, have shown some resonant response in sea states with peak periods of about 5 seconds. Often referred to as springing , this resonant motion can severely affect the fatigue life of tethers and increase their design loads. In order to calculate this springing motion at the design stage, it is necessary to identify and evaluate both the exciting loads and the mechanisms of energy dissipation. With the help of the French Norwegian Foundation a joint effort was

  13. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Nazarewicz, Witold

    2009-01-01

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation's nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  16. Fission mass yields of excited medium heavy nuclei

    International Nuclear Information System (INIS)

    Sandulescu, A.; Depta, K.; Herrmann, R.; Greiner, W.; Scheid, W.

    1985-01-01

    The mass distributions resulting from the fission of excited medium mass nuclei are discussed on the basis of the fragmentation theory. It is shown that very asymmetric fission events can be expected with rates which are only a few orders of magnitude smaller than the rates for symmetric fission. As an example a calculation of the fission mass distribution of the excited 172 Yb compound nucleus is presented. This mass distribution reveals observable structures over the entire range of the mass asymmetry due to valleys in the potential energy surface for fission fragments with closed proton and neutron shells

  17. Migration of fission products in UO2. Final report

    International Nuclear Information System (INIS)

    Prussin, S.G.; Olander, D.R.

    1995-01-01

    Results of an experimental and calculational effort to examine the fundamental mechanisms of fission product migration in and release from polycrystalline uranium dioxide are reported. The experiments were designed to provide diffusion parameters for the representative fission products tellurium, iodine, xenon, molybdenum and ruthenium under both reducing and oxidizing conditions. The calculational effort applied a new model of fission product release from reactor fuel that incorporates grain growth as well as grain boundary and lattice diffusion

  18. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high-temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products

  19. Heterogeneous Calculation of {epsilon}

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, Alf

    1961-02-15

    A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of {epsilon}. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer.

  20. Heterogeneous Calculation of ε

    International Nuclear Information System (INIS)

    Jonsson, Alf

    1961-02-01

    A heterogeneous method of calculating the fast fission factor given by Naudet has been applied to the Carlvik - Pershagen definition of ε. An exact calculation of the collision probabilities is included in the programme developed for the Ferranti - Mercury computer

  1. Les espérances de vie appliquées à des statuts spécifiques : historique des indicateurs et des méthodes de calcul

    OpenAIRE

    Nicolas Brouard; Jean-Marie Robine; Emmanuelle Cambois

    1998-01-01

    Cambois (Emmanuelle), Robin? (Jean-Marie), Brouard (Nicolas).- Life expectancies applied to specific statuses: a history of the indicators and the methods of calculation Indicators of life expectancy applied to specific statuses, such as the state of health or professional status, were introduced at the end of the 1930s and are currently the object of renewed interest. Because they relate mortality to different domains (health, professional activity) applied life expectancies reflect simultan...

  2. Angular momenta of fission fragments in the {alpha}-accompanied fission of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M.; Kliman, J.; Krupa, L.; Morhac, M. [Slovak Academy of Sciences, Department of Nuclear Physics, Bratislava (Slovakia); Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Hamilton, J.H.; Kormicki, J.; Ramayya, A.V.; Hwang, J.K.; Luo, Y.X.; Fong, D.; Gore, P. [Vanderbilt University, Department of Physics, Nashville, TN (United States); Ter-Akopian, G.M.; Oganessian, Yu.Ts.; Rodin, A.M.; Fomichev, A.S.; Popeko, G.S. [Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Daniel, A.V. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Rasmussen, J.O.; Macchiavelli, A.O.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Donangelo, R.; Cole, J.D.

    2005-06-01

    For the first time, average angular momenta of the ternary fission fragments {sup 100,102}Zr, {sup 106}Mo, {sup 144,146}Ba and {sup 138,140,142}Xe from the {alpha}-accompanied fission of {sup 252}Cf were obtained from relative intensities of prompt {gamma}-ray transitions with the use of the statistical model calculation. Average values of the angular momenta were compared with the corresponding values for the same fission fragments from the binary fission of {sup 252}Cf. Results indicate the presence of a decreasing trend in the average values of angular momenta induced in ternary fission fragments compared to the same binary fission fragments. On the average, the total angular momentum extracted for ternary fission fragments is {proportional_to}1.4{Dirac_h} lower than in binary fission. Consequently, results indicate that the mechanism of the ternary {alpha}-particles emission may directly effect an induction of angular momenta of fission fragments, and possible scenarios of such mechanisms are discussed. Further, the dependence of the angular momenta of {sup 106}Mo and {sup 140}Xe on the number of emitted neutrons from correlated pairs of primary fragments was obtained also showing a decreasing dependence of average angular momenta with increasing number of emitted neutrons. Consequences are briefly discussed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  4. Transport properties of fission product vapors

    International Nuclear Information System (INIS)

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

  5. Metallurgical study and phase diagram calculations of the Zr-Nb-Fe-(Sn,O) system; Etude metallurgique et calculs des diagrammes de phases des alliages base zirconium du systeme: Zr-Nb-Fe-(O,Sn)

    Energy Technology Data Exchange (ETDEWEB)

    Toffolon, C. [CEA/Saclay, Dept. d' Etudes du Comportement des Materiaux (DECM), 91 - Gif-sur-Yvette (France)]|[Paris-6 Univ., 75 (France)

    2000-07-01

    The Framatome M5{sup TM} Zr-Nb-O alloy with small amounts of Fe is of interest for nuclear applications (PWR fuel cladding).The behaviour of this kind of alloy for in-service conditions strongly depends on the microstructure. Therefore, a metallurgical study of alloys of the Zr-Nb-Fe-(O-Sn) system has been developed in order to study the influence of chemical composition variabilities of Nb, Fe and O and thermal treatments on the resultant microstructure. In order to get some insight on the physical metallurgy of Zr-Nb-Fe-(Sn,O) alloys and to minimize the experiments, it is useful to build a thermodynamic database. With this object, it was necessary to re-optimize and to calculate the low order binary systems such as Fe-Nb and Nb-Sn in order to assess the Zr-Nb-Fe-(Sn,O) system. Then, the experimental studies concerned: the influence of small variations in Nb and O contents on the {alpha}/{beta} transus temperatures. A comparison between experimental results and thermodynamic predictions showed a good agreement; the precipitation kinetics of {beta}Nb and intermetallic phases in the {alpha} phase domain. These experiments showed that the kinetics depends on the initial metallurgical conditions; the determination of the crystallographic structure and the stoichiometry of the ternary Zr-Nb-Fe intermetallic compounds as a function of the temperature. Finally, these experimental data were used to propose a first assessment of the Zr-Nb-Fe(O{approx}1200 ppm) system. (author)

  6. Rupture of the neck in nuclear fission

    International Nuclear Information System (INIS)

    Davies, K.T.R.; Managan, R.A.; Nix, J.R.; Sierk, A.J.

    1977-01-01

    We introduce a degree of freedom to describe the rupture of the neck in nuclear fission and calculate the point at which the neck ruptures as the nucleus descends dynamically from its fission saddle point. This is done by mentally slicing the system into two portions at its minimum neck radius and calculating the force required to separate the two portions while keeping their shapes fixed. This force is obtained by differentiating with respect to separation the sum of the Coulomb and nuclear interaction energies between the two portions. For nuclei throughout the Periodic Table we calculate this force along dynamical paths leading from the fission saddle point. The force is initially attractive but becomes repulsive when the neck reaches a critical size. For actinide nuclei the neck radius at which rupture occurs is about 2 fm. This increases the calculated translational kinetic energy of the fission fragments at infinity relative to that calculated for scission occurring at zero neck radius. With the effect of neck rupture taken into account, we calculate and compare with experimental results fission-fragment kinetic energies for two types of nuclear dissipation: ordinary two-body viscosity and one-body dissipation

  7. Sur la prise en compte du mode de financement dans les calculs de rentabilité. La méthode des intérêts seulement déductibles Financing Mix and Capital Budgeting Techniques. the Shadow Interest Method

    OpenAIRE

    Babusiaux D.; Karnik J. L.

    2006-01-01

    La pratique des calculs de rentabilité avec shadow interest se développe depuis quelques années dans les entreprises pétrolières, principalement dans le secteur exploration et développement des gisements. L'objet de cet article est de replacer cette approche parmi les méthodes classiques (calculs de rentabilité globale et de rentabilité des capitaux propres). Une relation simple entre les revenus actualisés relatifs à chacune des approches est obtenue sous certaines hypothèses simplificatrice...

  8. Theoretical descriptions of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1991-01-01

    Brief descriptions are given of the observables in neutron emission in fission together with early theoretical representations of two of these observables, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity ν-bar p . This is followed by summaries, together with examples, of modern approaches to the calculation of these two quantities. Here, emphasis is placed upon the predictability and accuracy of the new approaches. In particular, the dependencies of N(E) and ν-bar p upon the fissioning nucleus and its excitation energy are discussed. Then, recent work in multiple-chance fission and other recent work involving new measurements are presented and discussed. Following this, some properties of fission fragments are mentioned that must be better known and better understood in order to calculate N(E) and ν-bar p with higher accuracy than is currently possible. In conclusion, some measurements are recommended for the purpose of benchmarking simultaneous calculations of neutron emission and gamma emission in fission. (author). 32 refs, 26 figs

  9. Detection of radioactive gases in the CO{sub 2} cooling the reactors G 2 - G 3; Detection des gaz radioactifs dans le CO{sub 2} de refroidissement des piles G2 - G3

    Energy Technology Data Exchange (ETDEWEB)

    Pouthier, J; Rossi, J [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1968-07-01

    The carbon dioxide cooling the reactors G2 - G3 contains activation gases and fission gases. It is of interest to know their concentration, for example to be able to deduce rapidly the norms which would have to be applied in the case of an incident in the circuit. Gas-phase chromatography is applied daily for carrying out analyses. The chromatogram has separate peaks due to tritium, argon 41, krypton 85 and the 133 and 135 isotopes of xenon. By integrating each peak it is possible to calculate the specific activity of each product. The construction of an apparatus for carrying out continuous measurements is under consideration. (authors) [French] Le gaz carbonique, refroidissant les reacteurs G2 - G3, contient des gaz d'activation et des gaz de fission. Il est interessant de connaitre leur teneur par exemple pour etre en mesure de deduire rapidement les normes qu'il y aurait lieu d'appliquer en cas d'incidents sur le circuit. La methode de chromatographie en phase gazeuse est employee quotidiennement pour faire des analyses. Le chromatogramme se presente sous forme de pics distincts dus au tritium, a l'argon 41, au krypton 85 et aux isotopes 133 et 135 du xenon. L'integration de chaque pic permet de calculer l'activite specifique de chaque compose. Il est envisage de construire un appareil pour des mesures en continu. (auteurs)

  10. Investigation of exotic fission modes

    International Nuclear Information System (INIS)

    Poenaru, D. N.; Gherghescu, R. A.; Greiner, W.; Nagame, Y.; Hamilton, J. H.; Ramayya, A. V.

    2002-01-01

    Fission approach to the cluster radioactivities and α-decay has been systematically developed during the last two decades. A more complex process, the ternary fission, was observed since 1946 both in neutron-induced and spontaneous fission. We obtained interesting results concerning the binary fission saddle-point reflection asymmetric nuclear shapes, and we can explain how a possible nuclear quasimolecular state is formed during the 10 Be accompanied cold fission of 252 Cf. The equilibrium nuclear shapes in fission theory are usually determined by minimizing the deformation energy for a given surface equation. We developed a method allowing to obtain a very general saddle-point shape as a solution of a differential equation without an a priori introduction of a shape parametrization. In the approach based on a liquid drop model (LDM), saddle-point shapes are always reflection symmetric: the deformation energy increases with the mass-asymmetry parameter η = (A 1 - A 2 )/(A 1 + A 2 ). By adding the shell corrections to the LDM deformation energy, we obtained minima at a finite mass asymmetry for parent nuclei 238 U, 232,228 Th in agreement with experiments. This correction was calculated phenomenologically. A technique based on the fragment identification by using triple γ coincidences in the large arrays of Ge-detectors, like GAMMASPHERE, was employed at Vanderbilt University to discover new characteristics of the fission process, and new decay modes. The possibility of a whole family of new decay modes, the multicluster accompanied fission, was envisaged. Besides the fission into two or three fragments, a heavy or superheavy nucleus spontaneously breaks into four, five or six nuclei of which two are asymmetric or symmetric heavy fragments and the others are light clusters, e.g. α-particles, 10 Be, 14 C, or combinations of them. Examples were presented for the two-, three- and four cluster accompanied cold fission of 252 Cf and 262 Rf, in which the emitted

  11. Fifty years of nuclear fission: Nuclear data and measurements series

    International Nuclear Information System (INIS)

    Lynn, J.E.

    1989-06-01

    This report is the written version of a colloquium first presented at Argonne National Laboratory in January 1989. The paper begins with an historical preamble about the events leading to the discovery of nuclear fission. This leads naturally to an account of early results and understanding of the fission phenomena. Some of the key concepts in the development of fission theory are then discussed. The main theme of this discussion is the topography of the fission barrier, in which the interplay of the liquid-drop model and nucleon shell effects lead to a wide range of fascinating phenomena encompassing metastable isomers, intermediate-structure effects in fission cross-sections, and large changes in fission product properties. It is shown how study of these changing effects and theoretical calculations of the potential energy of the deformed nucleus have led to broad qualitative understanding of the nature of the fission process. 54 refs., 35 figs

  12. Comparison of Knolls Atomic Power Laboratory Clean Critical and Subcritical Experiments with Calculations; Etude Comparative des Calculs et des Experiences Critiques Non Empoisonnees et Souscritiques, Faits au Knolls Atomic Power Laboratory; Sravnenie chistykh kriticheskikh i podkriticheskikh opytov v nolle s raschetnymi dannymi; Comparacion de los Resultados Experimentales del Knolls Atomic Power Lamra Tory con los Valores Calculados Teoricamente en Conjuntos Criticos Limpios y Subcriticos

    Energy Technology Data Exchange (ETDEWEB)

    Bach, D. R.; Bistline, J. A.; Bunch, S. I.; Cerbone, R. J.; Feiner, F.; Skolnik, W.; Slovacek, R. E.; Weinstein, S. [Knolls Atomic Tower Laboratory, Schenectady, NY (United States)

    1964-02-15

    conditions show agreement to within a few per cent of the measured decay constants in the range from 100 to 45 000 s{sup -1}. Comparison of the slow chopper spectrum measurements with calculated spectra using the most recent bound-hydrogen kernels for both water and polyethylene show good agreement except for highly absorbing water systems. In the latter case, the measured spectra are apparently even softer than the free atom calculated spectra. (author) [French] Le Knolls Atomic Power Laboratory a entrepris d'executer un programme intensif d'experiences et de calculs connexes pour determiner si les methodes de calcul et les donnees nucleaires conviennent pour les assemblages hydrogenes. Le programme comporte trois grandes parties: 1. Experiences critiques non empoisonnees avec combustible homogene solide pour lesquelles le rapport metal/hydrogene (en volume) est eleve; 2. Experiences critiques non empoisonnees avec combustible en plaques portant sur des gammes etendues des caracteristiques suivantes: dimension du coeur, rapport metal/eau. densite du combustible et temperature; 3. Mesures de la constante de decroissance des neutrons instantanes et du spectre a l'etat stationnaire sur des assemblages sous-critiques appartenant a la meme gamme de compositions. Pour les experiences visees sous 1, les elements combustibles sont des comprimes de poudres de ZrO{sub 2}, et d'UO{sub 2} fortement enrichi, melanges uniformement a de la paraffine. Avec des blocs reflecteurs dans lesquels le rapport ZrO{sub 2}/parraffine est le meme que dans les blocs combustibles, on a pu construire des coeurs reflechis avec des caracteristiques de ralentissement uniformes. L'ensemble critique est concu de telle facon que les configurations critiques soient de simples parallelepipedes rectangles sans canaux pour les barres de securite ou de reglage. On a fait des etudes de dimension critique sur des ensembles (avec et sans reflecteur) caracterises par des pertes en neutrons rapides, et des comparaisons

  13. Ternary Fission of U{sup 235} by Resonance Neutrons; Fission Ternaire de {sup 235}U par des Neutrons de Resonance; 0422 0420 041e 0419 041d 041e 0415 0414 0415 041b 0415 041d 0418 0415 0423 0420 0410 041d 0410 -235 041d 0410 0420 0415 0417 041e 041d 0410 041d 0421 041d 042b 0425 041d 0415 0419 0422 0420 041e 041d 0410 0425 ; Fision Ternaria del {sup 235}U por Neutrones de Resonancia

    Energy Technology Data Exchange (ETDEWEB)

    Kvitek, I.; Popov, Ju. P.; Rjabov, Ju. V. [Ob' edinennyj Institut Jadernyh Issledovanij, Dubna, SSSR (Russian Federation)

    1965-07-15

    range 0.1 to 0.2 eV. The paper discusses the possible effect of the (n, {alpha}) reaction on the results obtained in experiments not using the coincidence of a long-range particle with a fission fragment. (author) [French] On a publie recemment des etudes dans lesquelles on signale que, lors du passage d'une resonance a une aune, le rapport entre les sections efficaces de fission ternaire et binaire de {sup 235}U subit des variations considerables. Or, les auteurs n'ont constate aucune variation de ce genre ni pour {sup 233}U ni pour {sup 2}'3{sup 9}Pu. Le memoire fait etat de recherches sur la fission ternaire de {sup 285}U par neutrons de 0,1 a 30 eV. Contrairement a d'autres travaux consacres a la fission ternaire de chaque fission ternaire a ete identifiee par coincidence entre l'un des fragments de fission et une particule legere a long parcours. Grace a ce procede, les fissions ternaires ont pu Etre separees de la contribution qui pourrait etre due a la reaction (n, {alpha}). Les mesures ont ete faites dans le reacteur a neutrons rapides de l'Institut unifie de recheiches nucleaires par la methode du temps de vol. Les auteucs ont utilise a cet effet une base de vol de 100 m, ce qui a permis d'obtenir un pouvoir de resolution de 0,6 {mu}s/m. Pour enregistrer les fragments de fission et la particule legere a long parcours, ils se sont servis de compteurs a scintillateur gazeux remplis de xenon sous une pression de 2 atm. Une couche de {sup 235}U enrichi d'environ 300 cm{sup 2} et de 2 mg/cm{sup 2} d'epaisseur a ete appliquee sur une feuille d'aluminium de 20 {mu}m. Dans le volume de gaz on a enregistre d'un cote de la feuille les scintillations dues aux fragments de fission et, de l'autre cote, celles qui pro venaient des particules legeres a long parcours. En vue d'evaluer le bruit de fond (par exemple les coincidences entre l'Impulsion provoquee par un fragment et l'impulsion due a un photon de fission gamma ou un proton provenant de la reaction (n, p) dans la

  14. Contribution and limits of geochemical calculation codes to evaluate the long term behavior of nuclear waste glasses; Apports et limites des modeles geochimiques pour l'evaluation du comportement a long terme des verres de confinement des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, B; Crovisier, J L [Universite Louis Pasteur, Centre de Geochimie de la Surface, CNRS ULP, Ecole et Observatoire des Sciences de la Terre, 67 - Strasbourg (France)

    1997-07-01

    Geochemical models have been intensively developed by researchers since more than twenty five years in order to be able to better understand and/or predict the long term stability/instability of water-rock systems. These geochemical codes were ail built first on a thermodynamic approach deriving from the application of Mass Action Law. The resulting first generation of models allowed to detect or predict the possible mass transfers (thermodynamic models) between aqueous and mineral phases including irreversible dissolutions of primary minerals and/or precipitation near equilibrium of secondary mineral phases. The recent development of models based on combined thermodynamics and kinetics opens the field of Lime dependent reactions prediction. This is crucial if one thinks to combine geochemical and hydrological studies in the so-called coupled models for transport and reaction calculations. All these models are progressively applied to the prediction of long term behavior of mineral phases, and more specifically glasses. In order to succeed in chat specific extension of the models, but also the data bases, there is a great need for additional new data from experimental approaches and from natural analogues. The modelling approach appears than also very useful in order to interpret the results of experimental data and to relate them to long term data extracted from natural analogues. Specific functions for modelling solid solution phases mat' also be used for describing the products of glasses alterations. (authors)

  15. Fission-product yields for thermal-neutron fission of curium-243

    International Nuclear Information System (INIS)

    Breederland, D.G.

    1982-01-01

    Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

  16. Burn up physics; Physique des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    requires samples of the order of several kilograms only. The relationships between these measurements and the investigations of lattices are discussed, and an outline is given of the way of carrying out the systematic study of fuels of various compositions. The method has been successfully applied to the systematic study of irradiated fuels (analysed independently by the methods mentioned above) thus giving the possibility of measuring in situ the absorption of fission products. (author) [French] Cette communication expose un ensemble d'etudes theoriques et d'experiences, effectuees au CEA et destinees a faire progresser la connaissance de l'evolution de la reactivite (au cours de l'irradiation du combustible) dans les reacteurs a uranium naturel ou faiblement enrichi.,. On rappelle les difficultes de l'experimentation directe sur des masses importantes de combustible irradie - en particulier dans les reacteurs de puissance en fonctionnement - et on souligne la necessite d'experiences a caractere fondamental distinguant: d'une part l'evolution de la composition des combustibles (chaines de noyaux lourds, produits de fission), d'autre part l'effet des modifications de composition sur le bilan de neutrons. Avant de presenter trois categories d'experiences que l'on est conduit a entreprendre, on rappelle l'importance des problemes lies aux spectres de neutrons et on decrit rapidement les methodes pratiques de calcul utilisees. L'irradiation systematique de quelques types de combustibles, suivie de leur analyse chimique et isotopique est en cours depuis plusieurs annees. On donne un apercu de l'ensemble du programme experimental et on decrit les moyens et les methodes mis en oeuvre: chaine {alpha}, {beta}, {gamma} pour la preparation des echantillons, dosage du Plutonium par coulommetrie et double dilution isotopique, separation du Bore utilise dans certains cas pour la mesure des densites de neutrons integrees. On discute sur quelques exemples l'interpretation des mesures

  17. Étude des distributions de contraintes autour d'un puits pétrolier. Application au calcul des pressions d'initiation de fracturation hydraulique Research on Stress Distribution Around an Oil Or Gas Well. Application to Calculating Pressures for Initiating Hydraulic Fracturing

    Directory of Open Access Journals (Sweden)

    Skoczylas F.

    2006-11-01

    Full Text Available Des méthodes de résolution analytiques et numériques ont été proposées, pour déterminer les distributions de contraintes autour de forages pétroliers dans le cas de matériaux peu résistants. Le calcul de ces distributions aboutit à la détermination des pressions de fracturation. Le critère de plasticité de Mohr-Coulomb a été choisi pour la résolution analytique. Cette étude permet d'établir trois schémas types de distribution de contraintes, au moment de la fracturation, pour un régime permanent d'écoulement. La géométrie axisymétrique du modèle a permis d'utiliser une méthode numérique par schéma aux différences finies. La loi de comportement du matériau est déduite du critère de plasticité de Drucker. Cette résolution permet le chargement incrémental en pression dans le forage et de simuler une injection rapide par un couplage simplifié fluide-squelette. Le modèle élastoplastique à deux surfaces de charge de Lade-Shao a finalement été retenu pour décrire l'écrouissage du matériau. La comparaison des résultats des différentes méthodes de calcul s'effectue sur la base de l'influence respective du coefficient de poussée au repos Ko, du coefficient de Poisson et de la résistance à la compression, Co, du matériau. Analytical and numerical methods are proposed, in order to determine stresses distribution around borehole for axisymetric problems in the case of low strengh materials. Calculation of these distributions leads to the determination of hydraulic fracture-pressure. Analytical solutions were found in the case of Mohr-Coulomb criterion and steady flow. These solutions leads to three typical stresses distributions when hydraulic fracture takes place. Because problem's axisymetry, the finite difference method was used. The constitutive law was deduced using Drücker plasticity criterion with an associated flow rule for elastic perfectly plastic behaviour. Such a numerical method can take into

  18. Generalized Energy-Dependent Q Values for Fission

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, R

    2010-03-31

    We extend Madland's parameterization of the energy release in fission to obtain the dependence of the fission Q value for major and minor actinides on the incident neutron energies in the range 0 {le} E{sub n} {le} 20 MeV. Our parameterization is based on the actinide evaluations recommended for the ENDF/B-VII.1 release. This paper describes the calculation of energydependent fission Q values based on the calculation of the prompt energy release in fission by Madland. This calculation was adopted for use in the LLNL ENDL database and then generalized to obtain the prompt fission energy release for all actinides. Here the calculation is further generalized to the total energy release in fission. There are several stages in a fission event, depending on the time scale. Neutrons and gammas may be emitted at any time during the fission event.While our discussion here is focussed on compound nucleus creation by an incident neutron, similar parameterizations could be obtained for incident gammas or spontaneous fission.

  19. A new set of parameters for 5 Gaussian fission yields systematics

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    2003-01-01

    A new set of parameters for 5 Gaussian-type fission yields systematics has been proposed for applying to high energy neutron or proton fission and to various kinds of fissioning systems including minor actinides. The mass yields calculated using the systematics were compared with various kinds of measured data including the fission with incident energy higher than 100 MeV and the fission of minor actinide nuclides. The comparisons showed rather good agreement between the calculated values and measured ones for various kinds of fissioning systems. (author)

  20. Mass distributions in nucleon-induced fission at intermediate energies

    CERN Document Server

    Duijvestijn, M C; Hambsch, F J

    2001-01-01

    Temperature-dependent fission barriers and fission-fragment mass distributions are calculated in the framework of the multimodal random neck-rupture model (MM-RNRM). It is shown how the distinction between the different fission modes disappears at higher excitation energies, due to the melting of shell effects. The fission-fragment mass yield calculations are coupled to the nuclear reaction code ALICE-91, which takes into account the competition between the other reaction channels and fission. With the combination of the temperature-dependent MM-RNRM and ALICE-91 nucleon-induced fission is investigated at energies between 10 and 200 MeV for nuclei varying from Au to Am. (72 refs).

  1. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  2. General Description of Fission Observables - JEFF Report 24. GEF Model

    International Nuclear Information System (INIS)

    Schmidt, Karl-Heinz; Jurado, Beatriz; Amouroux, Charlotte

    2014-06-01

    The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among the member countries of the OECD Nuclear Energy Agency (NEA) Data Bank to develop a reference nuclear data library. The JEFF library contains sets of evaluated nuclear data, mainly for fission and fusion applications; it contains a number of different data types, including neutron and proton interaction data, radioactive decay data, fission yield data and thermal scattering law data. The General fission (GEF) model is based on novel theoretical concepts and ideas developed to model low energy nuclear fission. The GEF code calculates fission-fragment yields and associated quantities (e.g. prompt neutron and gamma) for a large range of nuclei and excitation energy. This opens up the possibility of a qualitative step forward to improve further the JEFF fission yields sub-library. This report describes the GEF model which explains the complex appearance of fission observables by universal principles of theoretical models and considerations on the basis of fundamental laws of physics and mathematics. The approach reveals a high degree of regularity and provides a considerable insight into the physics of the fission process. Fission observables can be calculated with a precision that comply with the needs for applications in nuclear technology. The relevance of the approach for examining the consistency of experimental results and for evaluating nuclear data is demonstrated. (authors)

  3. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  4. Nuclear Dissipation from Fission Time

    International Nuclear Information System (INIS)

    Gontchar, I.; Morjean, M.; Basnary, S.

    2000-01-01

    Fission times, pre-scission neutron multiplicities and GDR pre-scission γ-ray multiplicities measured for uranium or thorium nuclei formed with temperatures T ∼ 1.8 MeV have been compared with calculations performed with CDSM2, a two-dimensional dynamical model combined with a statistical one. Among the three experimental approaches considered, fission times give access to the most precise pieces of information on nuclear dissipation at high excitation energy. For the temperature range under consideration, an agreement between the model and data is achieved if one-body dissipation is used with a strength factor k red ∼ 0.45 ± 0.10 applied to the wall term for the mononuclear configuration. (authors)

  5. Theoretical models of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1992-01-01

    A brief survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity bar v p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the new models. In particular, the dependencies of N(E) and bar v p upon the fissioning nucleus and its excitation energy are treated. Recent work in the calculation of the prompt fission neutron spectrum matrix N(E,E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of our ability to calculate these observables with confidence, the direction of future theoretical efforts, and limititations to current and future calculations. Finally, recommendations are presented as to which model should be used currently and which model should be pursued in future efforts

  6. Recent progress in the study of fission reaction

    International Nuclear Information System (INIS)

    Blons, J.; Paya, D.; Signarbieux, C.

    1977-01-01

    The different steps of the fission phenomenon are briefly recalled before a more detailed study of the static aspect of the fission barrier crossing. The experiments performed at Saclay during the last few years seem to confirm the calculations according to which a triple humped fission barrier is expected for the thorium isotopes. The last part deals with some dynamical aspects of the process. Recent results are presented which tend to prove that the nuclear viscosity decreases when the excitation energy increases [fr

  7. Daniel Gogny's vision for a microscopic theory of fission

    Science.gov (United States)

    Younes, W.

    2017-05-01

    Daniel Gogny made many contributions to our understanding of nuclear fission over a span of 35 years. This paper reviews some of those contributions, focusing in particular on fission dynamics, the challenges of describing scission in a quantum-mechanical context, and the calculation of fragment properties such as their mass, kinetic, and excitation energy distributions. The generator coordinate method provides the common theoretical framework within which these various aspects of fission are formulated.

  8. 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)

  9. Fission cross section and fission fragment angular distribution for oriented nucleus fission by intermediate energy neutrons (epsilon < or approximately 1 Mev)

    International Nuclear Information System (INIS)

    Barabanov, A.L.; Grechukhin, D.P.

    1985-01-01

    General analysis is conducted, and formulae for fission cross section and angular distribution of fission fragments of oriented nuclei by fast neutrons are presented. Geometrical coefficients making up the formulae permitting to carry out calculations for target nuclei with spins I=3/2, 5/2, 7/2 at interaction energies epsilon < or approximately 1 MeV are tabulated. Results of demonstrative calculation of fission fragment angular distribution of oriented sup(235)U nuclei by 0.1 <= epsilon <= 1.0 MeV neutrons reveal that angular distribution weakly depends on the set of permeability factors of neutron waves applied in the calculations

  10. Differential and integral characteristics of prompt fission neutrons in the statistical theory

    International Nuclear Information System (INIS)

    Gerasimenko, B.F.; Rubchenya, V.A.

    1989-01-01

    Hauser-Feshbach statistical theory is the most consistent approach to the calculation of both spectra and prompt fission neutrons characteristics. On the basis of this approach a statistical model for calculation of differential prompt fission neutrons characteristics of low energy fission has been proposed and improved in order to take into account the anisotropy effects arising at prompt fission neutrons emission from fragments. 37 refs, 6 figs

  11. Calculation of transient overvoltages in a 145-kV-cable termination by using the FDTD-method; Feldstaerkenberechnung bei transienten Spannungen an einem 145-kV-Endverschluss mit Hilfe des FDTD-Verfahrens

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard, G.; Ioffe, A. [Gerhard-Mercator-Univ.-GH Duisburg (Germany). Fachgebiet Elektrische Energieuebertragung

    1997-10-20

    The transient electromagnetic field occured from the overvoltages in a slip-on cable termination was calculated by using the FDTD-method. The electric field strength in a gap between the cable shield and the stress cone was studied. (orig.) [Deutsch] Mit Hilfe des FDTD-Feldberechnungsverfahrens wird das transiente elektromagnetische Feld in einem Aufschiebeendverschluss untersucht, wobei ein Luftspalt zwischen den leitenden Teilen der Kabelader und dem Feldsteuerelement beruecksichtigt wird. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

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

  13. The nuclear fission

    International Nuclear Information System (INIS)

    Fiorentino, J.

    1983-01-01

    The nuclear fission process considering initially the formation of compound nucleus and finishing with radioactive decay of fission products is studied. The process is divided in three parts which consist of the events associated to the nucleus of intermediate transitional state, the scission configuration, and the phenomenum of post scission. (M.C.K.) [pt

  14. Fission gas detection system

    International Nuclear Information System (INIS)

    Colburn, R.P.

    1984-01-01

    A device for collecting fission gas released by failed fuel rods which device uses a filter adapted to pass coolant but to block passage of fission gas bubbles due to the surface tension of the bubbles. The coolant may be liquid metal. (author)

  15. Muon-induced fission

    International Nuclear Information System (INIS)

    Polikanov, S.

    1980-01-01

    A review of recent experimental results on negative-muon-induced fission, both of 238 U and 232 Th, is given. Some conclusions drawn by the author are concerned with muonic atoms of fission fragments and muonic atoms of the shape isomer of 238 U. (author)

  16. Theoretical model application to the evaluation of fission neutron data up to 20 MeV incidence energy

    International Nuclear Information System (INIS)

    Ruben, A.; Maerten, H.; Seeliger, D.

    1990-01-01

    A complex statistical theory of fission neutron emission combined with a phenomenological fission model has been used to calculate fission neutron data for 238 U. Obtained neutron multiplicities and energy spectra as well as average fragment energies for incidence energies from threshold to 20 MeV (including multiple-chance fission) are compared with traditional data representations. (author). 19 refs, 6 figs

  17. What happens to the fission process above the 2nd- and 3rd-chance fission thresholds

    International Nuclear Information System (INIS)

    Stewart, L.; Howerton, R.J.

    1976-01-01

    Although the multiple fission process is important at high neutron energies, most of the evaluations available today do not include these individual fission cross sections or their associated fission spectra. The representations used in the Los Alamos and Livermore libraries are described and calculations compared with 14-MeV integral experiments available on 235 U, 238 U, and 239 Pu. Further work is needed to clearly delineate the specific problems in order to propose unique solutions

  18. Study of hypernuclei fission

    International Nuclear Information System (INIS)

    Malek, F.

    1990-01-01

    This work is about PS177 experience made on LEAR machine at CERN in 1988. The annihilation reaction of anti protons on a target of Bismuth or Uranium is studied. Lambda particles are produced by this reaction, in the nucleus in 2% of cases 7.1 10 -3 hypernuclei by stopped antiproton in the target are produced. The prompt hypernucleus fission probability of uranium is 75% and that of Bismuth 10%. The mass distribution of fission fragments is symmetrical ((≡ the excitation energy of the nucleus is very high). If the nucleus hasn't fissioned, the non-mesonic lambda decay, gives it an energy of 100 MeV, what allows to fission later. This fission is delayed because the hypernucleus lifetime is 1.3 +0.25 -0.21 10 -10 sec for Bismuth [fr

  19. Accidental behaviour of nuclear fuel in a warehousing site under air: investigation of the nuclear ceramic oxidation and of fission gas release; Comportement accidentel du combustible nucleaire dans un site d'entreposage sous air: Etude de l'oxydation de la ceramique nucleaire et du relachement des gaz de fission

    Energy Technology Data Exchange (ETDEWEB)

    Desgranges, L.

    2006-12-15

    After a brief presentation of the context of his works, i.e. the nuclear fuel, its behaviour in a nuclear reactor, and studies performed in high activity laboratory, the author more precisely presents its research topic: the behaviour of defective nuclear fuel in air. Then, he describes the researches performed in three main directions: firstly, the characterization and understanding of fission gas localisation (experimental localisation, understanding of the bubble forming mechanisms), secondly, the determination of mechanisms related to oxidation (atomic mechanisms related to UO{sub 2} oxidation, oxidation of fragments of irradiated fuel, the CROCODILE installation). He finally presents his scientific project which notably deals with fission gas release (from UO{sub 2} to U{sub 3}O{sub 7}, and from U{sub 3}O{sub 7} to U{sub 3}O{sub 8}), and with further high activity laboratory experiments

  20. The nuclear fission process

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Fifty years after its discovery, the nuclear fission phenomenon is of recurring interest. When its fundamental physics aspects are considered, fission is viewed in a very positive way, which is reflected in the great interest generated by the meetings and large conferences organized for the 50th anniversary of its discovery. From a purely scientific and practical point of view, a new book devoted to the (low energy) nuclear fission phenomenon was highly desirable considering the tremendous amount of new results obtained since the publication of the book Nuclear Fission by Vandenbosch and Huizenga in 1973 (Academic Press). These new results could be obtained thanks to the growth of technology, which enabled the construction of powerful new neutron sources, particle and heavy ion accelerators, and very performant data-acquisition and computer systems. The re-invention of the ionization chamber, the development of large fission fragment spectrometers and sophisticated multiparameter devices, and the production of exotic isotopes also contributed significantly to an improved understanding of nuclear fission. This book is written at a level to introduce graduate students to the exciting subject of nuclear fission. The very complete list of references following each chapter also makes the book very useful for scientists, especially nuclear physicists. The book has 12 chapters covering the fission barrier and the various processes leading to fission as well as the characteristics of the various fission reaction products. In order to guarantee adequate treatment of the very specialized research fields covered, several distinguished scientists actively involved in some of these fields were invited to contribute their expertise as authors or co-authors of the different chapters

  1. Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

    International Nuclear Information System (INIS)

    Nicoli, I.G.

    1981-06-01

    A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

  2. Measurement of cross-sections of fission reactions induced by neutrons on actinides from the thorium cycle at n-TOF facility; Mesures de sections efficaces de fission induite par neutrons sur des actinides du cycle du thorium a n-TOF

    Energy Technology Data Exchange (ETDEWEB)

    Ferrant, L

    2005-09-01

    In the frame of innovating energy source system studies, thorium fuel cycle reactors are considered. Neutron induced fission cross section on such cycle involved actinides play a role in scenario studies. To feed them, data bases are built with experimental results and nuclear models. For some nuclei, they are not complete or in disagreement. In order to complete these data bases, we have built an original set up, consisting in an alternation of PPACs (Parallel Plate Avalanche Chamber) and ultra - thin targets, which we installed on n-TOF facility. We describe detectors, set up, and the particular care brought to target making and characterization. Fission products in coincidence are detected with precise time measurement and localization with delay line read out method. We contributed, within the n-TOF collaboration, to the CERN brand new intense spallation neutron source characterization, based on time of flight measurement, and we describe its characteristics and performances. We were able to measure such actinide fission cross sections as {sup 232}Th, {sup 234}U, {sup 233}U, {sup 237}Np, {sup 209}Bi, and {sup nat}Pb relative to {sup 235}U et {sup 238}U standards, using an innovative acquisition system. We took advantage of the lame accessible energy field, from 0.7 eV to 1 GeV, combined with the excellent energy resolution in this field. Data treatment and analysis advancement are described to enlighten performance and limits of the obtained results. (author)

  3. Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

    International Nuclear Information System (INIS)

    2009-01-01

    This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

  4. Theory of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1998-01-01

    A survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity bar ν p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and some examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the recent models. In particular, the dependencies of N(E) and bar ν p upon the fissioning nucleus and its excitation energy are treated in detail for the Los Alamos model. Recent work in the calculation of the prompt fission neutron spectrum matrix N(E, E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of the ability to calculate these observables with confidence, the direction of future theoretical efforts, and limitations to current (and future) approaches. This paper is an extension of a similar paper presented at the International Centre for Theoretical Physics in 1996

  5. Calcul des propriétés élastiques des tissus utilisés dans les matériaux composites Computation of Elastic Properties of Fabrics Used in Composite Materials

    Directory of Open Access Journals (Sweden)

    Dal Maso F.

    2006-12-01

    Full Text Available Les renforts textiles s'imposent dès qu'il faut réaliser des structures massives ou complexes en matériaux composites, comme certains raccords et jonctions de tubes, des panneaux d'habitation légère, des carters de protection de têtes de puits en fond de mer, etc. Cet article expose les caractéristiques générales d'un renfort textile, puis différentes approches micromécaniques analytiques représentant les tissus à tissage bidimensionnel sont présentées. En partant du plus simple et en allant vers le plus complexe, ces modèles sont l'analogie à un stratifié [0°/90°], la mosaïque en série et en parallèle, les ondulations 1D et les ondulations 2D série-parallèle et parallèle-série. Toutes ces approches sont fondées sur la théorie mécanique des stratifiés. En analysant les résultats d'applications numériques de ces modèles et les résultats expérimentaux, on constate que les modèles des ondulations en 2D procurent les meilleures valeurs estimées des modules élastiques. Les autres modèles n'indiquent que des ordres de grandeurs. Woven fabric reinforcements are irreplaceable from the moment that heavy or complex composite structures should be manufactured, such as some pipe connections or flanges, light panels for housing, subsea well head protection panels, etc. In this paper overall characteristics of woven fabrics are described, followed by the review of different micromechanical analytical approaches. Starting with the simplest and continueing with the more complex, these models are : the analogy with a [0°/90°] laminate, the series and parallel mosaic models, the 1D fiber undulation model, and the 2D series-parallel and parallel-series fiber undulation models. All these approaches are based on the classical laminated plate theory. Analyzing the results of numerical applications of the models and experimental results, one can notice that both 2D fiber undulation models give the best estimated values for

  6. 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

  7. Effect of fission dynamics on the spectra and multiplicities of prompt fission neutrons

    International Nuclear Information System (INIS)

    Nix, J.R.; Madland, D.G.; Sierk, A.J.

    1985-01-01

    With the goal of examining their effect on the spectra and multiplicities of the prompt neutrons emitted in fission, we discuss recent advances in a unified macroscopic-microscopic description of large-amplitude collective nuclear dynamics. The conversion of collective energy into single-particle excitation energy is calculated for a new surface-plus-window dissipation mechanism. By solving the Hamilton equations of motion for initial conditions appropriate to fission, we obtain the average fission-fragment translational kinetic energy and excitation energy. The spectra and multiplicities of the emitted neutrons, which depend critically upon the average excitation energy, are then calculated on the basis of standard nuclear evaporation theory, taking into account the average motion of the fission fragments, the distribution of fission-fragment residual nuclear temperature, the energy dependence of the cross section for the inverse process of compound-nucleus formation, and the possibility of multiple-chance fission. Some illustrative comparisons of our calculations with experimental data are shown

  8. Compilation and evaluation of fission yield nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1991-09-01

    The task of this meeting was to review the progress made since the previous meeting on fission yield evaluation and to define the tasks for an IAEA Co-ordinated Research Programme in detail. Improvements have been noted in measured data, model calculations and the situation of fission yield evaluation. Tabs

  9. Neutron emission during acceleration of 252Cf fission fragments

    International Nuclear Information System (INIS)

    Batenkov, O.I.; Blinov, M.V.; Blinov, A.B.; Smirnov, S.N.

    1991-01-01

    We investigate neutron emission during acceleration of fission fragments in the process of spontaneous fission of 252 Cf. Experimental angular and energy distributions of neutrons are compared with the results of calculations of neutron evaporation during fragment acceleration. (author). 8 refs, 3 figs

  10. Metal cluster fission: jellium model and Molecular dynamics simulations

    DEFF Research Database (Denmark)

    Lyalin, Andrey G.; Obolensky, Oleg I.; Solov'yov, Ilia

    2004-01-01

    Fission of doubly charged sodium clusters is studied using the open-shell two-center deformed jellium model approximation and it ab initio molecular dynamic approach accounting for all electrons in the system. Results of calculations of fission reactions Na_10^2+ --> Na_7^+ + Na_3^+ and Na_18...

  11. Influence of nuclear dissipation on fission dynamics of the excited ...

    Indian Academy of Sciences (India)

    A stochastic approach to fission dynamics based on two-dimensional Langevin equations was applied to calculate the anisotropy of the fission fragments angular distribution and average pre-scission neutron multiplicities for the compound nucleus 248Cf formed in the $${16}$O+$^{232}$Th reactions. Postsaddle nuclear ...

  12. Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector

    International Nuclear Information System (INIS)

    Dufek, Jan; Holst, Gustaf

    2016-01-01

    Highlights: • Errors in the fission matrix eigenvector and fission source are correlated. • The error correlations depend on coarseness of the spatial mesh. • The error correlations are negligible when the mesh is very fine. - Abstract: Previous studies raised a question about the level of a possible correlation of errors in the cumulative Monte Carlo fission source and the fundamental-mode eigenvector of the fission matrix. A number of new methods tally the fission matrix during the actual Monte Carlo criticality calculation, and use its fundamental-mode eigenvector for various tasks. The methods assume the fission matrix eigenvector is a better representation of the fission source distribution than the actual Monte Carlo fission source, although the fission matrix and its eigenvectors do contain statistical and other errors. A recent study showed that the eigenvector could be used for an unbiased estimation of errors in the cumulative fission source if the errors in the eigenvector and the cumulative fission source were not correlated. Here we present new numerical study results that answer the question about the level of the possible error correlation. The results may be of importance to all methods that use the fission matrix. New numerical tests show that the error correlation is present at a level which strongly depends on properties of the spatial mesh used for tallying the fission matrix. The error correlation is relatively strong when the mesh is coarse, while the correlation weakens as the mesh gets finer. We suggest that the coarseness of the mesh is measured in terms of the value of the largest element in the tallied fission matrix as that way accounts for the mesh as well as system properties. In our test simulations, we observe only negligible error correlations when the value of the largest element in the fission matrix is about 0.1. Relatively strong error correlations appear when the value of the largest element in the fission matrix raises

  13. Fission gas in thoria

    Energy Technology Data Exchange (ETDEWEB)

    Kuganathan, Navaratnarajah, E-mail: n.kuganathan@imperial.ac.uk [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Ghosh, Partha S. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Galvin, Conor O.T. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Arya, Ashok K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Dutta, Bijon K. [Homi Bhabha National Institute, Trombay, Mumbai 400 094 (India); Dey, Gautam K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Grimes, Robin W. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom)

    2017-03-15

    The fission gases Xe and Kr, formed during normal reactor operation, are known to degrade fuel performance, particularly at high burn-up. Using first-principles density functional theory together with a dispersion correction (DFT + D), in ThO{sub 2} we calculate the energetics of neutral and charged point defects, the di-vacancy (DV), different neutral tri-vacancies (NTV), the charged tetravacancy (CTV) defect cluster geometries and their interaction with Xe and Kr. The most favourable incorporation point defect site for Xe or Kr in defective ThO{sub 2} is the fully charged thorium vacancy. The lowest energy NTV in larger supercells of ThO{sub 2} is NTV3, however, a single Xe atom is most stable when accommodated within a NTV1. The di-vacancy (DV) is a significantly less favoured incorporation site than the NTV1 but the CTV offers about the same incorporation energy. Incorporation of a second gas atom in a NTV is a high energy process and more unfavourable than accommodation within an existing Th vacancy. The bi-NTV (BNTV) cluster geometry studied will accommodate one or two gas atoms with low incorporation energies but the addition of a third gas atom incurs a high energy penalty. The tri-NTV cluster (TNTV) forms a larger space which accommodates three gas atoms but again there is a penalty to accommodate a fourth gas atom. By considering the energy to form the defect sites, solution energies were generated showing that in ThO{sub 2−x} the most favourable solution equilibrium site is the NTV1 while in ThO{sub 2} it is the DV. - Highlights: • We have considered Xe and Kr in point defects and defect clusters (neutral and charged) using Density Functional Theory (DFT) with a dispersion correction. • The most favourable charge state for a point defect (vacancy or interstitial) is that with full ionic charge and we have found that in all cases gas atoms occupy the fully charged vacancy sites. • The number of fission gas atoms accommodated in ThO{sub 2} is

  14. Session du Conseil du CERN : le ministre britannique, Robert Jackson, souligne l'intérêt de on pays pour l'avenir du CERN : décisions du Conseil pour la mise en oeuvre des recommandations du Comité d'évaluation du CERN: départ anticipé pour 200 membres au moins du personnel - mise à jour de la méthode de calcule pour les contributions des Etats Membres au budget

    CERN Multimedia

    CERN Press Office. Geneva

    1988-01-01

    Session du Conseil du CERN : le ministre britannique, Robert Jackson, souligne l'intérêt de on pays pour l'avenir du CERN : décisions du Conseil pour la mise en oeuvre des recommandations du Comité d'évaluation du CERN: départ anticipé pour 200 membres au moins du personnel - mise à jour de la méthode de calcule pour les contributions des Etats Membres au budget

  15. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    International Nuclear Information System (INIS)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-01-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  16. Calcul numérique des structures pour les Energies Marines Renouvelables: défis et structuration de la recherche autour du WeAMECenter

    OpenAIRE

    SCHOEFS , Franck

    2015-01-01

    International audience; Les structures offshore ont toujours constitué pour l'homme un défi logistique pour leur installation mais aussi en amont lors de leur conception. Apparues peu après la seconde guerre mondiale pour l'exploitation des hydrocarbures à quelques centaines de mètres des côtes, elles ont connues un essor important dans les années 70 et 80 suite aux deux chocs pétroliers. Le challenge depuis 15 ans est une exploitation à plusieurs milliers de mètres de profondeur. La recherch...

  17. Correlated prompt fission data in transport simulations

    Science.gov (United States)

    Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

    2018-01-01

    Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in

  18. Correlated prompt fission data in transport simulations

    Energy Technology Data Exchange (ETDEWEB)

    Talou, P.; Jaffke, P.; Kawano, T.; Stetcu, I. [Los Alamos National Laboratory, Nuclear Physics Group, Theoretical Division, Los Alamos, NM (United States); Vogt, R. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); University of California, Physics Department, Davis, CA (United States); Randrup, J. [Lawrence Berkeley National Laboratory, Nuclear Science Division, Berkeley, CA (United States); Rising, M.E.; Andrews, M.T.; Sood, A. [Los Alamos National Laboratory, Monte Carlo Methods, Codes, and Applications Group, Los Alamos, NM (United States); Pozzi, S.A.; Clarke, S.D.; Marcath, M.J. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI (United States); Verbeke, J.; Nakae, L. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); Jandel, M. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States); University of Massachusetts, Department of Physics and Applied Physics, Lowell, MA (United States); Meierbachtol, K. [Los Alamos National Laboratory, Nuclear Engineering and Nonproliferation, Los Alamos, NM (United States); Rusev, G.; Walker, C. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States)

    2018-01-15

    Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n-n, n-γ, and γ-γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX-PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation

  19. Fission tracks diameters in glasses

    International Nuclear Information System (INIS)

    Garzon Ruiperez, L.; Veiguela, J.

    1974-01-01

    Standard glass microscope slides have been irradiated with fission fragments from the uranium. The etching track conditions have been the same for the series, having changed the etching time only for each specimen. For each glass, a minimum of 250 measurements of the tracks diameters have been made, the distributions of which are the bimodal type. Diameters-etching dependence with time is roughly lineal. Energy determinations have been made with the help of the diameters-energy relations. The calculated values agree very well with the know ones. (author) [es

  20. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

  1. The Effect of Fission Products on the Radioactivity of some Hydrobionts of the Pacific Ocean; Effet des Produits de Fission sur la Radioactivite de Certains Organismes Marins de l'Ocean Pacifique; 0412 041b 0418 0414 ; Influencia de los Productos de Fision en la Radiactividad de Algunos Organismos Marinos del Oceano Pacifico

    Energy Technology Data Exchange (ETDEWEB)

    Moiseev, P. A.; Kardashev, A. V. [All-Union Marine Fisheries and Oceanographical Scientific Research Institute, Moscow, USSR (Russian Federation)

    1960-07-01

    The paper emphasizes the fact that currents in the water masses disperse radioactive elements over large distances, with the result that many marine fauna and flora far removed from the centres of contamination may accumulate dangerous concentrations of such elements. (author) [French] Dans ce memoire, on souligne que les masses d'eau des courants marins transportent les elements radioactifs a de grandes distances; il s'ensuit que de nombreux animaux et vegetaux marins vivant loin des foyers de contamination peuvent accumuler ces elements jusqu'a des concentrations dangereuses. (author) [Spanish] En el informe se pone de manifiesto que las corrientes de agua transportan elementos radiactivos a grandes distancias, por lo cual muchos animales y vegetales marinos, que habiten lejos de los focos de contaminacion, pueden acumular estos elementos en concentraciones peligrosas. (author) [Russian] V doklade podcherkivaetsja, chto vodnye massy raznosjat techenijami radioaktivnye jelementy na bol'shie rasstojanija, v rezul'tate chego mnogie morskie zhivotnye i rastenija, obitajushhie daleko ot ochagov zarazhenija, mogut nakaplivat' ztj jelementy do opasnyh koncentracij. (author)

  2. Calculation of the energy of particles emitted by the reactions {sup 3}{sub 1}T (d,n) {sup 4}{sub 2}He, D (d,n) {sup 3}{sub 2}He and D (d,p) T; Calcul de l'energie des particules emises par les reactions {sup 3}{sub 1}T (d,n) {sup 4}{sub 2}He, D (d,n) {sup 3}{sub 2}He et D (d,p) T

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    With a view to ease the work of research workers using 150 kV electrostatic accelerators, we have calculated the energy and the emission angle of particles emitted during the reactions {sub 1}{sup 3}T(d,n){sub 2}{sup 4}He, {sub 1}{sup 2}D(d,n){sub 2}{sup 3}He and {sub 1}{sup 2}D(d,p){sub 1}{sup 3}T. The results are classified in tables and arranged according to the acceleration energy of the deuterons. Since the energies considered are relatively low we have limited our study to the non-relativistic domain; this simplification results in a maximum energy variation with respect to the real energy values of 1 per cent. We give also two curves representing the variations in the total cross-sections for the reactions T (d,n){sub 2}{sup 4}He and D (d,n){sub 2}{sup 3}He. (authors) [French] De facon a faciliter la tache des experimentateurs utilisant des accelerateurs electrostatiques de 150 kV, nous avons calcule l'energie et l'angle d'emission des particules emises lors des reactions {sub 1}{sup 3}T(d,n){sub 2}{sup 4}He, {sub 1}{sup 2}D(d,n){sub 2}{sup 3}He and {sub 1}{sup 2}D(d,p){sub 1}{sup 3}T. Les resultats ont ete classes dans des tableaux, et ordonnes en fonction de l'energie d'acceleration des deuterons. Les energies considerees etant relativement peu elevees, nous avons limite notre etude au domaine non relativiste, cette simplification n'entraine qu'une variation maximale de 1 pour cent sur les valeurs reelles des energies. Nous avons joint a ce calcul deux courbes representant la variation des sections efficaces totales des reactions T (d,n){sub 2}{sup 4}He et D (d,n){sub 2}{sup 3}He. (auteurs)

  3. Microscopic Theory of Fission

    International Nuclear Information System (INIS)

    Younes, W; Gogny, D

    2008-01-01

    In recent years, the microscopic method has been applied to the notoriously difficult problem of nuclear fission with unprecedented success. In this paper, we discuss some of the achievements and promise of the microscopic method, as embodied in the Hartree-Fock method using the Gogny finite-range effective interaction, and beyond-mean-field extensions to the theory. The nascent program to describe induced fission observables using this approach at the Lawrence Livermore National Laboratory is presented

  4. Extraction of potential energy in charge asymmetry coordinate from experimental fission data

    Energy Technology Data Exchange (ETDEWEB)

    Pasca, H. [Joint Institute for Nuclear Research, Dubna (Russian Federation); ' ' Babes-Bolyai' ' Univ., Cluj-Napoca (Romania); Andreev, A.V.; Adamian, G.G. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Antonenko, N.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Tomsk Polytechnic Univ. (Russian Federation). Mathematical Physics Dept.

    2016-12-15

    For fissioning isotopes of Ra, Ac, Th, Pa, and U, the potential energies as a function of the charge asymmetry coordinate are extracted from the experimental charge distributions of the fission fragment and compared with the calculated scission-point driving potentials. The role of the potential energy surfaces in the description of the fission charge distribution is discussed. (orig.)

  5. Role of energy cost in the yield of cold ternary fission of Cf

    Indian Academy of Sciences (India)

    Abstract. The energy costs in the cold ternary fission of 252Cf for various light charged particle emission are calculated by including Wong's correction for Coulomb potential. Energy cost is found to be higher in cold fission than in normal fission. It is found that energy cost always increases with decrease in experimental yield ...

  6. Fission yield correlation generation and impact on nuclear problems - 15570

    International Nuclear Information System (INIS)

    Fiorito, L.; Stankovskiy, A.; Van den Eynde, G.

    2015-01-01

    In our work we defined a scheme to update fission yields and their covariance matrices. We implemented a Generalised Linear Least Square (GLLS) updating procedure to produce inter-isotope fission yield correlations. At each update, a constraining equation was selected and the related set of observables calculated using the prior knowledge of the fission yield data and uncertainties. Then, available extra information on each observable was introduced into the system - e.g. a data set of direct measurements or uncertainties. Our GLLS-based updating tool calculates best-estimate posterior fission yields and covariance matrices which merge both the extra and prior data. The major result of the update is the generation of fission yield correlations. We created complete updated covariance matrices for 6 nuclides (Th 232 , U 233 , U 235 , U 238 , Pu 239 and Pu 241 ) and a total of 14 fissioning systems using the JEFF-3.1.1 files. The fission yield covariance matrices were tested against the criticality and nuclide inventory calculations of the REBUS single pin benchmark after one irradiation cycle. It appears that fission yield correlations reduce the uncertainties to a very great extent, which in many cases are 4 times smaller than those obtained with uncorrelated data

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  8. Signatures of fission dynamics in highly excited nuclei produced in 197AU(800 A MeV) on proton collisions

    International Nuclear Information System (INIS)

    Benlliure, J.; Armbruster, P.; Bernas, M.

    2001-09-01

    197 Au(800 A MeV)-on-proton collisions are used to investigate the fission dynamics at high excitation energy. The kinematic properties together with the isotopic identification of the fission fragments allow to determine the mass, charge and excitation energy of the fissioning nucleus at saddle. The comparison of these observables and the measured total fission cross section with model calculations evidences a clear hindrance of fission at high excitation energy that can be explained in terms of nuclear dissipation. Assuming a statistical evaporation for other de-excitation channels than fission, an estimated value of the transient time of fission of (3 ± 1) . 10 -21 s is obtained. (orig.)

  9. Angular-momentum-bearing modes in fission

    International Nuclear Information System (INIS)

    Moretto, L.G.; Peaslee, G.F.; Wozniak, G.J.

    1989-03-01

    The angular-momentum-bearing degrees of freedom involved in the fission process are identified and their influence on experimental observables is discussed. The excitation of these modes is treated in the ''thermal'' limit, and the resulting distributions of observables are calculated. Experiments demonstrating the role of these modes are presented and discussed. 61 refs., 12 figs

  10. Neutron induced current pulses in fission chambers

    International Nuclear Information System (INIS)

    Taboas, A.L.; Buck, W.L.

    1978-01-01

    The mechanism of neutron induced current pulse generation in fission chambers is discussed. By application of the calculated detector transfer function to proposed detector current pulse shapes, and by comparison with actually observed detector output voltage pulses, a credible, semi-empirical, trapezoidal pulse shape of chamber current is obtained

  11. Characteristics of fission product release from a molten pool

    International Nuclear Information System (INIS)

    Yun, J.I.; Suh, K.Y.; Kang, C.S.

    2001-01-01

    The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

  12. Fission and r-process nucleosynthesis in neutron star mergers

    International Nuclear Information System (INIS)

    Giuliani, Samuel Andrea

    2018-01-01

    Fission plays a crucial role for the r-process nucleosynthesis in neutron star mergers. Due to the high neutron densities achieved in this astrophysical scenario the sequence of neutron captures and beta decays that constitutes the r process produces superheavy neutron rich nuclei that become unstable against fission. Fission determines thus the heaviest nuclei that can be produced by the r process and the fission yields shape the abundances of lighter nuclei. But despite the key role of fission the sensitivity of the r-process nucleosynthesis to uncertainties in fission predictions has not been explored. Nowadays there are only few set of fission rates suited for r-process calculations and most of them rely on a simplified treatment of the fission process. In this thesis we go beyond these approximations and compute the fission properties of r-process nuclei using the energy density functional approach. Fission is described as a tunneling process where the nucleus ''moves'' in a collective space characterized by coordinates describing the nuclear shape. Thus fission depends on the evolution of the energy with the deformation but also on the inertia due to the motion in the collective space. This is analogous to the quantum mechanical tunneling of a particle inside a potential well. In our study the relevant quantities for the description of the fission process are consistently computed for 3642 nuclei following the Hartree-Fock-Bogolyubov theory with constraining operators. We perform an extensive benchmark against the available experimental data and explore the variations of the fission properties along the superheavy landscape. We find that while collective inertias have a strong impact in the fission probabilities of light nuclei their role becomes less relevant in r -process nuclei. Within the statistical model we compute the neutron induced stellar reaction rates relevant for the r-process nucleosynthesis. These sets of stellar reaction

  13. A review of calculation methods for fast and intermediate reactors; Expose des methodes pour le calcul de reacteurs a neutrons rapides et intermediaires; Obzor metodov rascheta reaktorov na promezhutochnykh i bystrykh nejtronakh; Estudio panoramico de los metodos de calculo de los reactores rapidos e intermedios

    Energy Technology Data Exchange (ETDEWEB)

    Marchuk, G I [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

    1962-03-15

    This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author) [French] L'auteur examine la mise au point de methodes pour le calcul de reacteurs a neutrons rapides et intermediaires . Il decrit diverses manieres d'aborder les problemes des calculs sur la physique des reacteurs, notamment le calcul des effets de resonance. Il s'attache particulierement aux points suivants: systemes d'equations fondamentales et conjuguees a plusieurs groupes; diverses applications de la theorie des perturbations aux problemes de calculs sur la physique des reacteurs; methodes numeriques pour resoudre les equations fondamentales et conjuguees, voisines de la methode des harmoniques spheriques. L'auteur decrit ensuite une maniere d'appliquer la methode de la reponse aux problemes de la masse critique ainsi que des methodes pour le calcul de reacteurs ralentis a l'hydrogene. Il decrit les caracteristique s fondamentale s d'un modele de reacteur a un groupe effectif. (author) [Spanish] El autor analiza el desarrollo de los metodos de calculo de los reactores nucleares que trabajan con neutrones rapidos y con neutrones intermedios. Examina diversos planteos de los problemas del calculo fisico. Indica la forma de tomar en cuenta los efectos de resonancia y menciona los sistemas

  14. Systematics of neutron-induced fission yields

    International Nuclear Information System (INIS)

    Blachot, J.; Brissot, R.

    1983-10-01

    The main characteristics of the mass and charge distributions for thermal neutron induced fission of actinides are reviewed. We show that these distributions can be reasonably reproduced with only 24 data as input. We use a representation where the element yields together with the most probable mass Ap(Z) play the dominant role. The ability of this model to calculate mass yields for the fission of not yet measured actinides is also shown. The influence of the excitation energy of the fissile system on charge and mass distribution is also discussed

  15. Obsidian dating by fission track method

    International Nuclear Information System (INIS)

    Araya, A.M.O.

    1990-12-01

    The fission track method was employed to obtain the age of twelve obsidian sample from Ecuador. By using the plateau-age correction method, we obtained the true age of each sample and were able to identify four groups of ages in the studied area. Thereafter we studied the fading of fission tracks in two obsidian samples with different origins: Yanaurcu, Ecuador and Monte Arci, Italy. We constructed Arrhenius plots and calculated activation energies for both samples. The results from thermal annealing experiments were compared with theoretical curves obtained by integrating an equation proposed by Shukolyukov et al (1965). (author). 43 refs, 20 figs, 10 tabs

  16. Dynamics of process at the final stage of nuclear fission

    International Nuclear Information System (INIS)

    Koljari, I.G.; Mavlitov, N.D.

    2005-01-01

    Numerous experimental data show, that the final stage of nuclear fission near to a scission point plays an essential role at formation of characteristics of fission products. At the description of a final stage of fission there is a number of problems: Definition of the form of the nuclear near the scission point and definition forms of a fission fragments; The account of dynamic processes in compound nuclear directly before of fission. The condition of the quasistatic al adiabatic process - dS/dt=0 - is applied in a point of transition from the uniform compound nuclei to several forms for definition of generalized coordinates and speeds. Calculation of dependence of post neutrons from nuclear mass of fission fragments for reactions is α+ 83 Bi 209 → 85 At 213 (E lab = 45 MeV); α+ 92 U 242 → 94 Pu 242 (E lab = 45 MeV); 8 O 18 + 79 Au 197 → 97 Fr 215 (E lab = 159 MeV). System of equations, which describes behaviour of system in a point of nuclear fission-transition from the uniform form to system of a two (and, probably more) fission fragments is given. The system of the equations allows in a fission point to define the generalized coordinates, and the generalized speeds for each of the generalized coordinates of collective deformation variables

  17. Potentials of fissioning plasmas

    International Nuclear Information System (INIS)

    Karlheinz, Thom.

    1979-01-01

    Successful experiments with the nuclear pumping of lasers have demonstrated that in gaseous medium the kinetic energy of fission fragments can be converted directly into non-equilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in nearly thermal equilibrium varying up to a state of extreme-non-equilibrium. The accompanying variations of temperatures, pressure and radiative spectrum suggest wide ranges of applications. For example, in the gas-phase fission reactor concept enriched uranium hexafluoride or an uranium plasma replaces conventional fuel elements and permits operation above the melting point of solid materials. This potential has been motivation for the US National Aeronautics and Space Administration (NASA) to conduct relevant research for high specific impulse propulsion in space. The need to separate the high temperature gaseous fuel from the surfaces of a containing vessel and to protect them against thermal radiation has led to the concept of an externally moderated reactor in which the fissioning gaseous material is suspended by fluid dynamic means and the flow of opaque buffer gas removes the power. The gaseous nuclear fuel can slowly be circulated through the reactor for continuous on-site reprocessing including the annihilation of transuranium actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides can thus be achieved. These characteristics and the unique radiative properties led to the expectation that the gas-phase fission reactor could feature improved safety, safeguarding and economy, in addition to new technologies such as processing, photochemistry and the transmission of power over large distances in space

  18. De la théorie au modèle : les hélices comme sculptures calculées, le cas des Fonderies de l’Atlantique à Nantes

    Directory of Open Access Journals (Sweden)

    Jean-Louis Kerouanton

    2009-05-01

    Full Text Available La découverte, l’étude et la sauvegarde partielle des bâtiments des anciennes « Fonderies de l’Atlantique » à Nantes en 2001-2002 ont permis également de révéler puis de conserver une extraordinaire collection d’objets de fonderie au sable correspondant à la fabrication des hélices et des pales d’hélice pour les plus gros navires commerciaux et militaires. La technologie de fonderie d’alliage de cuivre pour les hélices avait été adoptée en 1937 dans l’entreprise. L’ensemble des modèles en bois de pales d’hélice est d’abord apparu comme une accumulation de sculptures monumentales particulièrement épurées. L’étude technologique plus fine montre cependant que cette « beauté » réelle ne correspond nullement à la volonté d’un artiste mais relève uniquement des calculs liés pour chaque production à un projet de navire. Et c’est seulement le calcul qui permet d’expliquer les réalisations d’hélices monoblocs par troussage, sans modèle aucun ; il ne reste alors de traces que les outils du troussage. Il s’agit bien dès lors d’un patrimoine scientifique - la transposition matérielle d’un calcul mathématique - tout autant que d’un patrimoine technique et industriel. C’est cette particularité qui fait la grande originalité de l’ensemble conservé par la communauté urbaine, en attendant, nous l’espérons, une valorisation future.In 2001 and 2002, the study and partial preservation of the buildings of the former Fonderies de l’Atlantic in Nantes also allowed for the preservation of a remarkable collection of objects related to the sand foundry casting carried out by this firm, specialised in propellers for major commercial and military vessels. The firm began to employ copper alloys in its foundry techniques from 1937. The collection of wooden models prepared for the casting of propellers appears, at first sight, as a remarkable accumulation of rather abstract monumental

  19. Méthodes de type éléments finis pour le calcul des champs électriques et magnétiques en électroencéphalographie et magnétoencéphalographie

    Science.gov (United States)

    Guérin, Christophe; Marin, Gildas; Garnero, Line; Meunier, Gérard

    1997-12-01

    So as to compute the electric potential and the flux density generated by the electrical activity of the brain, numerical methods based on the Finite Element Method have been developed. These methods, which can treat realistic head models and can take into account anisotropy of conductivity, for instance in the skull, are presented. Then two numerical examples are described: a spherical model and a realistic head model. Afin de calculer le potentiel électrique et l'induction créés par l'activité électrique du cerveau, nous avons développé des méthodes utilisant la Méthode des Éléments Finis. Ces méthodes, qui peuvent s'appliquer à des modèles réalistes de tête et qui permettent de tenir compte de la conductivité anisotrope de certains tissus comme l'os, sont présentées. Puis deux exemples numériques sont décrits : un modèle de sphères concentriques et un modèle réaliste de tête.

  20. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  1. Nuclear data in the problem of fission reactor decommissioning

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Kulagin, N.T.

    1993-01-01

    This report presents a review of the works published in Russia during last several years and devoted to the problem of nuclear data and calculations of nuclear facilities activation for fission reactor decommissioning. 6 refs

  2. Lunar surface fission power supplies: Radiation issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Lee, S.K.

    1994-01-01

    A lunar space fission power supply shield that uses a combination of lunar regolith and materials brought from earth may be optimal for early lunar outposts and bases. This type of shield can be designed such that the fission power supply does not have to be moved from its landing configuration, minimizing handling and required equipment on the lunar surface. Mechanisms for removing heat from the lunar regolith are built into the shield, and can be tested on earth. Regolith activation is greatly reduced compared with a shield that uses only regolith, and it is possible to keep the thermal conditions of the fission power supply close to these seen in free space. For a well designed shield, the additional mass required to be brought fro earth should be less than 1000 kg. Detailed radiation transport calculations confirm the feasibility of such a shield

  3. Lunar surface fission power supplies: Radiation issues

    International Nuclear Information System (INIS)

    Houts, M.G.; Lee, S.K.

    1994-01-01

    A lunar space fission power supply shield that uses a combination of lunar regolith and materials brought from earth may be optimal for early lunar outposts and bases. This type of shield can be designed such that the fission power supply does not have to be moved from its landing configuration, minimizing handling and required equipment on the lunar surface. Mechanisms for removing heat from the lunar regolith are built into the shield, and can be tested on earth. Regolith activation is greatly reduced compared with a shield that uses only regolith, and it is possible to keep the thermal conditions of the fission power supply close to those seen in free space. For a well designed shield, the additional mass required to be brought from earth should be less than 1,000 kg. Detailed radiation transport calculations confirm the feasibility of such a shield

  4. Processus of fission at medium energy

    International Nuclear Information System (INIS)

    Delagrange, H.

    1977-01-01

    Excitation functions for the reactions 233 U, 234 U, 235 U( 4 He,xn)Pu have been measured radiochimically between 20 and 45 MeV. Neutron emission probabilities have been deduced from these measurements, for 239 Pu, 238 Pu. These results have been analysed in the framework of a statistical model; the double humped fission barrier has been taken into account explicitly and the parameters of the barrier extracted. In the case of heavy ion reactions, fission probabilities have been calculated with the help of the Bohr-Wheeler formula and compared to experimental data from other authors. Deduced fission parameters (asub(f)/asub(n), Bsub(f)) are compared to different theoretical models [fr

  5. Fusion-Fission process and gamma spectroscopy of binary products in light heavy ion collisions (40 {<=} A{sub CN} {<=} 60); Processus de fusion-fission et spectroscopie gamma des produits binaires dans les collisions entre ions lourds legers (40 {<=} A{sub NC} {<=} 60)

    Energy Technology Data Exchange (ETDEWEB)

    Nouicer, Rachid [Institut de Recherche Subatomique, CNRS-IN2P3 - Universite Louis Pasteur, 67 - Strasbourg (France)

    1997-11-21

    During the work on which this Thesis is based, the significant role of the Fusion-Fission Asymmetric mechanism in light heavy ion collisions (A{sub NC} {<=} 60) has been emphasized. The Spin Dis-alignment in the oblate-oblate system has supplied evidence for the first time for the Butterfly mode in a resonant-like reaction. These two aspects, one macroscopic and the other more closely related to microscopic effects are certainly different from a conceptual point of view but are quite complementary for a global understanding of dinuclear systems. In the first part, inclusive and exclusive measurements of the {sup 35}Cl + {sup 12}C and {sup 35}Cl + {sup 24}Mg reaction have been performed at 8 MeV/nucleon in the Saclay experiment. These measurements have permitted us to verify the origin of products which have given rise of the asymmetric fusion-fission mechanism and which have demonstrated that the three-body process in this energy range is very weak. In the second part the {sup 28}Si + {sup 28}Si reaction has been performed at the resonance energy E{sub lab}> = 111.6 MeV at Strasbourg with the Eurogam phase II multi-detector array and VIVITRON accelerator. An angular momentum J{sup {pi}} 38{sup +} for inelastic and mutual channels of the {sup 28}Si + {sup 28}Si exit channel has been measured and has supplied evidence for a spin dis-alignment which has been interpreted in the framework of a molecular model by Butterfly motion. The spectroscopic study of {sup 32}S nucleus, has revealed the occurrence of a new {gamma}-ray transition 0{sup +}(8507.8 keV) {yields} 2{sub 1}{sup +}(2230.2 keV). (author) 105 refs., 116 figs., 26 tabs.

  6. Fundamentals of the double-humped fission barrier

    International Nuclear Information System (INIS)

    Brack, M.

    1980-01-01

    We review the development of the theory of the fission barrier over the past forty years. Special emphasis is put on the shell-correction method of Strutinsky and its foundation and numerical verification from microscopical Hartree-Fock calculations. The different practical realisations of the method and its applications to the calculation of deformation energy surfaces are reviewed. The influence of the different shape degrees of freedom of the nucleus on the form of the fission barrier is discussed. Finally, we summarize some more recent developments concerning both experimental and theoretical aspects of the double-humped fission barrier. (author)

  7. Determination of fission gas yields from isotope ratios

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg

    1983-01-01

    This paper describes a method of calculating the actual fission yield of Kr and Xe in nuclear fuel including the effect of neutron capture reactions and decay. The bases for this calculation are the cumulative yields (ref. 1) of Kr and Xe isotopes (or pairs of isotopes) which are unaffected...... by neutron capture reactions, and measured Kr and Xe isotope ratios. Also the burnup contribution from the different fissile heavy isotopes must be known in order to get accurate fission gas yields....

  8. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer

  9. Extraction liquide-liquide des sels minéraux. Approximation analytique des isothermes de partage. Application au calcul sur ordinateur d'une installation Liquid-Liquid Extraction of Mineral Salts. Analytical Approximation of Partition Isotherms. Application to the Computer Designing of an Installation

    Directory of Open Access Journals (Sweden)

    Chaid M. T.

    2006-11-01

    Full Text Available Cet article comporte deux parties -dans la première on a calculé les isothermes de partage en partant des considérations thermodynamiques basées sur les travaux de Guggenheim. Les résultats ne sont bons que dans le cas des solutions idéales. On a donc simulé les isothermes par des fonctions analytiques. Celles-ci permettent de représenter convenablement tous les faisceaux d'isothermes quelles que soient l'acidité de la solution aqueuse et les concentrations dans les deux phases; - la deuxième partie est réservée au calcul d'une installation complète d'extraction avec colonne de récupération; le calcul, mené sur ordinateur, utilise les fonctions analytiques précédentes et permet d'obtenir les débits et les concentrations à chaque plateau. This article contains two parts: - the first part discusses the calculating of partition isotherms using thermodynamic considerations based on research by Guggenheim. The results are good only in the case of ideal solutions. Therefore, isotherms were simulated by ànalyticfunctionswhich are suitable for representing all isotherm bundles no matter what the acidity of the aqueous solution and the concentrations in both phases moy be. - the second part 'deals' with designing a complete extraction installation with a recovery column. Computerized designing makes use of the preceding analytic functions and enables flowrates and concentrations ta be determined at each plate.

  10. True ternary fission of 252Cf

    International Nuclear Information System (INIS)

    Vijayaraghavan, K.R.; Balasubramaniam, M.; Oertzen, W. von

    2014-01-01

    Splitting of heavy radioactive nucleus into three fragments is known as ternary fission. If the size of the fragments are almost equal it is referred to as true ternary fission. Recently, Yu. V. Pyatkov et al observed/reported the experimental observation of true ternary fission in 252 Cf. In this work, the possibilities of different true ternary fission modes of 252 Cf through potential energy surface (PES) calculations based on three cluster model (TCM) are discussed. In TCM a condition on the mass numbers of the fission fragments is implied as A 1 ≥ A 2 ≥ A 3 in order to avoid repetition of combinations. Due to this condition, the values of Z 3 vary from 0 to 36 and Z 2 vary from 16 to 51. Of the different pairs having similar (Z 2 , Z 3 ) with different potential energy, a pair possessing minimum potential energy is chosen. Thus identified favourable combinations are plotted. For the PES calculations the arrangement of the fragments is considered in the order of A 1 +A 2 +A 3 . i.e. the heavy and the lightest fragments are kept at the ends. It is seen that the deepest minimum in the PES occurs for Z 3 =2 labelled as (Z 2 ; 2) indicating He accompanied breakup as the most favourable one. Of which, the breakup with Z 2 around 46 to 48 is the least (shown by dashed (Z 1 = 50) and dotted (Z 1 = 52) lines indicating a constant Z 1 value). The other notable minima in the PES are labelled and they correspond to the (Z 2 , Z 3 ) pairs viz., (20, 20), (28, 20), (28, 28) and (32, 32). Of these four minima, the first three are associated with the magic numbers 20 and 28. For Z 3 =20, there are two minimums at (20,20) and (28,20) among them (28,20) is the lowest minimum through which the minimum-path passes, and it is the ternary decay observed by Yu. V. Pyatkov et al. The fourth minima is the most interesting due to the fact that it corresponds to true ternary fission mode with Z 2 =32, Z 3 =32 and Z 1 =34. The minimum potential energy path also goes through this true

  11. Methodology and application of the WIMS-D4M fission product data

    International Nuclear Information System (INIS)

    Mo, S.C.

    1995-01-01

    The WIMS-D4 code has been modified (WIMS-D4m) to generate burn-up dependent microscopic cross sections for use in full core depletion calculations. The calculation of neutron absorption by fission products can be obtained from a reduced fission-product-chain model that includes the 135 Xe and 149 Sm chains, and a lumped fission product to account for the absorption by fission products not explicitly treated. Burn-up calculations were performed for the ANS MEU core using WIMS and EPRI-CELL cross sections. The calculated eigenvalues and material loadings are in good agreements

  12. Fission track dating and estimation of uranium in some garnets of Rajasthan (India)

    Energy Technology Data Exchange (ETDEWEB)

    Singh, S; Virk, H S [Punjabi Univ., Patiala (India). Dept. of Physics

    1978-09-01

    The experimental procedure, involving the preparation, etching, thermal neutron irradiation and scanning of the garnet samples, is described. The calculated fission track ages and uranium concentration are tabulated.

  13. Specific fission J-window and angular momentum dependence of the fission barrier

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Hiroshi; Saito, Tadashi; Takahashi, Naruto; Yokoyama, Akihiko [Osaka Univ., Suita (Japan); Shinohara, Atsushi

    1997-04-01

    A method to determine a unique J-window in the fission process was devised and the fissioning nuclide associated with thus extracted J-window was identified for each of the heavy-ion reaction systems. Obtained fission barriers at the resulting J-window were compared with the calculated values by the rotating finite range model (RFRM). The deduced barriers for individual nuclides were compared with the RFRM barriers to reproduce more or less the angular momentum dependence the RFRM prediction. The deduced systematic behavior of the fission barrier indicates no even-odd and shell corrections are necessary. The nuclear dissipation effect based on Kramer`s model revealed substantial reduction of the statistically deduced barrier heights and brought a fairly large scattering from the RFRM J-dependence. However, introduction of the temperature-dependent friction coefficient ({gamma} = 2 for T {>=} 1.0 MeV and 0.5 for T < 1.0 MeV) was found to bring about satisfactory agreement with both RFRM fission barriers and the pre-fission neutron multiplicity systematics. (author). 81 refs.

  14. Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

    Science.gov (United States)

    Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

    2018-01-01

    Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using

  15. The evaluation for reference fission yield of 238U fission

    International Nuclear Information System (INIS)

    Liang Qichang; Liu Tingjin

    1998-01-01

    In the fission yield data evaluation and measurement, the reference yield is very important, good or poor recommended or measurement values depend upon the reference data to a great extent. According to the CRP's requirement, the evaluation of reference fission yields have been and will be carried out in CNDC, as a part of the whole work (contract No.9504/R 0 /Regular Budget Fund), the evaluation for 29 reference fission yields of 15 product nuclides from 238 U fission have been completed

  16. Indirect Determination of Half-Lives of Short-Lived Fission Gases Using a Gas Flow Method; Determination Indirecte des Courtes PeRiodes de Produits de Fission Gazeux, par une Methode a Courant Gazeux; 041a 041e 0421 0412 0415 041d 041d 041e 0415 041e 041f 0420 0415 0414 0415 041b 0415 041d 0418 0415 041f 0415 0420 0418 041e 0414 041e 0412 041f 041e 041b 0423 0420 0410 0421 041f 0410 0414 0410 041a 041e 0420 041e 0422 041a 041e 0416 0418 0412 0423 0429 0418 0425 0413 0410 0417 041e - 041e 0411 0420 0410 0417 041d 042b 0425 041f 0420 041e 0414 0423 041a 0422 041e 0412 0414 0415 041b 0415 041d 0418 042f 0421 0418 0421 041f 041e 041b 042c 0417 041e 0412 0410 041d 0418 0415 041c 041c 0415 0422 041e 0414 041e 0412 0413 0410 0417 041e 0412 041e 0413 041e 041f 041e 0422 041e 041a 0410 ; Determinacion Indirecta del T 1/2 de los Cases de Fision de Periodo Corto por un Metodo de Circulacion

    Energy Technology Data Exchange (ETDEWEB)

    Patzelt, P.; Herrmann, G. [Institut fuer Anorganische Chemie und Kernchemie der Universitaet Mainz, Mainz, Federal Republic of Germany (Germany)

    1965-07-15

    ) [French] Etant donne que l'etude de la repartition des charges dans la fission nucleaire exige que l'un determine les rendements primaires en produits de fission qui, dans la plupart des cas, sont eux-memes des produits a courte periode ou proviennent de precurseurs a courte periode, il faut proceder a des separations tres rapides et efficaces de membres voisins d'une serie radioactive. Si l'on utilise le pouvoir d'emanation extremement eleve de certains composes de l'uranium, on separe le plus facilement les produits de fission gazeux dans plus d'une douzaine de series radioactives. Cependant, les calculs de rendement, de meme que le pouvoir d'emanation effectif des cibles en couches epaisses, peuvent etre sensiblement influences par les donnees relatives a la periode des gaz. C'est pourquoi on a determine avec une grande precision les periodes des isotopes du krypton et du xenon, de {sup 89}Kr a {sup 93}Kr et de {sup 131}Xe a {sup 143}Xe, par une methode a courant gazeux et un procede perfectionne avec electrode chargee. Les auteurs ont effectue des mesures par l'intermediaire des produits de desintegration a longue periode deposes sur les electrodes. Le debit, le temps d'irradiation, le temps de refroidissement et les dispositions relatives au comptage on ete choisis de maniere que les mesures des divers produits de filiation soient grandement facilitees. Dans certains cas, des separations radiochimiques se sont revelees indispensables pour des raisons de selectivite. Les resultats obtenus sont les suivants: {sup 89}Kr, 3.11 {+-} 0.06 min: {sup 90}Kr, 32.1 {+-} 0.8 s; {sup 91}Kr, 8.36 {+-} 0.15 s; {sup 92}Kr, 1,92 {+-} 0.07 s; {sup 93}Kr, 1.17 {+-} 0.04 s; {sup 137}Xe, 3.87 {+-} 0.15 min; {sup 138}Xe, 14.5 {+-} 0.5 min; {sup 139}Xe, 40.4 {+-} 1.3 s: {sup 140}Xe, 13.33 {+-} 0.27s; {sup 141}Xe, 1.70 {+-} 0.05s; {sup 142}Xe, 1.15 {+-} 0.04s- et {sup 143}Xe, 0.96 {+-} 0.02s. Les auteurs ont obtenu les periodes en extrapolant les courbes de desintegration beta ou de

  17. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

  18. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1982-02-01

    In these lectures we present the liquid drop model of fission and compare some of its prediction with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. We then discuss, using the example of the oscillator model, the generality of shell effects. We show how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  19. Fission of heavy hypernuclei

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1993-01-01

    The results on delayed and prompt fission of heavy hypernuclei obtained by the LEAR PS177 collaboration are recalled and discussed. It is shown that the hypernuclei life-times can be explained in term of a weak strangeness violating lambda-nucleon interaction with a cross section close to 6.0 10 -15 barns. The lambda attachment function is shown to be sensitive to the scission configuration, just before fission, and to the neck dynamics. This function provides a new way to study the nuclear scission process. (author)

  20. Fission gas measuring technology

    International Nuclear Information System (INIS)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok.

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs

  1. Fission gas measuring technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs.

  2. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1980-08-01

    In these lectures the liquid drop model of fission is presented and some of its predictions compared with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. It is shown how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  3. Actinide neutron-induced fission up to 20 MeV

    International Nuclear Information System (INIS)

    Maslov, V.M.

    2001-01-01

    Fission and total level densities modelling along with double-humped fission barrier parameters allow to describe available actinide neutron-induced fission cross section data in the incident neutron energy range of ∼ 10 keV - 20 MeV. Saddle asymmetries relevant to shell correction model calculations influence fission barriers, extracted by cross section data analysis. The inner barrier was assumed axially symmetric in case of U, Np and Pu neutron-deficient nuclei. It is shown that observed irregularities in neutron-induced fission cross section data in the energy range of 0.5-3 MeV could be attributed to the interplay of few-quasiparticle excitations in the level density of fissioning and residual nuclei. Estimates of first-chance fission cross section and secondary neutron spectrum model were validated by 238 U fission, (n,2n) and (n,3n) data description up to 20 MeV. (author)

  4. Actinide neutron-induced fission up to 20 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, V M [Radiation Physics and Chemistry Problems Institute, Minsk-Sosny (Belarus)

    2001-12-15

    Fission and total level densities modelling along with double-humped fission barrier parameters allow to describe available actinide neutron-induced fission cross section data in the incident neutron energy range of {approx} 10 keV - 20 MeV. Saddle asymmetries relevant to shell correction model calculations influence fission barriers, extracted by cross section data analysis. The inner barrier was assumed axially symmetric in case of U, Np and Pu neutron-deficient nuclei. It is shown that observed irregularities in neutron-induced fission cross section data in the energy range of 0.5-3 MeV could be attributed to the interplay of few-quasiparticle excitations in the level density of fissioning and residual nuclei. Estimates of first-chance fission cross section and secondary neutron spectrum model were validated by {sup 238}U fission, (n,2n) and (n,3n) data description up to 20 MeV. (author)

  5. Comment calculer le contenu énergétique des produits d'origine pétrolière et de leurs substituts d'origine charbonnière ou végétale How to Calculate the Energy Content of Products of Petroleum Origin and of Their Substitutes from Coal Or Biomass

    Directory of Open Access Journals (Sweden)

    Leprince P.

    2006-11-01

    Full Text Available Le contenu énergétique des produits pétroliers issus du raffinage, des produits de vapocraquage de naphta et de quelques produits de la pétrochimie, a été calculé en prenant en considération, la dépense énergétique entre coproduits ou sous-produits d'une même transformation, les variantes de procédés et l'effet des divers modes de fourniture d'énergie nécessaire à la transformation considérée. The energy content of petroleum products coming from refining, of naphta steam-cracking products and of various petrochemical products is calculated by taking into consideration the energy expenditure between the coproducts or by products from the same transformation, the different processes and the effect of different ways in which to supply the energy required for the transformation being considered.

  6. Fission fragment mass and total kinetic energy distributions of spontaneously fissioning plutonium isotopes

    Science.gov (United States)

    Pomorski, K.; Nerlo-Pomorska, B.; Bartel, J.; Schmitt, C.

    2018-03-01

    The fission-fragment mass and total kinetic energy (TKE) distributions are evaluated in a quantum mechanical framework using elongation, mass asymmetry, neck degree of freedom as the relevant collective parameters in the Fourier shape parametrization recently developed by us. The potential energy surfaces (PES) are calculated within the macroscopic-microscopic model based on the Lublin-Strasbourg Drop (LSD), the Yukawa-folded (YF) single-particle potential and a monopole pairing force. The PES are presented and analysed in detail for even-even Plutonium isotopes with A = 236-246. They reveal deep asymmetric valleys. The fission-fragment mass and TKE distributions are obtained from the ground state of a collective Hamiltonian computed within the Born-Oppenheimer approximation, in the WKB approach by introducing a neck-dependent fission probability. The calculated mass and total kinetic energy distributions are found in good agreement with the data.

  7. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  8. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  9. Electronic structure of the actinides and their dioxides. Application to the defect formation energy and krypton solubility in uranium dioxide; Etude de la structure electronique des actinides et de leurs dioxydes. Application aux defauts ponctuels et aux gaz de fission dans le dioxyde d`uranium

    Energy Technology Data Exchange (ETDEWEB)

    Petit, T. [CEA Centre d`Etudes Nucleaires de Grenoble, 38 (France)]|[CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique

    1996-09-28

    Uranium dioxide is the standard nuclear fuel used in French h power plants. During irradiation, fission products such as krypton and xenon are created inside fuel pellets. So, gas release could become, at very high burnup, a limiting factor in the reactor exploitation. To study this subject, we have realised calculations using the Density Functional Theory (DFT) into the Local Density Approximation (LDA) and the Atomic Sphere Approximation (ASA). First, we have validated our approach by calculating cohesive properties of thorium, protactinium and uranium metals. The good agreement between our results and experimental values implies that 5f electrons are itinerant. Calculated lattice parameter, cohesive energy and bulk modulus for uranium and thorium dioxides are in very good agreement with experiment. We show that binding between uranium and oxygen atoms is not completely ionic but partially covalent. The question of the electrical conductivity still remains an open problem. We have been able to calculate punctual defect formation energies in uranium dioxide. Accordingly to experimental observations, we find that it is easier to create a defect in the oxygen sublattice than in the uranium sublattice. Finally, we have been able to predict a probable site of krypton atoms in nuclear fuel: the Schottky trio. Experiences of Extended X-ray Absorption Fine structure Spectroscopy (EXAFS) and X-ray Photoelectron Spectroscopy (XPS) on uranium dioxide doped by ionic implantation will help us in the comprehension of the studied phenomena and the interpretation of our calculations. (author). 256 refs.

  10. Volatilization and reaction of fission products in flowing steam

    International Nuclear Information System (INIS)

    Johnson, I.; Steidl, D.V.; Johnson, C.E.

    1985-01-01

    The principal risk to the public from nuclear power plants derives from the highly radioactive atoms (fission products) generated as energy is produced in the nuclear fuel. The revolatilization of fission products from reactor system surfaces due to self-heating by radioactive decay has become a complicating factor in the source-term redefinition effort. It has had a major impact on calculations of fission product distributions in accident safety analyses. The focus of this research effort was to investigate the volatilization and transport of fission products and control rod materials in a flowing gaseous steam-hydrogen mixture. Fission product and control rod materials in various combinations were studied including CsI, CsOH, TeO 2 , SrO, Ag, In, Cd and Mn. The vaporization behavior of the deposits were characterized with respect to vaporization rates, chemical species and downstream transport behavior

  11. Modelling the widths of fission observables in GEF

    Directory of Open Access Journals (Sweden)

    Schmidt K.-H.

    2013-03-01

    Full Text Available The widths of the mass distributions of the different fission channels are traced back to the probability distributions of the corresponding quantum oscillators that are coupled to the heat bath, which is formed by the intrinsic degrees of freedom of the fissioning system under the influence of pairing correlations and shell effects. Following conclusion from stochastic calculations of Adeev and Pashkevich, an early freezing due to dynamical effects is assumed. It is shown that the mass width of the fission channels in low-energy fission is strongly influenced by the zero-point motion of the corresponding quantum oscillator. The observed variation of the mass widths of the asymmetric fission channels with excitation energy is attributed to the energy-dependent properties of the heat bath and not to the population of excited states of the corresponding quantum oscillator.

  12. Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method; Qualification des methodes de calculs de fluence dans les reacteurs a eau pressurisee. Amelioration du traitement des sections efficaces par la methode des tables de probabilite

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S H

    1994-01-01

    It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ``benchmarks`` have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section`s fluctuation. In this thesis, we propose a new method called ``Probability Table Method`` to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup`s results and probability table`s results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs.

  13. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5 wt. % Fs) are presented. (The symbol 'Fs' designates fissium, a 'pseudo-element' which, in reality, is an alloy whose composition is representative of fission products that remain in reprocessed fuel). The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations with the measurements shows quantitative agreement in both the magnitude and the axial variation of the retained gas content. (orig.)

  14. Process for treating fission waste

    International Nuclear Information System (INIS)

    Rohrmann, C.A.; Wick, O.J.

    1983-01-01

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste

  15. 50 years of nuclear fission

    International Nuclear Information System (INIS)

    Hilscher, D.

    1989-01-01

    The article tells the story of the discovery of nuclear fission in Berlin 50 years ago by Otto Hahn and Fritz Strassmann in cooperation with Lise Meitner. 50 years later nuclear fission is still a subject of research. Some question remain unanswered. Selected new research results are used to discuss the dynamics of the collective movement of the elementary nuclear fission process. (orig.) [de

  16. Fission dynamics of hot nuclei

    Indian Academy of Sciences (India)

    2014-04-05

    Apr 5, 2014 ... across the fission barrier is very small or in other words, the fission barrier is much ... of this shape evolution, the gross features of the fissioning nucleus can be described ..... [7] Y Abe, C Gregoire and H Delagrange, J. Phys.

  17. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Maeck, W.J.

    1979-01-01

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  18. The discovery of fission

    International Nuclear Information System (INIS)

    McKay, H.A.C.

    1978-01-01

    In this article by the retired head of the Separation Processes Group of the Chemistry Division, Atomic Energy Research Establishment, Harwell, U.K., the author recalls what he terms 'an exciting drama, the unravelling of the nature of the atomic nucleus' in the years before the Second World War, including the discovery of fission. 12 references. (author)

  19. A project for storing the radioactive sludges from Marcoule in the soil (1960); Une solution de stockage dans le sol des boues radioactives de Marcoule (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, P; Gailledreau, C [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    A study of the feasibility of storing radioactive sludges in the soil. Special attention is paid to the case of the Marcoule plant, producing three kinds of sludges. The movement of the different fission products is given. (author) [French] Etude de la possibilite de stocker dans le sol des boues radioactives. Le cas de la station de traitement des effluents de Marcoule, produisant trois types de boues, est etudie plus specialement. Le mouvement des divers produits de fission est precise. (auteur)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-01-15

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

  1. Elastocapillary Instability in Mitochondrial Fission

    Science.gov (United States)

    Gonzalez-Rodriguez, David; Sart, Sébastien; Babataheri, Avin; Tareste, David; Barakat, Abdul I.; Clanet, Christophe; Husson, Julien

    2015-08-01

    Mitochondria are dynamic cell organelles that constantly undergo fission and fusion events. These dynamical processes, which tightly regulate mitochondrial morphology, are essential for cell physiology. Here we propose an elastocapillary mechanical instability as a mechanism for mitochondrial fission. We experimentally induce mitochondrial fission by rupturing the cell's plasma membrane. We present a stability analysis that successfully explains the observed fission wavelength and the role of mitochondrial morphology in the occurrence of fission events. Our results show that the laws of fluid mechanics can describe mitochondrial morphology and dynamics.

  2. A threshold for dissipative fission

    International Nuclear Information System (INIS)

    Thoennessen, M.; Bertsch, G.F.

    1993-01-01

    The empirical domain of validity of statistical theory is examined as applied to fission data on pre-fission data on pre-fission neutron, charged particle, and γ-ray multiplicities. Systematics are found of the threshold excitation energy for the appearance of nonstatistical fission. From the data on systems with not too high fissility, the relevant phenomenological parameter is the ratio of the threshold temperature T thresh to the (temperature-dependent) fission barrier height E Bar (T). The statistical model reproduces the data for T thresh /E Bar (T) thresh /E Bar (T) independent of mass and fissility of the systems

  3. Fission product yield evaluation for the USA evaluated nuclear data files

    International Nuclear Information System (INIS)

    Rider, B.F.; England, T.R.

    1994-01-01

    An evaluated set of fission product yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  4. The uranium bearing shale ore-body at St-Hippolyte (Haut-Rhin). An example of research with statistical methods; Le gisement des schistes uraniferes de St-Hippolyte (Haut-Rhin). Exemple d'etude par calculs statistiques

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

    The uranium bearing shale ore-body at St-Hippolyte was mainly proved by drillings, the results of which were studied through statistical methods. The author gives an account of his conclusions regarding the distribution of mineralization, its origin and, the estimate of reserves. The uranium mineralization is irregularly distributed in shales. On a vertical line, levels could be singled out: they are similar, as far as lithology is concerned, but each of them is characterized by a grade-population, according to a lognormal distribution. Horizontally, a connection is noted between grades and the overlying barren sandstone bed. These considerations, as well as a statistical study of U/Ra ratio, induced the author to consider that the mineralization of the richest level has an hydrothermal origin. It is only through an uranium diffusion from that level that the others are mineralized. The uranium which is contained in poorest beds has a syn-genetic origin. Furthermore, statistical methods bring us to an evaluation of reserves. In such a case, the evaluation is equivalent to the ore obtained by common arithmetical methods. Moreover, we are able to state precisely the upper and lower limits where a true tonnage or a true grade could be given with a definite value of statistical certainty. Then the author has been able to study the separation of reserves in grade-groups and to foresee the effect of sorting in connection with the lower possible grade and with extraction units (wagons, lorries, etc...), on which the sorting will be done. To conclude, the author indicates the value of both classical and statistical methods. These two techniques are completing each other and they solve different problems. (author) [French] Le gisement des schistes uraniferes de St-Hippolyte a ete reconnu essentiellement par sondages, dont les resultats ont ete etudies par les methodes du calcul statistique. L'auteur expose les conclusions auxquelles il a ete amene et qui concernent: la

  5. Prompt fission neutron spectra and average prompt neutron multiplicities

    International Nuclear Information System (INIS)

    Madland, D.G.; Nix, J.R.

    1983-01-01

    We present a new method for calculating the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity anti nu/sub p/ as functions of the fissioning nucleus and its excitation energy. The method is based on standard nuclear evaporation theory and takes into account (1) the motion of the fission fragments, (2) the distribution of fission-fragment residual nuclear temperature, (3) the energy dependence of the cross section sigma/sub c/ for the inverse process of compound-nucleus formation, and (4) the possibility of multiple-chance fission. We use a triangular distribution in residual nuclear temperature based on the Fermi-gas model. This leads to closed expressions for N(E) and anti nu/sub p/ when sigma/sub c/ is assumed constant and readily computed quadratures when the energy dependence of sigma/sub c/ is determined from an optical model. Neutron spectra and average multiplicities calculated with an energy-dependent cross section agree well with experimental data for the neutron-induced fission of 235 U and the spontaneous fission of 252 Cf. For the latter case, there are some significant inconsistencies between the experimental spectra that need to be resolved. 29 references

  6. New fission valley for 258Fm and nuclei beyond

    International Nuclear Information System (INIS)

    Moeller, P.; Nix, J.R.; Swiatecki, W.J.

    1986-01-01

    Experimental results on the fission properties of nuclei close to 264 Fm show sudden and large changes with a change of only one or two neutrons or protons. The nucleus 258 Fm, for instance, undergoes symmetric fission with a half-life of about 0.4 ms and a kinetic energy peaked at about 235 MeV whereas 256 Fm undergoes asymmetric fission with a half-life of about 3 h and a kinetic energy peaked at about 200 MeV. Qualitatively, these sudden changes hve been postulated to be due to the emergence of fragment shells in symmetric fission products close to 132 Sn. A quantitative calculation that shows where high-kinetic-energy symmetric fission occurs and why it is associated with a sudden and large decrease in fission half-lives. The study is based on calculations of potential-energy surfaces in the macroscopic-microscopic model and a semi-empirical model for the nuclear inertia. The implications of the new fission valley on the stability of the heaviest elements is discussed. 33 refs., 12 figs

  7. Barium 139 as Fission Indicator

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1943-07-01

    This report is based on a measurement performed at the Cavendish Laboratory (Cambridge) by E. Broda in December 1943 where a technique has been worked out for measuring the fission density in a uranium containing medium in relative units by determining the amount of a suitable fission product formed. Generally a given fission product will be formed in natural uranium by slow neutron fission of U235 or by fast neutron fission of either U235 or U238. It is intended to translate the relative units into absolute units by comparison of the Ba yield with the indication of UF6 fission chamber in the same medium. This has to be done separately for fast and slow neutron fission as the yields may be different. Another application of the technique developed is the measurement of thermal neutron density in an uraniferous medium without using a detector subject to variations of sensitivity according to the properties of the medium. (nowak)

  8. Description of light charged particle emission in ternary fission

    International Nuclear Information System (INIS)

    Andreev, A. V.; Adamian, G. G.; Antonenko, N. V.; Kuklin, S. N.; Scheid, W.

    2010-01-01

    We consider the motion of three fragments starting from the scission point of ternary system. In the alpha-accompanied ternary fission the initial conditions are not the free parameters and determined by minimization of potential energy at scission point. In the trajectory calculations the angular distribution and mean value of the kinetic energy of the alpha-particles are well described in the spontaneous ternary fission of 252 Cf. In the Be- and C-accompanied ternary fission we found that the emission of the third particle occurs from one of the heavy fragments after their separation. (authors)

  9. Alecto 1 - criticality experiment on a solution of plutonium and of uranium 235. Experimental results and calculations on tank number 2 ({phi} 300 mm); Alecto 1 - experience de criticite sur une solution de plutonium et d'uranium enrichi a 90 pour cent. Resultats experimentaux et calculs concernant la cuve no. 2 ({phi} = 300 mm)

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, J G; Brunet, J P; Clouet D' Orval, Ch; Kremser, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Experiments on plutonium and 90 pour cent enriched uranium solutions have been made in the Alecto reactor with a tank of external diameter 300 mm. Various geometries Lave been tested, for variable concentrations of fissionable salts. The critical mass was studied as a function of the concentration in various reflector conditions (water, concrete, wood) and the experimental values were compared with calculated values. The effects of cadmium as a reflector and of the stainless steel tank were also studied. Lastly were carried out measurements of {beta}/{tau}, ratio of the effective fraction of delayed neutrons to the average lifetime of the neutrons in the reactor. (authors) [French] Des experiences sur des solutions de plutonium et d'uranium enrichi a 90 pour cent ont ete effectuees dans le reacteur Alecto, avec une cuve de diametre exterieur 300 mm. Diverses configurations geometriques ont ete realisees, pour des concentrations variables du sel fissile. On a etudie la masse critique en fonction de la concentration, dans plusieurs conditions de reflexion (eau, beton, bois), et on a compare les resultats experimentaux aux valeurs donnees par le calcul. On a egalement etudie l'influence du cadmium comme reflecteur et celle de la cuve d'acier inoxydable. Enfin on a effectue des mesures de {beta}/{tau}, rapport de la proportion effective des neutrons retardes au temps de vie moyen des neutrons dans la pile. (auteurs)

  10. Measurement of the fission cross section of uranium-235 between 4 eV and 20 keV; Mesure de la section efficace de fission de l'uranium-235 entre 4 eV et 20 keV

    Energy Technology Data Exchange (ETDEWEB)

    Michaudon, A; Genin, R; Joly, R; Vendryes, G

    1959-01-01

    The neutron fission cross section of uranium-235 has been measured between 4 ev and 20 kev by the time of flight method with the Saclay electron linear accelerator as a pulsed neutron source. After a brief description of the experimental apparatus and the conditions of work during the experiment, the curve {sigma}{sub F} {radical}E in the energy range studied is shown. This curve is then analyzed by the ''area'' method and a set of {sigma}{sub 0} {gamma}{sub F} values is obtained. With {sigma}{sub 0} {gamma} values measured in other laboratories, it is possible to compute fission widths for several resonances and to study their distribution. This distribution is then compared to Porter-Thomas distributions with different values of the number of exit channels. (authors) [French] La section efficace de fission de l'uranium--235 a ete mesuree entre 4 eV et 20 KeV par la methode du temps de vol en utilisant l'accelerateur lineaire a electrons de Saclay comme source pulses de neutrons. Apres une rapide description de l'appareillage experimental et des conditions de fonctionnement au cours de l'experience, on presente la courbe {sigma}{sub F} {radical}E obtenue dans la game d'energie etudiee. Cette courbe est ensuite analysee par la methode de surface des resonances et un lot de valeurs de {sigma}{sub 0} {gamma}{sub F} est obtenue. Conjuguee avec les valeurs de {sigma}{sub 0} {gamma} obtenues dans d'autres laboratoires, cette analyse permet de calculer les largeurs de fission pour plusieurs resonances et d'etudier leur distribution. Cette distribution est ensuite comparee aux distributions de Porter et Thomas correspondant a differentes valeurs du nombre de voies de sortie. (auteurs)

  11. Remarks on the fission barriers of super-heavy nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, S. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Goethe-Universitaet Frankfurt, Institut fuer Physik, Frankfurt (Germany); Heinz, S.; Mann, R.; Maurer, J.; Muenzenberg, G.; Barth, W.; Dahl, L.; Kindler, B.; Kojouharov, I.; Lang, R.; Lommel, B.; Runke, J.; Scheidenberger, C.; Tinschert, K. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Antalic, S. [Comenius University, Department of Nuclear Physics and Biophysics, Bratislava (Slovakia); Eberhardt, K.; Thoerle-Pospiech, P.; Trautmann, N. [Johannes Gutenberg-Universitaet Mainz, Mainz (Germany); Grzywacz, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Tennessee, Knoxville, TN (United States); Hamilton, J.H. [Vanderbilt University, Department of Physics and Astronomy, Nashville, TN (United States); Henderson, R.A.; Kenneally, J.M.; Moody, K.J.; Shaughnessy, D.A.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Miernik, K. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Warsaw, Warsaw (Poland); Miller, D. [University of Tennessee, Knoxville, TN (United States); Morita, K. [RIKEN Nishina Center for Accelerator-Based Science, Wako, Saitama (Japan); Nishio, K. [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Popeko, A.G.; Yeremin, A.V. [Joint Institute for Nuclear Research, Dubna (Russian Federation); Roberto, J.B.; Rykaczewski, K.P. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Uusitalo, J. [University of Jyvaeskylae, Department of Physics, Jyvaeskylae (Finland)

    2016-04-15

    Shell-correction energies of super-heavy nuclei are approximated by using Q{sub α} values of measured decay chains. Five decay chains were analyzed, which start at the isotopes {sup 285}Fl, {sup 294}118, {sup 291}Lv, {sup 292}Lv and {sup 293}Lv. The data are compared with predictions of macroscopic-microscopic models. Fission barriers are estimated that can be used to eliminate uncertainties in partial fission half-lives and in calculations of evaporation-residue cross-sections. In that calculations, fission probability of the compound nucleus is a major factor contributing to the total cross-section. The data also provide constraints on the cross-sections of capture and quasi-fission in the entrance channel of the fusion reaction. Arguments are presented that fusion reactions for synthesis of isotopes of elements 118 and 120 may have higher cross-sections than assumed so far. (orig.)

  12. Ternary fission in an effective liquid drop model

    International Nuclear Information System (INIS)

    Duarte, Sergio B.; Tavares, Odilon A.P.; Dimarco, A.; Goncalves, Marcello; Guzman, Fernando; Trallero-Herrera, Carlos; Rodriguez, Oscar; Garcia, Fermin

    2001-01-01

    Full text follows: The nuclear partition in three fragments has been observed in recent experiments for fission process of 252 Cf and 24 '0 Pu. We apply the Effective Liquid Drop Model (ELDM), successfully used for discussing binary cold fission and cluster emissions for a three center geometric shape parametrization, describing the quasi-molecular deformation which can lead to ternary fragmentation. A preliminary calculation for rates of these processes are performed and the results are compared to the rate of the dominant binary fission process. A large range of parent nuclei (spherical and deformed) is covered in the calculation. The purpose is to point out others possible ternary fission process experimentally measurable. (author)

  13. Fission gas retention in irradiated metallic fuel

    International Nuclear Information System (INIS)

    Fenske, G.R.; Gruber, E.E.; Kramer, J.M.

    1987-01-01

    Theoretical calculations and experimental measurements of the quantity of retained fission gas in irradiated metallic fuel (U-5Fs) are presented. The calculations utilize the Booth method to model the steady-state release of gases from fuel grains and a simplified grain-boundary gas model to predict the gas release from intergranular regions. The quantity of gas retained in as-irradiated fuel was determined by collecting the gases released from short segments of EBR-II driver fuel that were melted in a gas-tight furnace. Comparison of the calculations to the measurements shows quantitative agreement with both the magnitude and the axial variation of the retained gas content

  14. Influence of the cosmic-ray induced fission tracks on the fission track of extraterrestric minerals via the 238U spontaneous fission

    International Nuclear Information System (INIS)

    Damm, G.; Thiel, K.

    1977-01-01

    The age determined by counting fission tracks of lunar and meteorite materials is obviously falsified by additional fission track parts not to be accounted for by the spontaneous fission of uranium 238. For this p and n induced fissions of U, Th and other hreavy elements through the cosmic radiation come into consideration. In order to determine the possible part of such interference factors, a simulation experiment at the proton synchrocycloton (CERN, Geneva) has been carried out and independently of this, the production rates for the p and n induced U, Th, Bi, Pb and Au in the surface-near regolith layers of the moon were calculated. It could be seen that the irradiation age as well as the spacial distribution of the heavy metals in the samples to be dated must be considered. (RB) [de

  15. Angular distribution of oriented nucleus fission neutrons

    International Nuclear Information System (INIS)

    Barabanov, A.L.; Grechukhin, D.P.

    1982-01-01

    Calculations of anisotropy of angular distribution of oriented 235 U nuclei thermal fission neutrons have been carried out. the neutrons were assumed to evaporate isotropically by completely accelerated fragements in the fragment system with only its small part, i. e. fission-producing neutrons, emitted at the moment of neck break. It has been found out that at low energies of neutrons Esub(n)=1-2 MeV the sensitivity of the angular distribution anisotropy to variations of spectrum of neutron evaporation from fragments and the magnitude of a share of fission-producing neutrons reaches approximately 100%, which at high energies, Esub(n) > 5 MeV it does not exceed approximately 20%. Therefore the angular distribution of fast neutrons to a greater degree of confidence may be used for restoring the angular distribution anisotropy of fragments while the angular distribution of low energy neutrons may be used for deriving information on the fission process, but only in case 6f the experiment accuracy is better than approximately 3%

  16. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  17. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  18. Calculation system for physical analysis of boiling water reactors; Modelisation des phenomenes physiques specifiques aux reacteurs a eau bouillante, notamment le couplage neutronique-thermohydraulique

    Energy Technology Data Exchange (ETDEWEB)

    Bouveret, F

    2001-07-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  19. The fission track method

    International Nuclear Information System (INIS)

    Hansen, K.

    1990-01-01

    During the last decade fission track (FT) analysis has evolved as an important tool in exploration for hydrocarbon resources. Most important is this method's ability to yield information about temperatures at different times (history), and thus relate oil generation and time independently of other maturity parameters. The purpose of this paper is to introduce the basics of the method and give an example from the author's studies. (AB) (14 refs.)

  20. GROGi-F. Modified version of GROGi 2 nuclear evaporation computer code including fission decay channel

    International Nuclear Information System (INIS)

    Delagrange, H.

    1977-01-01

    This report is the user manual of the GR0GI-F code, modified version of the GR0GI-2 code. It calculates the cross sections for heavy ion induced fission. Fission probabilities are calculated via the Bohr-Wheeler formalism

  1. The use of averages and other summation quantities in the testing of evaluated fission product yield and decay data. Applications to ENDF/B(IV)

    International Nuclear Information System (INIS)

    Walker, W.H.

    1976-01-01

    Averages of some fission product properties can be obtained by multiplying the fission product yield for each fission product by the value of the property (e.g. mass, atomic number, mass defect) for that fission product and summing all significant contributions. These averages can be used to test the reliability of the yield set or provide useful data for reactor calculations. The report gives the derivation of these averages and discusses their application using the ENDF/B(IV) fission product library. The following quantities are treated here: the number of fission products per fission ΣYsub(i); the average mass number and the average number of neutrons per fission; the average atomic number of the stable fission products and the average number of β-decays per fission; the average mass defect of the stable fission products and the total energy release per fission; the average decay energy per fission (beta, gamma and anti-neutrino); the average β-decay energy per fission; individual and group-averaged delayed neutron emission; the total yield for each fission product element. Wherever it is meaningful to do so, a sum is subdivided into its light and heavy mass components. The most significant differences between calculated values based on ENDF/B(IV) and measurements are the β and γ decay energies for 235 U thermal fission and delayed neutron yields for other fissile nuclides, most notably 238 U. (author)

  2. Measurement of MA fission cross sections at YAYOI

    Energy Technology Data Exchange (ETDEWEB)

    Ohkawachi, Yasushi; Ohki, Shigeo; Wakabayashi, Toshio [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Fission cross section ratios of minor actinide nuclides (Am-241, Am-243) relative to U-235 in the fast neutron energy region have been measured using a back-to-back (BTB) fission chamber at YAYOI fast neutron source reactor. A small BTB fission chamber was developed to measure the fission cross section ratios in the center of the core at YAYOI reactor. Dependence of the fission cross section ratios on neutron spectra was investigated by changing the position of the detector in the reactor core. The measurement results were compared with the fission cross sections in the JENDL-3.2, ENDF/B-VI and JEF-2.2 libraries. It was found that calculated values of Am-241 using the JENDL-3.2, ENDF/B-VI and JEF-2.2 data are lower by about 15% than the measured value in the center of the core (the neutron average energy is 1.44E+6(eV)). And, good agreement can be seen the measured value and calculated value of Am-243 using the JENDL-3.2 data in the center of the core (the neutron average energy is 1.44E+6)(eV), but calculated values of Am-243 using the ENDF/B-VI and JEF-2.2 data are lower by 11% and 13% than the measured value. (author)

  3. Fission blanket benchmark experiment on spherical assembly of uranium and PE with PE reflector

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Tonghua; Lu, Xinxin; Wang, Mei; Han, Zijie, E-mail: neutron_integral@aliyun.com; Jiang, Li; Wen, Zhongwei; Liu, Rong

    2016-04-15

    Highlights: • The fission rate distribution on two depleted uranium assemblies was measured with plate fission chambers. • We do calculations using MCNP code and ENDF/B-V.0 library. • The overestimation of calculations to the measured fission rates was found. • The observed discrepancy are discussed. - Abstract: New concept of fusion-fission hybrid for energy generation has been proposed. To validate the nuclear performance of fission blanket of hybrid, as part of series of validation experiment, two types of fission blanket assemblies were setup in this work and measurements were made of the reaction rate distribution for uranium fission in the spherical assembly of depleted uranium and polyethylene by Plate Fission Chamber (PFC). There are two PFCs in experiment, one is depleted uranium chamber and the other is enriched uranium chamber. The Monte-Carlo transport code MCNP5 and continuous energy cross sections library ENDF/BV.0 were used for the analysis of fission rate distribution in the two types of assemblies. The calculated results were compared with the experimental ones. The overestimation of fission rate for depleted uranium and enriched uranium were found in the inner boundary of the two assemblies. However, the C/E ratio tends to decrease for the distance from the core slightly and the results for enriched uranium are better than that for depleted uranium.

  4. Quality control of the treatment planning systems dose calculations in external radiation therapy using the Penelope Monte Carlo code; Controle qualite des systemes de planification dosimetrique des traitements en radiotherapie externe au moyen du code Monte-Carlo Penelope

    Energy Technology Data Exchange (ETDEWEB)

    Blazy-Aubignac, L

    2007-09-15

    The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)

  5. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  6. Robust singlet fission in pentacene thin films with tuned charge transfer interactions.

    Science.gov (United States)

    Broch, K; Dieterle, J; Branchi, F; Hestand, N J; Olivier, Y; Tamura, H; Cruz, C; Nichols, V M; Hinderhofer, A; Beljonne, D; Spano, F C; Cerullo, G; Bardeen, C J; Schreiber, F

    2018-03-05

    Singlet fission, the spin-allowed photophysical process converting an excited singlet state into two triplet states, has attracted significant attention for device applications. Research so far has focused mainly on the understanding of singlet fission in pure materials, yet blends offer the promise of a controlled tuning of intermolecular interactions, impacting singlet fission efficiencies. Here we report a study of singlet fission in mixtures of pentacene with weakly interacting spacer molecules. Comparison of experimentally determined stationary optical properties and theoretical calculations indicates a reduction of charge-transfer interactions between pentacene molecules with increasing spacer molecule fraction. Theory predicts that the reduced interactions slow down singlet fission in these blends, but surprisingly we find that singlet fission occurs on a timescale comparable to that in pure crystalline pentacene. We explain the observed robustness of singlet fission in such mixed films by a mechanism of exciton diffusion to hot spots with closer intermolecular spacings.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

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

  8. Developpements numeriques recents realises en aeroelasticite chez Dassault Aviation pour la conception des avions de combat modernes et des avions d’affaires

    Science.gov (United States)

    2003-03-01

    Cost through Advanced Modelling and Virtual Simulation [La reduction des couts et des delais d’acquisition des vehicules militaires par la modelisation...sont les 6quations de restitution, par le mod~e, des frdquences et des amortissements des modes adrodlastiques mesurds h une prdcision F- donnde. Afin... amortissements mesurds h 37800 Pa et 60000 Pa (points nettemnent inferieurs A la vitesse critique). Comme le montre ce diagramme, le calcul, recal6 h

  9. Sorption Regularities in Behaviour of Fission-Product Elements during Filtration of their Solutions through Ground; Lois de l'Absorption des Elements Radioactifs Lors du Filtrage des Solutions a Travers les Terrains; 0417 0410 041a 041e 041d 0414 ; Leyes de la Absorcion de los Elementos Radiactivos Cuando se Filtran Soluciones a Traves de las Formaciones Geologicas

    Energy Technology Data Exchange (ETDEWEB)

    Spitsyn, V. I.; Balukova, V. D.; Gromov, V. V.; Zakharov, S. I.; Zhagin, B. P.; Spiridonov, F. M.

    1960-07-01

    Research on the sorption of radioisotopes under natural conditions employing the controlled filtration process was performed. Radioisotopes were introduced into the solution as soon as filtration had become steady and the process continued for four months. Soil samples were then taken by drilling at different depths and analysed to determine their radioisotope content. Diffusion of radioisotopes was observed at depths of up to 10 m; two distinct boundaries of soil-activity decrease were ascertained: at the surface of the site and at the depth of the solution filtration front. In addition, the radiostrontium absorption by natural sorbents, principally pure minerals widely distributed in soils and subsoils, was investigated separately. The presence of calcium ions, even in small quantities, sharply reduces the degree of radiostrontium sorption. However, other conditions being equal, strontium may be absorbed to a greater extent than calcium, according to the composition of the sorbent. The field investigations of radiostrontium sorption and migration showed that when filtering radioactive solutions two possible variants have to be taken into account. In the first case the solutions are discharged into soil unaffected by any flow of ground water. In this situation the radiostrontium is retained by the soil. In the second case, the radioisotopes proceed directly into the water-bearing horizon. The radiostrontium will then migrate with the ground water flow and through the soil and this migration will be further affected by the sorption and desorption processes occurring. The experiments performed demonstrate the ease with which long-lived radioisotopes migrate under natural conditions and call attention to the need for thorough study of ground water problems in connexion with various methods of disposing of radioactive waste into ground. (author) [French] En declenchant un processus de filtration, il a ete possible d'etudier l'absorption des elements radioactifs dans

  10. Symmetric and asymmetric ternary fission of hot nuclei

    International Nuclear Information System (INIS)

    Siwek-Wilczynska, K.; Wilczynski, J.; Leegte, H.K.W.; Siemssen, R.H.; Wilschut, H.W.; Grotowski, K.; Panasiewicz, A.; Sosin, Z.; Wieloch, A.

    1993-01-01

    Emission of α particles accompanying fusion-fission processes in the 40 Ar + 232 Th reaction at E( 40 Ar) = 365 MeV was studied in a wide range of in-fission-plane and out-of-plane angles. The exact determination of the emission angles of both fission fragments combined with the time-of-flight measurements allowed us to reconstruct the complete kinematics of each ternary event. The coincident energy spectra of α particles were analyzed by using predictions of the energy spectra of the statistical code CASCADE . The analysis clearly demonstrates emission from the composite system prior to fission, emission from fully accelerated fragments after fission, and also emission during scission. The analysis is presented for both symmetric and asymmetric fission. The results have been analyzed using a time-dependent statistical decay code and confronted with dynamical calculations based on a classical one-body dissipation model. The observed near-scission emission is consistent with evaporation from a dinuclear system just before scission and evaporation from separated fragments just after scission. The analysis suggests that the time scale of fission of the hot composite systems is long (about 7x10 -20 s) and the motion during the descent to scission almost completely damped

  11. Dynamic treatment of fission and fusion in two dimensions

    International Nuclear Information System (INIS)

    Nazareth, R.A.M.S.

    1977-01-01

    The barrier penetrability in two dimensions for nuclear fusion and fission phenomena is studied. The equations of fission static trajectories (minimum potential) in Hofmann formalism are derived and the influence of inertia parameters on the penetrability is verified. For fusion case, a realistic potential for exactly penetrability calculation is proposed. The transverse momentum to the fusion and the unidimensional calculation in classical approximation by choose the trajectory which turn into maximum the penetrability are considered. The exact penetrability is compared with calculation in the classical approximation which takes in account the possibility of appearing discontinuity in the barrier along of fusion pathway. (M.C.K.) [pt

  12. Recovery of trans-plutonium elements; Recuperation des elements transplutoniens

    Energy Technology Data Exchange (ETDEWEB)

    Espie, J Y; Poncet, B; Simon, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1970-07-01

    The object of this work is to study the recovery of americium and curium from the fission-product solution obtained from the processing of irradiated fuel elements made of natural metallic uranium alloyed with aluminium, iron and silicon; these elements have been subjected to an average irradiation of 4000 MW days/ton in a gas-graphite type reactor having a thermal power of 3.7 MW/ton of uranium. The process used consists of 3 extraction cycles and one americium-curium separation: - 1) extraction cycle in 40 per cent TBP: extraction of actinides and lanthanides; elimination of fission products; - 2) extraction cycle in 8 per cent D2EHPA: decontamination from the fission products, decontamination of actinides from lanthanides; - 3) extraction cycle in 40 per cent TBP: separation of the complexing agent and concentration of the actinides; - 4) americium-curium separation by precipitation. (authors) [French] Cette etude a pour objet, la recuperation de l'americium et du curium de la solution de produits de fission provenant du traitement de combustibles irradies a base d'uranium naturel metallique allie a l'aluminium, le fer, et le silicium, et ayant subi une irradiation moyenne de 4000 MWj/t dans une pile du type graphite-gaz, dont la puissance thermique est de 3.7 MW/t d'uranium. Le procede utilise comprend 3 cycles d'extraction et une separation americium-curium: - 1. cycle d'extraction dans le TBP a 40 pour cent: extraction des actinides et des lanthanides, elimination des produits de fission; - 2. cycle d'extraction dans le D2EHPA a 8 pour cent: decontamination en produits de fission, decontamination des actinides en lanthanides; - 3. cycle d