Calculation of flux density distribution on irradiation field of electron accelerator

International Nuclear Information System (INIS)

Tanaka, Ryuichi

1977-03-01

The simple equation of flux density distribution in the irradiation field of an ordinary electron accelerator is a function of the physical parameters concerning electron irradiation. Calculation is based on the mean square scattering angle derived from a simple multiple scattering theory, with the correction factors of air scattering, beam scanning and number transmission coefficient. The flux density distribution was measured by charge absorption in a graphite target set in the air. For the calculated mean square scattering angles of 0.089-0.29, the values of calculation agree with those by experiment within about 10% except at large scattering angles. The method is applicable to dose evaluation of ordinary electron accelerators and design of various irradiators for radiation chemical reaction. Applicability of the simple multiple scattering theory in calculation of the scattered flux density and periodical variation of the flux density of scanning beam are also described. (auth.)

SNS Sample Activation Calculator Flux Recommendations and Validation

Energy Technology Data Exchange (ETDEWEB)

McClanahan, Tucker C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS); Gallmeier, Franz X. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS); Iverson, Erik B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS); Lu, Wei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Spallation Neutron Source (SNS)

2015-02-01

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) uses the Sample Activation Calculator (SAC) to calculate the activation of a sample after the sample has been exposed to the neutron beam in one of the SNS beamlines. The SAC webpage takes user inputs (choice of beamline, the mass, composition and area of the sample, irradiation time, decay time, etc.) and calculates the activation for the sample. In recent years, the SAC has been incorporated into the user proposal and sample handling process, and instrument teams and users have noticed discrepancies in the predicted activation of their samples. The Neutronics Analysis Team validated SAC by performing measurements on select beamlines and confirmed the discrepancies seen by the instrument teams and users. The conclusions were that the discrepancies were a result of a combination of faulty neutron flux spectra for the instruments, improper inputs supplied by SAC (1.12), and a mishandling of cross section data in the Sample Activation Program for Easy Use (SAPEU) (1.1.2). This report focuses on the conclusion that the SAPEU (1.1.2) beamline neutron flux spectra have errors and are a significant contributor to the activation discrepancies. The results of the analysis of the SAPEU (1.1.2) flux spectra for all beamlines will be discussed in detail. The recommendations for the implementation of improved neutron flux spectra in SAPEU (1.1.3) are also discussed.

Pellet by pellet neutron flux calculations coupled with nodal expansion method

International Nuclear Information System (INIS)

Aldo, Dall'Osso

2003-01-01

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

A new calculation of atmospheric neutrino flux: the FLUKA approach

International Nuclear Information System (INIS)

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

1999-01-01

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

Calculation of conventional and prompt lepton fluxes at very high energy

CERN Document Server

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Can, B.

1982-01-01

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

Calculation of neutron flux in the presence of a source

International Nuclear Information System (INIS)

Planchard, J.

1993-09-01

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

Calculation of conventional and prompt lepton fluxes at very high energy

Directory of Open Access Journals (Sweden)

Fedynitch Anatoli

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1981-02-01

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

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

International Nuclear Information System (INIS)

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

1981-02-01

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

Maximum power flux of auroral kilometric radiation

International Nuclear Information System (INIS)

Benson, R.F.; Fainberg, J.

1991-01-01

The maximum auroral kilometric radiation (AKR) power flux observed by distant satellites has been increased by more than a factor of 10 from previously reported values. This increase has been achieved by a new data selection criterion and a new analysis of antenna spin modulated signals received by the radio astronomy instrument on ISEE 3. The method relies on selecting AKR events containing signals in the highest-frequency channel (1980, kHz), followed by a careful analysis that effectively increased the instrumental dynamic range by more than 20 dB by making use of the spacecraft antenna gain diagram during a spacecraft rotation. This analysis has allowed the separation of real signals from those created in the receiver by overloading. Many signals having the appearance of AKR harmonic signals were shown to be of spurious origin. During one event, however, real second harmonic AKR signals were detected even though the spacecraft was at a great distance (17 R E ) from Earth. During another event, when the spacecraft was at the orbital distance of the Moon and on the morning side of Earth, the power flux of fundamental AKR was greater than 3 x 10 -13 W m -2 Hz -1 at 360 kHz normalized to a radial distance r of 25 R E assuming the power falls off as r -2 . A comparison of these intense signal levels with the most intense source region values (obtained by ISIS 1 and Viking) suggests that multiple sources were observed by ISEE 3

Radiative flux calculations at UV and visible wavelengths

International Nuclear Information System (INIS)

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

1993-10-01

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

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

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

Maximum allowable heat flux for a submerged horizontal tube bundle

International Nuclear Information System (INIS)

McEligot, D.M.

1995-01-01

For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose quotes) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration

THE RISE AND FALL OF OPEN SOLAR FLUX DURING THE CURRENT GRAND SOLAR MAXIMUM

International Nuclear Information System (INIS)

Lockwood, M.; Rouillard, A. P.; Finch, I. D.

2009-01-01

We use geomagnetic activity data to study the rise and fall over the past century of the solar wind flow speed V SW , the interplanetary magnetic field strength B, and the open solar flux F S . Our estimates include allowance for the kinematic effect of longitudinal structure in the solar wind flow speed. As well as solar cycle variations, all three parameters show a long-term rise during the first half of the 20th century followed by peaks around 1955 and 1986 and then a recent decline. Cosmogenic isotope data reveal that this constitutes a grand maximum of solar activity which began in 1920, using the definition that such grand maxima are when 25-year averages of the heliospheric modulation potential exceeds 600 MV. Extrapolating the linear declines seen in all three parameters since 1985, yields predictions that the grand maximum will end in the years 2013, 2014, or 2027 using V SW , F S , or B, respectively. These estimates are consistent with predictions based on the probability distribution of the durations of past grand solar maxima seen in cosmogenic isotope data. The data contradict any suggestions of a floor to the open solar flux: we show that the solar minimum open solar flux, kinematically corrected to allow for the excess flux effect, has halved over the past two solar cycles.

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

International Nuclear Information System (INIS)

Kurosawa, M.

2005-01-01

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

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

Science.gov (United States)

Kurosawa, Masahiko

2005-01-01

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

Stationary neutrino radiation transport by maximum entropy closure

International Nuclear Information System (INIS)

Bludman, S.A.

1994-11-01

The authors obtain the angular distributions that maximize the entropy functional for Maxwell-Boltzmann (classical), Bose-Einstein, and Fermi-Dirac radiation. In the low and high occupancy limits, the maximum entropy closure is bounded by previously known variable Eddington factors that depend only on the flux. For intermediate occupancy, the maximum entropy closure depends on both the occupation density and the flux. The Fermi-Dirac maximum entropy variable Eddington factor shows a scale invariance, which leads to a simple, exact analytic closure for fermions. This two-dimensional variable Eddington factor gives results that agree well with exact (Monte Carlo) neutrino transport calculations out of a collapse residue during early phases of hydrostatic neutron star formation

Error estimates for ice discharge calculated using the flux gate approach

Science.gov (United States)

Navarro, F. J.; Sánchez Gámez, P.

2017-12-01

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

Rate maximum calculation of Dpa in CNA-II pressure vessel

International Nuclear Information System (INIS)

Mascitti, J. A

2012-01-01

The maximum dpa rate was calculated for the reactor in the following state: fresh fuel, no Xenon, a Boron concentration of 15.3 ppm, critical state, its control rods in the criticality position, hot, at full power (2160 MW). It was determined that the maximum dpa rate under such conditions is 3.54(2)x10 12 s -1 and it is located in the positions corresponding to θ=210 o in the azimuthal direction, and z=20 cm and -60 cm respectively in the axial direction, considering the calculation mesh centered at half height of the fuel element (FE) active length. The dpa rate spectrum was determined as well as the contribution to it for 4 energy groups: a thermal group, two epithermal groups and a fast one. The maximum dpa rate considering the photo-neutrons production from (γ, n) reaction in the heavy water of coolant and moderator was 3.93(4)x10 12 s -1 that is 11% greater than the obtained without photo-neutrons. This verified significant difference between both cases, suggest that photo-neutrons in large heavy water reactors such as CNA-II should not be ignored. The maximum DPA rate in the first mm of the reactor pressure vessel was calculated too and it was obtained a value of 4.22(6)x10 12 s -1 . It should be added that the calculation was carried out with the reactor complete accurate model, with no approximations in spatial or energy variables. Each value has, between parentheses, a percentage relative error representing the statistical uncertainty due to the probabilistic Monte Carlo method used to estimate it. More representative values may be obtained with this method if equilibrium burn-up distribution is used (author)

Chain Rule Approach for Calculating the Time-Derivative of Flux

Energy Technology Data Exchange (ETDEWEB)

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

2017-10-03

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

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

International Nuclear Information System (INIS)

Martinez C, E.

2011-01-01

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

MURALB - a programme for calculating neutron fluxes in many groups

International Nuclear Information System (INIS)

MacDougall, J.

1977-09-01

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

Alize 3 - first critical experiment for the franco-german high flux reactor - calculations; Alize 3 - premiere experience critique pour le reacteur a haut flux franco-allemand. Calculs

Energy Technology Data Exchange (ETDEWEB)

Scharmer, K [Commissariat a l' Energie Atomique, Dir. des Piles Atomiques, Saclay (France). Centre d' Etudes Nucleaires

1969-07-01

The results of experiments in the light water cooled D{sub 2}O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k{sub eff} was smaller than 0.5 per cent {delta}k/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D{sub 2}O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [French] Les resultats des experiences faites dans la maquette critique ALIZE III, refrigeree a l'eau legere et reflechie par l'eau lourde, ont ete compares aux calculs. On a utilise un modele de la theorie de diffusion a trois groupes rapides et epithermiques et deux groupes thermiques qui se recouvrent. Ce modele a permis de calculer la distribution de puissance dans le coeur en bon accord avec les mesures, meme dans le cas d'une forte variation du spectre des neutrons dans le coeur. L'erreur entre k{sub eff} calcule et mesure etait inferieure a 0,5 pour cent {delta}k/k. Le coefficient de vide et des materiaux de structure, la reactivite des barres 'noires', les variations du spectre (rapport Cd, rapport Pu/U) et la fraction des photo-neutrons retardes sont egalement calcules. Les mesures de reactivite et de perturbation de flux dans le reflecteur, dues aux canaux, ont ete interpretees du point de vue d'un arrangement optimum des canaux pour le Reacteur a Haut Flux Franco-Allemand. (auteur)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

Schuerrer, F.

1980-01-01

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

SPACETRAN, Radiation Leakage from Cylinder with ANISN Flux Calculation

International Nuclear Information System (INIS)

Cramer, S.N.; Solomito, M.

1974-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1968-07-15

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1982-01-01

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

LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-12-15

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

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

International Nuclear Information System (INIS)

Abdel Krim, M.S.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Rapid calculation of maximum particle lifetime for diffusion in complex geometries

Science.gov (United States)

Carr, Elliot J.; Simpson, Matthew J.

2018-03-01

Diffusion of molecules within biological cells and tissues is strongly influenced by crowding. A key quantity to characterize diffusion is the particle lifetime, which is the time taken for a diffusing particle to exit by hitting an absorbing boundary. Calculating the particle lifetime provides valuable information, for example, by allowing us to compare the timescale of diffusion and the timescale of the reaction, thereby helping us to develop appropriate mathematical models. Previous methods to quantify particle lifetimes focus on the mean particle lifetime. Here, we take a different approach and present a simple method for calculating the maximum particle lifetime. This is the time after which only a small specified proportion of particles in an ensemble remain in the system. Our approach produces accurate estimates of the maximum particle lifetime, whereas the mean particle lifetime always underestimates this value compared with data from stochastic simulations. Furthermore, we find that differences between the mean and maximum particle lifetimes become increasingly important when considering diffusion hindered by obstacles.

Comparison of calculated energy flux of internal tides with microstructure measurements

Directory of Open Access Journals (Sweden)

Saeed Falahat

2014-10-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1976-07-01

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

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

International Nuclear Information System (INIS)

Fu Yanzhang; Zhu Yubao; Chen Juequan

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

TORT application in reactor pressure vessel neutron flux calculations

International Nuclear Information System (INIS)

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

1994-01-01

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

Variability of the Lyman alpha flux with solar activity

International Nuclear Information System (INIS)

Lean, J.L.; Skumanich, A.

1983-01-01

A three-component model of the solar chromosphere, developed from ground based observations of the Ca II K chromospheric emission, is used to calculate the variability of the Lyman alpha flux between 1969 and 1980. The Lyman alpha flux at solar minimum is required in the model and is taken as 2.32 x 10 11 photons/cm 2 /s. This value occurred during 1975 as well as in 1976 near the commencement of solar cycle 21. The model predicts that the Lyman alpha flux increases to as much as 5 x 10 11 photons/cm 2 /s at the maximum of the solar cycle. The ratio of the average fluxes for December 1979 (cycle maximum) and July 1976 (cycle minimum) is 1.9. During solar maximum the 27-day solar rotation is shown to cause the Lyman alpha flux to vary by as much as 40% or as little as 5%. The model also shows that the Lyman alpha flux varies over intermediate time periods of 2 to 3 years, as well as over the 11-year sunspot cycle. We conclude that, unlike the sunspot number and the 10.7-cm radio flux, the Lyman alpha flux had a variability that was approximately the same during each of the past three cycles. Lyman alpha fluxes calculated by the model are consistent with measurements of the Lyman alpha flux made by 11 of a total of 14 rocket experiments conducted during the period 1969--1980. The model explains satisfactorily the absolute magnitude, long-term trends, and the cycle variability seen in the Lyman alpha irradiances by the OSO 5 satellite experiment. The 27-day variability observed by the AE-E satellite experiment is well reproduced. However, the magntidue of the AE-E 1 Lyman alpha irradiances are higher than the model calculations by between 40% and 80%. We suggest that the assumed calibration of the AE-E irradiances is in error

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

International Nuclear Information System (INIS)

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

1975-09-01

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

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

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

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

Science.gov (United States)

Palmer, Grant

1987-01-01

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

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

Science.gov (United States)

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

2018-01-01

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

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

Science.gov (United States)

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

2009-04-01

) including a quadratic runoff module (Bustillo, 2005), 1 linear interpolation method and 2 discharge-weighted concentration methods ("M18", "M19", Philipps et al, 1999). As expected, based on 55 stations and 430 years, SPM fluxes are the most uncertain ones with maximum biases determined on annual fluxes (monthly sampling simulations). ranging at stations from -75% to +55% by the classical rating-curves approach ("M1", "M2") droping to -60% to +5% for the M18 method. At this frequency, biases are much less for Ptot and PO4-3 (-30% to +10%), nitrate (-5% to +10%) and are negligible for TDS. For higher frequencies, the biases are reduced: for instance for weekly surveys they drop to -25% for SPM and to -20% to 5% for Ptot for the M18 method. The river basin size is influencing the performance of calculations methods: SPM flux errors are much higher for smaller basins (103 to 104 km2) than for larger ones (> 104 km2), probably in relation with the flow duration in 2% of time which is a key control factor of flux duration in 2% of time (Moatar et al, 2006). This indicator based on daily flow (Q) records is generally available at water quality stations. Other indicators based on discrete water quality surveys are being tested to explain the performance of flux methods for each variable: concentrations (C) variability, C vs Q relationship, concentration seasonality. For each variable and each station the optimal flux calculation method will be derived from the future expert system. BUSTILLO V., Biogéochimie et hydroclimatologie appliquées à l'aménagement des bassins fluviaux .PhD Thesis, INP Toulouse,232 p+annexes (2005). FERGUSON R.I., Accuracy and precision of methods for estimating river loads. Earth Surface Processes and Landforms, vol. 12,95-104 (1987). MOATAR F., PERSON G., MEYBECK M., COYNEL A., ETCHEBER H., CROUZET P., The influence of contrasting suspended particulate matter transport regimes on the bias and precision of flux estimates Science of the Total Environment

Analytical 3-D force calculation of a transverse flux machine

NARCIS (Netherlands)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Jung, Sang Hoon

2006-02-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Shi, Chong; Nakajima, Teruyuki

2015-01-01

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

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Scharmer, K.

1969-01-01

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

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

CERN Document Server

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

2002-01-01

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

Application of generalized perturbation theory to flux disadvantage factor calculations

International Nuclear Information System (INIS)

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

1979-01-01

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

The calculation of maximum permissible exposure levels for laser radiation

International Nuclear Information System (INIS)

Tozer, B.A.

1979-01-01

The maximum permissible exposure data of the revised standard BS 4803 are presented as a set of decision charts which ensure that the user automatically takes into account such details as pulse length and pulse pattern, limiting angular subtense, combinations of multiple wavelength and/or multiple pulse lengths, etc. The two decision charts given are for the calculation of radiation hazards to skin and eye respectively. (author)

Simultaneous global calculation of flux and importance with forward Monte Carlo

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1963-05-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Bosevski, T.

1986-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1961-05-15

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

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

Directory of Open Access Journals (Sweden)

Di Zheng

2017-02-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

CERN Document Server

Linden, P; Dahl, B; Pázsit, I; Por, G

1999-01-01

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

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

Science.gov (United States)

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

2012-01-01

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

The calculation of neutron flux using Monte Carlo method

Science.gov (United States)

Günay, Mehtap; Bardakçı, Hilal

2017-09-01

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

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

International Nuclear Information System (INIS)

Kulacsy, K.; Lux, I.

1997-01-01

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

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

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1974-01-15

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

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

Directory of Open Access Journals (Sweden)

Masahito Ueyama

2012-07-01

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

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

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

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

Calculations for HFIR [High Flux Isotope Reactor] fuel plate non- bonding and fuel segregation uncertainty factors

International Nuclear Information System (INIS)

Kirkpatrick, J.R.

1990-10-01

The effects of non-bonds and of fuel segregation on the package factors of the heat flux in the High Flux Isotope Reactor (HFIR) are examined. The effects of the two defects are examined both separately and together. It is concluded that the peaking factors that are used in the present HFIR thermal analysis code are conservative and thus no changes in the peaking factors are necessary to continue to ensure that HFIR is safe. A study was made of the effect of the non-bond spot diameter on the peaking factor. The conclusion is that the spot can have diameter more than three times the maximum value allowed by the specifications before the peaking factor is greater than the maximum value specified in the present HFIR thermal analysis code. 6 refs., 7 figs., 8 tabs

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

International Nuclear Information System (INIS)

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

1994-08-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

Khattab, K.

2005-03-01

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

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

International Nuclear Information System (INIS)

Khattab, K.

2006-01-01

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

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

Science.gov (United States)

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

2016-12-01

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

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

Directory of Open Access Journals (Sweden)

Avramović Ivana

2007-01-01

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

Flux and brightness calculations for various synchrotron radiation sources

International Nuclear Information System (INIS)

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

1991-11-01

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

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

Science.gov (United States)

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

2016-04-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

1994-07-01

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

Neutron flux calculation by means of Monte Carlo methods

International Nuclear Information System (INIS)

Barz, H.U.; Eichhorn, M.

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Slater, C.O.

1990-07-01

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

Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1967-05-15

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

BRIEF COMMUNICATION: The negative ion flux across a double sheath at the formation of a virtual cathode

Science.gov (United States)

McAdams, R.; Bacal, M.

2010-08-01

For the case of negative ions from a cathode entering a plasma, the maximum negative ion flux and the positive ion flux before the formation of a virtual cathode have been calculated for particular plasma conditions. The calculation is based on a simple modification of an analysis of electron emission into a plasma containing negative ions. The results are in good agreement with a 1d3v PIC code model.

Thermal neutron measurement using the instrumented test bundle and assessment of maximum linear power in HANARO

Energy Technology Data Exchange (ETDEWEB)

Lee, C. S.; Seo, C. K.; Lee, B. C.; Kim, H. N.; Kang, B. W. [KAERI, Taejon (Korea, Republic of)

2000-10-01

The HANARO fuel, U{sub 3}Si-Al, has been developed by AECL and tested in NRU reactor. Due to the lack of the data performed under the high power, the repetitive conduct of the irradiation test was required under the power greater than 108kW/m, which is the estimated maximum linear power in the design stage. Accordingly, the instrumented test bundle with SPND(Self Powered Neutron Detector) was fabricated and its irradiation test was performed in IR2 of HANARO. The measured thermal neutron flux with SPND is compared with calculation results by HANAFMS(HANARO Fuel Management System). The difference in the measured and calculated thermal flux values are below {+-}11% and the accuracy of the linear power predicted by HANAFMS is consequently accompanied. Therefore, it is believed that the maximum linear power above 120kW/m is achieved during the irradiation test of the test bundle.

Maximum Expected Wall Heat Flux and Maximum Pressure After Sudden Loss of Vacuum Insulation on the Stratospheric Observatory for Infrared Astronomy (SOFIA) Liquid Helium (LHe) Dewars

Science.gov (United States)

Ungar, Eugene K.

2014-01-01

The aircraft-based Stratospheric Observatory for Infrared Astronomy (SOFIA) is a platform for multiple infrared observation experiments. The experiments carry sensors cooled to liquid helium (LHe) temperatures. A question arose regarding the heat input and peak pressure that would result from a sudden loss of the dewar vacuum insulation. Owing to concerns about the adequacy of dewar pressure relief in the event of a sudden loss of the dewar vacuum insulation, the SOFIA Program engaged the NASA Engineering and Safety Center (NESC). This report summarizes and assesses the experiments that have been performed to measure the heat flux into LHe dewars following a sudden vacuum insulation failure, describes the physical limits of heat input to the dewar, and provides an NESC recommendation for the wall heat flux that should be used to assess the sudden loss of vacuum insulation case. This report also assesses the methodology used by the SOFIA Program to predict the maximum pressure that would occur following a loss of vacuum event.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

Wasastjerna, F.; Lux, I.

1980-03-01

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

Calculation of fluences of fast neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor. Part I. Theory, calculations, comparison with the experiment

International Nuclear Information System (INIS)

Rataj, J.

1993-10-01

The method of calculating neutron spectra and integral flux densities of neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor is outlined. The one-dimensional and two-dimensional calculations were performed by means of the DORT code in R, R-Z, and R-Θ geometries using the cross sections from the ELXSIR library. In the R-Θ geometry, the coupled neutron flux densities were determined. The calculated values of the maximum activation of detectors differ less than 15% from the values measured in surveillance specimens, which is within the limit of uncertainty associated with the position of the detector in the casing. The differences between the calculated and observed data behind the pressure vessel were below 4%. 10 tabs., 3 figs., 41 refs

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

Fabrication of Anodic Aluminum Oxide Membrane for High Heat Flux Evaporation

OpenAIRE

McGrath, Kristine

2016-01-01

As electronics become more powerful and have higher energy densities, it is becoming more and more necessary to find solutions to dissipate these high heat fluxes. One solution to this problem is nanopore evaporative cooling. Based on current literature, the experimental data is far below what is expected from the theoretical calculations.In this thesis, the experimental results produced heat fluxes much closer to the theoretical values. Experimentally, a maximum heat dissipation of 103 W was...

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

International Nuclear Information System (INIS)

Dorning, J.J.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Flux pinning by voids in surface-oxidized superconducting niobium and vanadium

International Nuclear Information System (INIS)

Meij, G.P. van der.

1984-03-01

The volume pinning force in several niobium and vanadium samples with voids is determined at various temperatures. Reasonable agreement is found with the collective pinning theory of Larkin and Ovchinnikov above the field of maximum pinning, if the flux line lattice is assumed to be amorphous in this region and if the elementary pinning force is calculated from the quasi-classical theory of Thuneberg, Kurkijaervi, and Rainer. Also some history and relaxation effects are studied in an alternating field. A qualitative explanation is given in terms of flux line dislocations, which reduce the shear strength of the flux line lattice. (Auth.)

Calculate the maximum expected dose for technical radio physicists a cobalt machine

International Nuclear Information System (INIS)

Avila Avila, Rafael; Perez Velasquez, Reytel; Gonzalez Lapez, Nadia

2009-01-01

Considering the daily operations carried out by technicians Radiophysics Medical Service Department of Radiation Oncology Hospital V. General Teaching I. Lenin in the city of Holguin, during a working week (Between Monday and Friday) as an important element in calculating the maximum expected dose (MDE). From the exponential decay law which is subject the source activity, we propose corrections to the cumulative doses in the weekly period, leading to obtaining a formula which takes into a cumulative dose during working days and sees no dose accumulation of rest days (Saturday and Sunday). The estimate factor correction is made from a power series expansion convergent is truncated at the n-th term coincides with the week period for which you want to calculate the dose. As initial condition is adopted ambient dose equivalent rate as a given, which allows estimate MDE in the moments after or before this. Calculations were proposed use of an Excel spreadsheet that allows simple and accessible processing the formula obtained. (author)

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

Science.gov (United States)

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

2013-01-01

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

Solar Modulation of Inner Trapped Belt Radiation Flux as a Function of Atmospheric Density

Science.gov (United States)

Lodhi, M. A. K.

2005-01-01

No simple algorithm seems to exist for calculating proton fluxes and lifetimes in the Earth's inner, trapped radiation belt throughout the solar cycle. Most models of the inner trapped belt in use depend upon AP8 which only describes the radiation environment at solar maximum and solar minimum in Cycle 20. One exception is NOAAPRO which incorporates flight data from the TIROS/NOAA polar orbiting spacecraft. The present study discloses yet another, simple formulation for approximating proton fluxes at any time in a given solar cycle, in particular between solar maximum and solar minimum. It is derived from AP8 using a regression algorithm technique from nuclear physics. From flux and its time integral fluence, one can then approximate dose rate and its time integral dose.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

The topside ionosphere above Arecibo at equinox during sunspot maximum

International Nuclear Information System (INIS)

Bailey, G.J.

1980-01-01

The coupled time-dependent 0 + and H + continuity and momentum equations and 0 + , H + and electron heat balance equations are solved simultaneously within the L = 1.4 (Arecibo) magnetic flux tube between an altitude of 120 km and the equatorial plane. The results of the calculations are used in a study of the topside ionosphere above Arecibo at equinox during sunspot maximum. Magnetically quiet conditions are assumed. The results of the calculations show that the L = 1.4 magnetic flux tube becomes saturated from an arbitrary state within 2-3 days. During the day the ion content of the magnetic flux tube consists mainly of 0 + whereas 0 + and H + are both important during the night. There is an altitude region in the topside ionosphere during the day where ion-counterstreaming occurs with H + flowing downward and 0 + flowing upward. The conditions causing this ion-counterstreaming are discussed. There is a net chemical gain of H + at the higher altitudes. This H + diffuses both upwards and downwards whilst 0 + diffuses upwards from its solar e.u.v. production source which is most important at the lower altitudes. During the night the calculated 0 + and H + temperatures are very nearly equal whereas during the day there are occasions when the H + temperature exceeds the 0 - temperature by about 300 K. (author)

Method of Relative Magnitudes for Calculating Magnetic Fluxes in Electrical Machine

Directory of Open Access Journals (Sweden)

Oleg A.

2018-03-01

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

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

Science.gov (United States)

Walitt, L.

1982-01-01

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

Semi-analytic flux formulas for shielding calculations

International Nuclear Information System (INIS)

Wallace, O.J.

1976-06-01

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

Application of maximum values for radiation exposure and principles for the calculation of radiation doses

International Nuclear Information System (INIS)

2007-08-01

The guide presents the definitions of equivalent dose and effective dose, the principles for calculating these doses, and instructions for applying their maximum values. The limits (Annual Limit on Intake and Derived Air Concentration) derived from dose limits are also presented for the purpose of monitoring exposure to internal radiation. The calculation of radiation doses caused to a patient from medical research and treatment involving exposure to ionizing radiation is beyond the scope of this ST Guide

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

International Nuclear Information System (INIS)

Andrejsek, K.

1977-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

Mohammadnia Meysam

2013-01-01

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

Real-time diamagnetic flux measurements on ASDEX Upgrade.

Science.gov (United States)

Giannone, L; Geiger, B; Bilato, R; Maraschek, M; Odstrčil, T; Fischer, R; Fuchs, J C; McCarthy, P J; Mertens, V; Schuhbeck, K H

2016-05-01

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

Increased terrestrial to ocean sediment and carbon fluxes in the northern Chesapeake Bay associated with twentieth century land alteration

Science.gov (United States)

Saenger, C.; Cronin, T. M.; Willard, D.; Halka, J.; Kerhin, R.

2008-01-01

We calculated Chesapeake Bay (CB) sediment and carbon fluxes before and after major anthropogenic land clearance using robust monitoring, modeling and sedimentary data. Four distinct fluxes in the estuarine system were considered including (1) the flux of eroded material from the watershed to streams, (2) the flux of suspended sediment at river fall lines, (3) the burial flux in tributary sediments, and (4) the burial flux in main CB sediments. The sedimentary maximum in Ambrosia (ragweed) pollen marked peak land clearance (~1900 a.d.). Rivers feeding CB had a total organic carbon (TOC)/total suspended solids of 0.24??0.12, and we used this observation to calculate TOC fluxes from sediment fluxes. Sediment and carbon fluxes increased by 138-269% across all four regions after land clearance. Our results demonstrate that sediment delivery to CB is subject to significant lags and that excess post-land clearance sediment loads have not reached the ocean. Post-land clearance increases in erosional flux from watersheds, and burial in estuaries are important processes that must be considered to calculate accurate global sediment and carbon budgets. ?? 2008 Coastal and Estuarine Research Federation.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

The Labrador Sea during the Last Glacial Maximum: Calcite dissolution or low biogenic carbonate fluxes?

Science.gov (United States)

Marshall, Nicole; de Vernal, Anne; Mucci, Alfonso; Filippova, Alexandra; Kienast, Markus

2017-04-01

Low concentrations of biogenic carbonate characterize the sediments deposited in the Labrador Sea during the last glaciation. This may reflect poor calcite preservation and/or low biogenic carbonate productivity and fluxes. Regional bottom water ventilation was reduced during the Last Glacial Maximum (LGM), so the calcite lysocline might have been shallower than at present in the deep Labrador Sea making dissolution of calcite shells in the deep Labrador Sea possible. To address the issue, a multi-proxy approach based on micropaleontological counts (coccoliths, foraminifers, palynomorphs) and biogeochemical analyses (alkenones) was applied in the investigation of core HU2008-029-004-PC recovered in the northwestern Labrador Sea. Calcite dissolution indices based on the relative abundance benthic foraminifera shells to their organic linings as well as on fragmentation of planktonic foraminifera shells were used to evaluate changes in calcite dissolution/ preservation since the LGM. In addition, the ratio of the concentrations of coccoliths, specifically of the alkenone-producer Emiliania huxleyi, and alkenones (Emiliania huxleyi: alkenones) was explored as a potential new proxy of calcite dissolution. A sharp increase in coccoliths, foraminifers and organic linings from nearly none to substantial concentrations at 12 ka, reflect a jump to significantly greater biogenic fluxes at the glacial-interglacial transition. Furthermore, conventional dissolution indices (shells/linings of benthic foraminifera and fragmentation of planktic foraminifers) reveal that dissolution is not likely responsible for the lower glacial abundances of coccoliths and foraminifers. Only the low Emiliania huxleyi: alkenones ratios in glacial sediments could be interpreted as evidence of increased dissolution during the LGM. Given the evidence of allochthonous alkenone input into the glacial Labrador Sea, the latter observations must be treated with caution. Overall, the records indicate that

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

International Nuclear Information System (INIS)

Delchambre, E; Counsell, G; Kirk, A

2009-01-01

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

Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

Energy Technology Data Exchange (ETDEWEB)

Boscary, J

1995-10-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs.

Transport methods: general. 6. A Flux-Limited Diffusion Theory Derived from the Maximum Entropy Eddington Factor

International Nuclear Information System (INIS)

Yin, Chukai; Su, Bingjing

2001-01-01

The Minerbo's maximum entropy Eddington factor (MEEF) method was proposed as a low-order approximation to transport theory, in which the first two moment equations are closed for the scalar flux f and the current F through a statistically derived nonlinear Eddington factor f. This closure has the ability to handle various degrees of anisotropy of angular flux and is well justified both numerically and theoretically. Thus, a lot of efforts have been made to use this approximation in transport computations, especially in the radiative transfer and astrophysics communities. However, the method suffers numerical instability and may lead to anomalous solutions if the equations are solved by certain commonly used (implicit) mesh schemes. Studies on numerical stability in one-dimensional cases show that the MEEF equations can be solved satisfactorily by an implicit scheme (of treating δΦ/δx) if the angular flux is not too anisotropic so that f 32 , the classic diffusion solution P 1 , the MEEF solution f M obtained by Riemann solvers, and the NFLD solution D M for the two problems, respectively. In Fig. 1, NFLD and MEEF quantitatively predict very close results. However, the NFLD solution is qualitatively better because it is continuous while MEEF predicts unphysical jumps near the middle of the slab. In Fig. 2, the NFLD and MEEF solutions are almost identical, except near the material interface. In summary, the flux-limited diffusion theory derived from the MEEF description is quantitatively as accurate as the MEEF method. However, it is more qualitatively correct and user-friendly than the MEEF method and can be applied efficiently to various steady-state problems. Numerical tests show that this method is widely valid and overall predicts better results than other low-order approximations for various kinds of problems, including eigenvalue problems. Thus, it is an appealing approximate solution technique that is fast computationally and yet is accurate enough for a

Variability of fractal dimension of solar radio flux

Science.gov (United States)

Bhatt, Hitaishi; Sharma, Som Kumar; Trivedi, Rupal; Vats, Hari Om

2018-04-01

In the present communication, the variation of the fractal dimension of solar radio flux is reported. Solar radio flux observations on a day to day basis at 410, 1415, 2695, 4995, and 8800 MHz are used in this study. The data were recorded at Learmonth Solar Observatory, Australia from 1988 to 2009 covering an epoch of two solar activity cycles (22 yr). The fractal dimension is calculated for the listed frequencies for this period. The fractal dimension, being a measure of randomness, represents variability of solar radio flux at shorter time-scales. The contour plot of fractal dimension on a grid of years versus radio frequency suggests high correlation with solar activity. Fractal dimension increases with increasing frequency suggests randomness increases towards the inner corona. This study also shows that the low frequency is more affected by solar activity (at low frequency fractal dimension difference between solar maximum and solar minimum is 0.42) whereas, the higher frequency is less affected by solar activity (here fractal dimension difference between solar maximum and solar minimum is 0.07). A good positive correlation is found between fractal dimension averaged over all frequencies and yearly averaged sunspot number (Pearson's coefficient is 0.87).

Studies of vertical fluxes of horizontal momentum in the lower atmosphere using the MU-radar

Directory of Open Access Journals (Sweden)

F. S. Kuo

2008-11-01

Full Text Available We study the momentum flux of the atmospheric motions in the height ranges between 6 and 22 km observed using the MU radar at Shigaraki in Japan during a 3 day period in January 1988. The data were divided by double Fourier transformation into data set of waves with downward- phase- velocity and data set of waves with upward-phase-velocity for independent momentum flux calculation. The result showed that both the 72 h averaged upward flux and downward flux of zonal momentum were negative at nearly each height, meaning that the upward flux was dominated by westward propagating waves while the downward flux was dominated by eastward propagating waves. The magnitude of the downward flux was approximately a factor of 1.5 larger than the upward flux for waves in the 2~7 h and 7~24 h period bands, and about equal to the upward flux in the 10–30 min and 30 min–2 h period bands. It is also observed that the vertical flux of zonal momentum tended to be small in each frequency band at the altitudes below the jet maximum (10~12 km, and the flux increased toward more negative values to reach a negative maximum at some altitude well above the jet maximum. Daily averaged flux showed tremendous variation: The 1st 24 h (quiet day was relatively quiet, and the fluxes of the 2nd and 3rd 24 h (active days increased sharply. Moreover, the upward fluxes of zonal momentum below 17 km in the quiet day for each period band (10~30 min, 30 min~2 h, 2~7 h, and 7~24 h were dominantly positive, while the corresponding downward fluxes were dominantly negative, meaning that the zonal momentum below 17 km in each period band under study were dominantly eastward (propagating along the mean wind. In the active days, both the upward fluxes and downward fluxes in each frequency band were dominantly negative throughout the whole altitude range 6.1–18.95 km.

Neutron flux monitor

International Nuclear Information System (INIS)

Oda, Naotaka.

1993-01-01

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

Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

Energy Technology Data Exchange (ETDEWEB)

Boscary, J [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Sciences de la Matiere; [Association Euratom-CEA, Centre d` Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

1997-03-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author) 197 refs.

About Merging Threshold and Critical Flux Concepts into a Single One: The Boundary Flux

Directory of Open Access Journals (Sweden)

Marco Stoller

2014-01-01

Full Text Available In the last decades much effort was put in understanding fouling phenomena on membranes. One successful approach to describe fouling issues on membranes is the critical flux theory. The possibility to measure a maximum value of the permeate flux for a given system without incurring in fouling issues was a breakthrough in membrane process design. However, in many cases critical fluxes were found to be very low, lower than the economic feasibility of the process. The knowledge of the critical flux value must be therefore considered as a good starting point for process design. In the last years, a new concept was introduced, the threshold flux, which defines the maximum permeate flow rate characterized by a low constant fouling rate regime. This concept, more than the critical flux, is a new practical tool for membrane process designers. In this paper a brief review on critical and threshold flux will be reported and analyzed. And since the concepts share many common aspects, merged into a new concept, called the boundary flux, the validation will occur by the analysis of previously collected data by the authors, during the treatment of olive vegetation wastewater by ultrafiltration and nanofiltration membranes.

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

Science.gov (United States)

Khattab, K; Sulieman, I

2009-04-01

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

Use of sup(233)U for high flux reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Liem, P.H.

1991-01-01

The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

Neutron flux monitoring device

International Nuclear Information System (INIS)

Shimazu, Yoichiro.

1995-01-01

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

Burnout in a channel with non-uniform circumferential heat flux

International Nuclear Information System (INIS)

Lee, D.H.

1966-03-01

Burnout experiments are reported for uniform flux and circumferential flux tilt (maximum/average flux about 1.25) with tubes and annuli, all the experiments having uniform axial heating. These show similar results, the burnout power with flux tilt being within 10% of that with uniform flux. For the same mean exit steam quality, the local maximum flux is higher than the predicted burnout value and generally a better prediction is obtained using the average flux. (author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Ferreira, C.R.

1984-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

Grimstone, M.J.

1978-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-10-01

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

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

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

2004-01-01

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

Response of actinides to flux changes in high-flux systems

International Nuclear Information System (INIS)

Sailor, W.C.

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Aalto, E

1964-10-15

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

Application of maximum values for radiation exposure and principles for the calculation of radiation dose

International Nuclear Information System (INIS)

2000-01-01

The guide sets out the mathematical definitions and principles involved in the calculation of the equivalent dose and the effective dose, and the instructions concerning the application of the maximum values of these quantities. further, for monitoring the dose caused by internal radiation, the guide defines the limits derived from annual dose limits (the Annual Limit on Intake and the Derived Air Concentration). Finally, the guide defines the operational quantities to be used in estimating the equivalent dose and the effective dose, and also sets out the definitions of some other quantities and concepts to be used in monitoring radiation exposure. The guide does not include the calculation of patient doses carried out for the purposes of quality assurance

Application of maximum values for radiation exposure and principles for the calculation of radiation dose

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

The guide sets out the mathematical definitions and principles involved in the calculation of the equivalent dose and the effective dose, and the instructions concerning the application of the maximum values of these quantities. further, for monitoring the dose caused by internal radiation, the guide defines the limits derived from annual dose limits (the Annual Limit on Intake and the Derived Air Concentration). Finally, the guide defines the operational quantities to be used in estimating the equivalent dose and the effective dose, and also sets out the definitions of some other quantities and concepts to be used in monitoring radiation exposure. The guide does not include the calculation of patient doses carried out for the purposes of quality assurance.

Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow

International Nuclear Information System (INIS)

Boscary, J.

1995-10-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Natural Elemental Concentrations and Fluxes: Their Use as Indicators of Repository Safety

International Nuclear Information System (INIS)

Miller, Bill; Lind, Andy; Savage, Dave; Maul, Philip; Robinson, Peter

2002-03-01

controlling the release and transport of contaminants from the repository. In further calculations, the elemental mass fluxes of U, Th, K and Rb were used to calculate total alpha and non-alpha radioactive fluxes. For U and Th, activity fluxes were calculated for the radioelements alone as well as for their respective decay chains, assuming secular equilibrium in the chains and considering only the longer-lived nuclides with half-lives longer than one day. For K and Rb, activity fluxes were calculated for the non-series nuclides 40 K and 87 Rb. These natural activity fluxes are considered to be particularly useful safety indicators because they can be readily compared with the results from PAs, because the calculated repository releases normally expressed as dose can be recast in terms of equivalent activity fluxes. Lastly, ore bodies and hydrothermal systems were considered briefly because they provide the potential for maximum concentrations and maximum fluxes, respectively, in geological systems. Although it would be unlikely that a repository would ever be located in these geological systems, they are useful to consider here because they provide further context to the broadest variability in natural systems for comparison with the repository releases. This study has demonstrated that it is possible to compile from the published literature a substantial database of elemental abundances in natural materials and, using this data, to calculate a range of elemental and activity fluxes arising due to different processes at different spatial scales. Although it was not attempted in this work, these fluxes should be comparable to standard PA results, with some modification to the PA calculations explicitly to output the concentrations and activities associated with the repository releases

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Near bed suspended sediment flux by single turbulent events

Science.gov (United States)

Amirshahi, Seyed Mohammad; Kwoll, Eva; Winter, Christian

2018-01-01

The role of small scale single turbulent events in the vertical mixing of near bed suspended sediments was explored in a shallow shelf sea environment. High frequency velocity and suspended sediment concentration (SSC; calibrated from the backscatter intensity) were collected using an Acoustic Doppler Velocimeter (ADV). Using quadrant analysis, the despiked velocity time series was divided into turbulent events and small background fluctuations. Reynolds stress and Turbulent Kinetic Energy (TKE) calculated from all velocity samples, were compared to the same turbulent statistics calculated only from velocity samples classified as turbulent events (Reevents and TKEevents). The comparison showed that Reevents and TKEevents was increased 3 and 1.6 times, respectively, when small background fluctuations were removed and that the correlation with SSC for TKE could be improved through removal of the latter. The correlation between instantaneous vertical turbulent flux (w ‧) and SSC fluctuations (SSC ‧) exhibits a tidal pattern with the maximum correlation at peak ebb and flood currents, when strong turbulent events appear. Individual turbulent events were characterized by type, strength, duration and length. Cumulative vertical turbulent sediment fluxes and average SSC associated with individual turbulent events were calculated. Over the tidal cycle, ejections and sweeps were the most dominant events, transporting 50% and 36% of the cumulative vertical turbulent event sediment flux, respectively. Although the contribution of outward interactions to the vertical turbulent event sediment flux was low (11%), single outward interaction events were capable of inducing similar SSC ‧ as sweep events. The results suggest that on time scales of tens of minutes to hours, TKE may be appropriate to quantify turbulence in sediment transport studies, but that event characteristics, particular the upward turbulent flux need to be accounted for when considering sediment transport

Core calculations of JMTR

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Practical Wide-speed-range Sensorless Control System for Permanent Magnet Reluctance Synchronous Motor Drives via Active Flux Model

DEFF Research Database (Denmark)

Ancuti, Mihaela Codruta; Tutelea, Lucian; Andreescu, Gheorghe-Daniel

2014-01-01

This article introduces a control strategy to obtain near-maximum available torque in a wide speed range with sensorless operation via the active flux concept for permanent magnet-reluctance synchronous motor drives. A new torque dq current reference calculator is proposed, with reference torque...

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

Science.gov (United States)

Desai, Trunil S; Srivastava, Shireesh

2018-01-01

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

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

Directory of Open Access Journals (Sweden)

Nielsen Lars K

2009-05-01

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

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

International Nuclear Information System (INIS)

Forkl, A.; Kronmueller, H.

1995-01-01

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

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

Science.gov (United States)

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

2017-09-01

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

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

Directory of Open Access Journals (Sweden)

Brovchenko Mariya

2017-01-01

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

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

Science.gov (United States)

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

2017-10-01

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

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

International Nuclear Information System (INIS)

Ghassoun, Jillali; Jehoauni, Abdellatif

2000-01-01

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

Measurement and evaluation of fast neutron flux of CT and OR5 irradiation hole in HANARO

International Nuclear Information System (INIS)

Yang, Seong Woo; Choo, Kee Nam; Lee, Seung-Kyu; Kim, Yong Kyun

2012-01-01

The irradiation test has been conducted to evaluate the irradiation performance of many materials by a material capsule at HANARO. Since the fast neutron fluence above 1 MeV is important for the irradiation test of material, it must be measured and evaluated exactly at each irradiation hole. Therefore, a fast neutron flux was measured and evaluated by a 09M-02K capsule irradiated in an OR5 irradiation hole and a 10M-01K capsule irradiated in a CT irradiation hole. Fe, Ni, and Ti wires as the fluence monitor were used for the detection of fast neutron flux. Before the irradiation test, the neutron flux and spectrum was calculated for each irradiation hole using an MCNP code. After the irradiation test, the activity of the fluence monitor was measured by an HPGe detector and the reaction rate was calculated. For the OR5 irradiation hole, the radial difference of the fast neutron flux was observed from a calculated data due to the OR5 irradiation hole being located outside the core. Furthermore, a control absorber rod was withdrawn from the core as the increase of the irradiation time at the same irradiation cycle, so the distribution of neutron flux was changed from the beginning to the end of the cycle. These effects were considered to evaluate the fast neutron flux. Neutron spectrums of the CT and OR5 irradiation hole were adjusted by the measured data. The fluxes of a fast neutron above 1 MeV were compared with calculated and measured value. Although the maximum difference was shown at 18.48%, most of the results showed good agreement. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1960-07-01

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

Calculation on maximum accumulation of Pu-239 and Pu-241 from aqueous homogeneous reactor

International Nuclear Information System (INIS)

Ikhlas H Siregar; Frida Agung R; Suharyana; Azizul Khakim; Dahman Siregar

2016-01-01

Calculations on maximum accumulation of Pu-239 and Pu-241 using MCNPX computer code with UO_2(NO_3)_2 fuel solution enriched by 19.75% operating at temperature 80°C have been conducted. AHR design was simulated with cylindrical core having diameter of 63.4 cm and 122 cm high. From this geometry we found the reactor was critical with density 108 gr U/L of UO_2(NO_3)_2 solution. The result showed that multiplication factor (k_e_f_f) of AHR was 1.05284. Then the burn up calculations were done for various time intervals from 5 days until 285 years to analyze the result. From calculation, it was found out that the saturated concentration of Pu-239 was reached after 40-50 years of operation, producing 1.23 x 102 gr and the activity 7.645 Ci. While for operate time of AHR to produce Pu-241 should under 80 years with mass 21.7 gr and the activity 2.247 x 103 Ci. The accumulations of both isotopes are considered to be small, having low potential for misusing them for producing nuclear weapon. (author)

Calculating the Maximum Density of the Surface Packing of Ions in Ionic Liquids

Science.gov (United States)

Kislenko, S. A.; Moroz, Yu. O.; Karu, K.; Ivaništšev, V. B.; Fedorov, M. V.

2018-05-01

The maximum density of monolayer packing on a graphene surface is calculated by means of molecular dynamics (MD) for ions of characteristic size and symmetry: 1-butyl-3-methylimidazolium [BMIM]+, tetrabutylammonium [TBA]+, tetrafluoroborate [BF4]-, dicyanamide [DCA]-, and bis(trifluoromethane) sulfonimide [TFSI]-. The characteristic orientations of ions in a closely packed monolayer are found. It is shown that the formation of a closely packed monolayer is possible for [DCA]- and [BF4]- anions only at surface charges that exceed the limit of the electrochemical stability of the corresponding ionic liquids. For the [TBA]+ cation, a monolayer structure can be observed at the charge of nearly 30 μC/cm2 attainable in electrochemical experiment.

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

Science.gov (United States)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube

International Nuclear Information System (INIS)

Boscary, J.; Association Euratom-CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

1997-03-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author)

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

International Nuclear Information System (INIS)

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

Science.gov (United States)

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

Monte Carlo calculation of standard graphite block

International Nuclear Information System (INIS)

Ljubenov, V.

2000-01-01

This paper presents results of calculation of neutron flux space and energy distribution in the standard graphite block (SGB) obtained by the MCNP TM code. VMCCS nuclear data library, based on the ENDF / B-VI release 4 evaluation file, is used. MCNP model of the SGB considers detailed material, geometric and spectral properties of the neutron source, source carrier, graphite moderator medium, aluminium foil holders and proximate surrounding of SGB Geometric model is organised to provide the simplest homogeneous volume cells in order to obtain the maximum acceleration of neutron history tracking (author)

Neutron point-flux calculation by Monte Carlo

International Nuclear Information System (INIS)

Eichhorn, M.

1986-04-01

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

PODESY program for flux mapping of CNA II reactor:

International Nuclear Information System (INIS)

Ribeiro Guevara, Sergio

1988-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Atmospheric neutrino fluxes

International Nuclear Information System (INIS)

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

1990-02-01

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

Increasing the neutron flux study for the TRR-II core design

International Nuclear Information System (INIS)

Chen, C.-H.; Yang, J.-T.; Chou, Y.-C.

1999-01-01

The maximum unperturbed thermal flux of the originally proposed core design, which is a 6x6 square arrangement with power level of 20 MW and has been presented at the 6th Meeting of IGORR, for the TRR-II reactor is about 2.0x10 14 n/cm 2 -sec. However, it is no longer satisfied the user's requirement, that is, it must reach at least 2.5x10 14 n/cm 2 -sec. In order to enhance the thermal neutron flux, one of the most effective ways is to increase the average power density. Therefore, two new designs with more compact cores are then proposed and studied. One is 5x6 rectangular arrangement with power of 20 MW; the other one is 5x5 square arrangement with power of 16 MW. It is for sure that both core designs can satisfy thermal hydraulic safety limits. The designed parameters related to neutronics are listed and compared fundamentally. According to our calculation, although both cores have similar average power density, the results show that the 5x6/20 MW design has the maximum unperturbed thermal flux in the D 2 O region about 2.7x10 14 n/cm 2 -sec, and the 5x5/16 MW design has 2.5x10 14 n/cm 2 -sec. The maximum thermal flux in the neighborhood of the longer side of the 5x6 core is about 7% higher than the one in the neighborhood of any side of the 5x5 core. This 'long-side effect' gives the 5x6/20 MW core design an advantage of the utilization of the thermal neutron flux in the D 2 O region. In addition, the 5x5 core is also more sensitive to the reactivity change on account of in-core irradiation test facilities. Therefore, under overall considerations the 5x6/20 MW core design is chosen for further detailed design. (author)

Calculation of thermal stress condition in long metal cylinder under heating by continuous laser radiation

International Nuclear Information System (INIS)

Uglov, A.A.; Uglov, S.A.; Kulik, A.N.

1997-01-01

The method of determination of temperature field and unduced thermal stresses in long metallic cylinder under its heating by cw-laser normally distributed heat flux is offered. The graphically presented results of calculation show the stress maximum is placed behind of center of laser heat sport along its movement line on the cylinder surface

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

DEFF Research Database (Denmark)

Richards, Meryl; Metzel, Ruth; Chirinda, Ngonidzashe

2016-01-01

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

Pollutant Flux Estimation in an Estuary Comparison between Model and Field Measurements

Directory of Open Access Journals (Sweden)

Yen-Chang Chen

2014-08-01

Full Text Available This study proposes a framework for estimating pollutant flux in an estuary. An efficient method is applied to estimate the flux of pollutants in an estuary. A gauging station network in the Danshui River estuary is established to measure the data of water quality and discharge based on the efficient method. A boat mounted with an acoustic Doppler profiler (ADP traverses the river along a preselected path that is normal to the streamflow to measure the velocities, water depths and water quality for calculating pollutant flux. To know the characteristics of the estuary and to provide the basis for the pollutant flux estimation model, data of complete tidal cycles is collected. The discharge estimation model applies the maximum velocity and water level to estimate mean velocity and cross-sectional area, respectively. Thus, the pollutant flux of the estuary can be easily computed as the product of the mean velocity, cross-sectional area and pollutant concentration. The good agreement between the observed and estimated pollutant flux of the Danshui River estuary shows that the pollutant measured by the conventional and the efficient methods are not fundamentally different. The proposed method is cost-effective and reliable. It can be used to estimate pollutant flux in an estuary accurately and efficiently.

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

Science.gov (United States)

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

2018-06-01

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

The Global Character of the Flux of Downward Longwave Radiation

Science.gov (United States)

Stephens, Graeme L.; Wild, Martin; Stackhouse, Paul W., Jr.; L'Ecuyer, Tristan; Kato, Seiji; Henderson, David S.

2012-01-01

Four different types of estimates of the surface downwelling longwave radiative flux (DLR) are reviewed. One group of estimates synthesizes global cloud, aerosol, and other information in a radiation model that is used to calculate fluxes. Because these synthesis fluxes have been assessed against observations, the global-mean values of these fluxes are deemed to be the most credible of the four different categories reviewed. The global, annual mean DLR lies between approximately 344 and 350 W/sq m with an error of approximately +/-10 W/sq m that arises mostly from the uncertainty in atmospheric state that governs the estimation of the clear-sky emission. The authors conclude that the DLR derived from global climate models are biased low by approximately 10 W/sq m and even larger differences are found with respect to reanalysis climate data. The DLR inferred from a surface energy balance closure is also substantially smaller that the range found from synthesis products suggesting that current depictions of surface energy balance also require revision. The effect of clouds on the DLR, largely facilitated by the new cloud base information from the CloudSat radar, is estimated to lie in the range from 24 to 34 W/sq m for the global cloud radiative effect (all-sky minus clear-sky DLR). This effect is strongly modulated by the underlying water vapor that gives rise to a maximum sensitivity of the DLR to cloud occurring in the colder drier regions of the planet. The bottom of atmosphere (BOA) cloud effect directly contrast the effect of clouds on the top of atmosphere (TOA) fluxes that is maximum in regions of deepest and coldest clouds in the moist tropics.

Neutron flux distribution forecasting device of reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1991-01-01

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

Specification of ROP flux shape

Energy Technology Data Exchange (ETDEWEB)

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

1997-06-01

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

Specification of ROP flux shape

International Nuclear Information System (INIS)

Min, Byung Joo; Gray, A.

1997-06-01

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

Studies in boiling heat transfer in two phase flow through tube arrays: nucleate boiling heat transfer coefficient and maximum heat flux as a function of velocity and quality of Freon-113

International Nuclear Information System (INIS)

Rahmani, R.

1983-01-01

The nucleate boiling heat-transfer coefficient and the maximum heat flux were studied experimentally as functions of velocity, quality and heater diameter for single-phase flow, and two-phase flow of Freon-113 (trichlorotrifluorethane). Results show: (1) peak heat flux: over 300 measured peak heat flux data from two 0.875-in. and four 0.625-in.-diameter heaters indicated that: (a) for pool boiling, single-phase and two-phase forced convection boiling the only parameter (among hysteresis, rate of power increase, aging, presence and proximity of unheated rods) that has a statistically significant effect on the peak heat flux is the velocity. (b) In the velocity range (0 0 position or the point of impact of the incident fluid) and the top (180 0 position) of the test element, respectively

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

Energy Technology Data Exchange (ETDEWEB)

Adams, C. H.

1976-07-01

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

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

International Nuclear Information System (INIS)

Adams, C.H.

1976-07-01

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

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

International Nuclear Information System (INIS)

Bobkov, V.P.

2000-01-01

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

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

Science.gov (United States)

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

2014-02-01

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

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

International Nuclear Information System (INIS)

Abit, K.

1984-01-01

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

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

Science.gov (United States)

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

2016-05-01

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

Critical heat flux in subcooled and low quality boiling

International Nuclear Information System (INIS)

Maroti, L.

1976-06-01

A semi-empirical relationship for critical heat flux prediction in a light water cooled power reactor core is developed. The method of developing this relationship is the extension of the analysis of pool boiling crisis for forced convective boiling. In the calculations the energy conservation equation is used together with additional condition for the crisis. Assuming that in the vicinity of the crisis the heat is transported only by the latent heat of the vapour this condition for the crisis can be characterized by the maximum departure velocity of the vapour. Because only flow boiling crisis associating with bubbling at the heating surface is considered the model could be applied only for low quality boiling crisis. The calculated results are compared to the available experimental ones. (Sz.N.Z.)

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

Science.gov (United States)

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

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

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

International Nuclear Information System (INIS)

Kawasaki, Hiromitsu; Seki, Yasushi

1983-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

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

Critical heat flux determination in an annulus section

International Nuclear Information System (INIS)

Reyes C, C.A.

1997-01-01

The present report explains the phenomenon of Critical heat flux. The study of this physical phenomenon is carried out during the boiling of a liquid and is of supreme importance for the calculation and operation of a nuclear reactor even in the moderns generators of steam (thermoelectric and nucleoelectrics), industrial cooling and in all those industrial process that use a liquid subject to sources of heating and to conditions of work excessively high (temperatures and pressures) so that stay in operation in an appropriate manner and sure. Once well-known this value, the equipment used in these process works with a maximum heat that is smaller than the Critical Heat Flux. The study of the Critical Heat Flux has achieved important advances in the last years, mainly for the enormous obligation that in this moment involved the safety to world level, this has forced to researchers and designers of this type of equipment to center their attention in the obtaining of a correlation which of general way explains it. In this reports two correlations will be compared that they contribute to the evaluation of the Critical Heat Flux in annulus and that they try to be generals in this type of geometry, the Shah correlation's and the Katto correlation's. The same as most of the correlations, these have been calculated so that the fluid of work is water, although they have also been proven with others fluids. The results obtained in this report only will show the degree of advance which the investigation of this phenomenon has achieved in annulus and to low amounts of flow of liquid, like which they are in the Experimental Heat Transfer Circuit located in the Department of Physics of the National Institute of Nuclear Research. (Author)

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

DEFF Research Database (Denmark)

Zhang, Xiaoxu; Liu, Xiao; Chen, Zhe

2015-01-01

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

Solar Flux Deposition And Heating Rates In Jupiter's Atmosphere

Science.gov (United States)

Perez-Hoyos, Santiago; Sánchez-Lavega, A.

2009-09-01

We discuss here the solar downward net flux in the 0.25 - 2.5 µm range in the atmosphere of Jupiter and the associated heating rates under a number of vertical cloud structure scenarios focusing in the effect of clouds and hazes. Our numerical model is based in the doubling-adding technique to solve the radiative transfer equation and it includes gas absorption by CH4, NH3 and H2, in addition to Rayleigh scattering by a mixture of H2 plus He. Four paradigmatic Jovian regions have been considered (hot-spots, belts, zones and Polar Regions). The hot-spots are the most transparent regions with downward net fluxes of 2.5±0.5 Wm-2 at the 6 bar level. The maximum solar heating is 0.04±0.01 K/day and occurs above 1 bar. Belts and zones characterization result in a maximum net downward flux of 0.5 Wm-2 at 2 bar and 0.015 Wm-2 at 6 bar. Heating is concentrated in the stratospheric and tropospheric hazes. Finally, Polar Regions are also explored and the results point to a considerable stratospheric heating of 0.04±0.02 K/day. In all, these calculations suggest that the role of the direct solar forcing in the Jovian atmospheric dynamics is limited to the upper 1 - 2 bar of the atmosphere except in the hot-spot areas. Acknowledgments: This work has been funded by Spanish MEC AYA2006-07735 with FEDER support and Grupos Gobierno Vasco IT-464-07.

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

International Nuclear Information System (INIS)

Patrakka, E.

1994-01-01

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

Effect of flux discontinuity on spatial approximations for discrete ordinates methods

International Nuclear Information System (INIS)

Duo, J.I.; Azmy, Y.Y.

2005-01-01

This work presents advances on error analysis of the spatial approximation of the discrete ordinates method for solving the neutron transport equation. Error norms for different non-collided flux problems over a two dimensional pure absorber medium are evaluated using three numerical methods. The problems are characterized by the incoming flux boundary conditions to obtain solutions with different level of differentiability. The three methods considered are the Diamond Difference (DD) method, the Arbitrarily High Order Transport method of the Nodal type (AHOT-N), and of the Characteristic type (AHOT-C). The last two methods are employed in constant, linear and quadratic orders of spatial approximation. The cell-wise error is computed as the difference between the cell-averaged flux computed by each method and the exact value, then the L 1 , L 2 , and L ∞ error norms are calculated. The results of this study demonstrate that the level of differentiability of the exact solution profoundly affects the rate of convergence of the numerical methods' solutions. Furthermore, in the case of discontinuous exact flux the methods fail to converge in the maximum error norm, or in the pointwise sense, in accordance with previous local error analysis. (authors)

Updated stomatal flux and flux-effect models for wheat for quantifying effects of ozone on grain yield, grain mass and protein yield.

Science.gov (United States)

Grünhage, Ludger; Pleijel, Håkan; Mills, Gina; Bender, Jürgen; Danielsson, Helena; Lehmann, Yvonne; Castell, Jean-Francois; Bethenod, Olivier

2012-06-01

Field measurements and open-top chamber experiments using nine current European winter wheat cultivars provided a data set that was used to revise and improve the parameterisation of a stomatal conductance model for wheat, including a revised value for maximum stomatal conductance and new functions for phenology and soil moisture. For the calculation of stomatal conductance for ozone a diffusivity ratio between O(3) and H(2)O in air of 0.663 was applied, based on a critical review of the literature. By applying the improved parameterisation for stomatal conductance, new flux-effect relationships for grain yield, grain mass and protein yield were developed for use in ozone risk assessments including effects on food security. An example of application of the flux model at the local scale in Germany shows that negative effects of ozone on wheat grain yield were likely each year and on protein yield in most years since the mid 1980s. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-03-01

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

High-heat-flux testing of helium-cooled heat exchangers for fusion applications

International Nuclear Information System (INIS)

Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

1996-01-01

High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

Burnout calculation

International Nuclear Information System (INIS)

Li, D.

1980-01-01

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

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

Science.gov (United States)

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

2010-09-01

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

Calculation of deuterium retention in, re-emission and reflection from a tungsten material under D+ ions irradiation with ACAT-DIFFUSE

International Nuclear Information System (INIS)

Ono, T.; Muramoto, T.; Kenmotsu, T.; Kawamura, T.

2008-08-01

We calculated, with a dynamic Monte Carlo code ACAT-DIFFUSE, fluxes of thermal D 2 re-emission, reflection and self-sputtering from a wrought tungsten material during a time sequence of 100 eV D + implantation, post-implanted isothermal out-gassing and thermal desorption spectroscopy. The obtained result agreed well with an existing experiment, where diffusion was considered in the calculations from the beginning of implantation. The three fluxes in the implantation period were shown to be almost comparable. The integrated deuterium flux released in the same period was estimated. The depth profiles of deuterium retained at 300 K in that period indicate that, while their maximum locations did not move, the profiles were broadened out because of fast diffusion. The amount of deuterium retained at 300 K was one order of magnitude higher than that at 473 K. (author)

A study of influence of material properties on magnetic flux density induced in magneto rheological damper through finite element analysis

Directory of Open Access Journals (Sweden)

Gurubasavaraju T. M.

2018-01-01

Full Text Available Magnetorheological fluids are smart materials, which are responsive to the external stimulus and changes their rheological properties. The damper performance (damping force is dependent on the magnetic flux density induced at the annular gap. Magnetic flux density developed at fluid flow gap of MR damper due to external applied current is also dependent on materials properties of components of MR damper (such as piston head, outer cylinder and piston rod. The present paper discus about the influence of different materials selected for components of the MR damper on magnetic effect using magnetostatic analysis. Different materials such as magnetic and low carbon steels are considered for piston head of the MR damper and magnetic flux density induced at fluid flow gap (filled with MR fluid is computed for different DC current applied to the electromagnetic coil. Developed magnetic flux is used for calculating the damper force using analytical method for each case. The low carbon steel has higher magnetic permeability hence maximum magnetic flux could pass through the piston head, which leads to higher value of magnetic effect induction at the annular gap. From the analysis results it is observed that the magnetic steel and low carbon steel piston head provided maximum magnetic flux density. Eventually the higher damping force can be observed for same case.

Calculating flux to predict future cave radon concentrations

Czech Academy of Sciences Publication Activity Database

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

2016-01-01

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

Uncertainty calculations made easier

International Nuclear Information System (INIS)

Hogenbirk, A.

1994-07-01

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

Methane Flux of Amazonian Peatland Ecosystems: Large Ecosystem Fluxes with Substantial Contribution from Palm (maritia Flexuosa) STEM Emissions

Science.gov (United States)

Van Haren, J. L. M.; Cadillo-Quiroz, H.

2015-12-01

Methane (CH4) emissions through plants have long been known in wetlands. However, most measurements have focused on stem tops and leaves. Recently, measurements at the lower parts of stems have shown that stem emissions can exceed soil CH4 emissions in Asian peatlands (Pangala et al. 2013). The addition of stem fluxes to soil fluxes for total ecosystem fluxes has the potential to bridge the discrepancy between modeled to measured and bottom-up to top-down flux estimates. Our measurements in peatlands of Peru show that especially Mauritia flexuosa, a palm species, can emit very large quantities of CH4, although most trees emitted at least some CH4. We used flexible stem chambers to adapt to stems of any size above 5cm in diameter. The chambers were sampled in closed loop with a Gasmet DX4015 for flux measurements, which lasted ~5 minutes after flushing with ambient air. We found that M. flexuosa stem fluxes decrease with height along the stem and were positively correlated with soil fluxes. Most likely CH4 is transported up the stem with the xylem water. Measured M. flexuosa stem fluxes below 1.5m averaged 11.2±1.5 mg-C m-2 h-1 (±95% CI) with a maximum of 123±3.5 mg-C m-2 h-1 (±SE), whereas soil fluxes averaged 6.7±1.7 mg-C m-2 h-1 (±95% CI) with a maximum of 31.6±0.4 mg-C m-2 h-1 (±SE). Significant CH4 fluxes were measured up to 5 m height along the stems. Combined with the high density of ~150 M. flexuosa individuals per hectare in these peatlands and the consistent diameter of ~30cm, the high flux rates add ~20% to the soil flux. With anywhere between 1 and 5 billion M. flexuosa stems across Amazon basin wetlands, stem fluxes from this palm species could represent a major addition to the overall Amazon basin CH4 flux.

Numerical Calculation of the Peaking Factor of a Water-Cooled W/Cu Monoblock for a Divertor

International Nuclear Information System (INIS)

Han Le; Chang Haiping; Zhang Jingyang; Xu Tiejun

2015-01-01

In order to accurately predict the incident critical heat flux (ICHF, the heat flux at the heated surface when CHF occurs) of a water-cooled W/Cu monoblock for a divertor, the exact knowledge of its peaking factors (f p ) under one-sided heating conditions with different design parameters is a key issue. In this paper, the heat conduction in the solid domain of a water-cooled W/Cu monoblock is calculated numerically by assuming the local heat transfer coefficients (HTC) of the cooling wall to be functions of the local wall temperature, so as to obtain f p . The reliability of the calculation method is validated by an experimental example result, with the maximum error of 2.1% only. The effects of geometric and flow parameters on the f p of a water-cooled W/Cu monoblock are investigated. Within the scope of this study, it is shown that the f p increases with increasing dimensionless W/Cu monoblock width and armour thickness (the shortest distance between the heated surface and Cu layer), and the maximum increases are 43.8% and 22.4% respectively. The dimensionless W/Cu monoblock height and Cu thickness have little effect on f p . The increase of Reynolds number and Jakob number causes the increase of f p , and the maximum increases are 6.8% and 9.6% respectively. Based on the calculated results, an empirical correlation on peaking factor is obtained via regression. These results provide a valuable reference for the thermal-hydraulic design of water-cooled divertors. (paper)

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

DEFF Research Database (Denmark)

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

2016-01-01

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

rf SQUID system as tunable flux qubit

Energy Technology Data Exchange (ETDEWEB)

Ruggiero, B. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy)]. E-mail: b.ruggiero@cib.na.cnr.it; Granata, C. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Vettoliere, A. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Rombetto, S. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Russo, R. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Russo, M. [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Corato, V. [Dipartimento di Ingegneria dell' Informazione, Seconda Universita di Napoli, I-81031 Aversa (Italy); Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy); Silvestrini, P. [Dipartimento di Ingegneria dell' Informazione, Seconda Universita di Napoli, I-81031 Aversa (Italy); Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I-80078 Pozzuoli (Italy)

2006-08-21

We present a fully integrated rf SQUID-based system as flux qubit with a high control of the flux transfer function of the superconducting transformer modulating the coupling between the flux qubit and the readout system. The control of the system is possible by including into the superconducting flux transformer a vertical two-Josephson-junctions interferometer (VJI) in which the Josephson current is precisely modulated from a maximum to zero by a transversal magnetic field parallel to the flux transformer plane. The proposed system can be also used in a more general configuration to control the off-diagonal terms in the Hamiltonian of the flux qubit and to turn on and off the coupling between two or more qubits.

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

International Nuclear Information System (INIS)

Godoy, William F.; DesJardin, Paul E.

2010-01-01

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

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

Directory of Open Access Journals (Sweden)

Barthelemy Mathieu

2014-01-01

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

Neutron flux measurements in C-9 capsule pressure tube

International Nuclear Information System (INIS)

Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.

2001-01-01

C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s

A Fourier analysis on the maximum acceptable grid size for discrete proton beam dose calculation

International Nuclear Information System (INIS)

Li, Haisen S.; Romeijn, H. Edwin; Dempsey, James F.

2006-01-01

We developed an analytical method for determining the maximum acceptable grid size for discrete dose calculation in proton therapy treatment plan optimization, so that the accuracy of the optimized dose distribution is guaranteed in the phase of dose sampling and the superfluous computational work is avoided. The accuracy of dose sampling was judged by the criterion that the continuous dose distribution could be reconstructed from the discrete dose within a 2% error limit. To keep the error caused by the discrete dose sampling under a 2% limit, the dose grid size cannot exceed a maximum acceptable value. The method was based on Fourier analysis and the Shannon-Nyquist sampling theorem as an extension of our previous analysis for photon beam intensity modulated radiation therapy [J. F. Dempsey, H. E. Romeijn, J. G. Li, D. A. Low, and J. R. Palta, Med. Phys. 32, 380-388 (2005)]. The proton beam model used for the analysis was a near mono-energetic (of width about 1% the incident energy) and monodirectional infinitesimal (nonintegrated) pencil beam in water medium. By monodirection, we mean that the proton particles are in the same direction before entering the water medium and the various scattering prior to entrance to water is not taken into account. In intensity modulated proton therapy, the elementary intensity modulation entity for proton therapy is either an infinitesimal or finite sized beamlet. Since a finite sized beamlet is the superposition of infinitesimal pencil beams, the result of the maximum acceptable grid size obtained with infinitesimal pencil beam also applies to finite sized beamlet. The analytic Bragg curve function proposed by Bortfeld [T. Bortfeld, Med. Phys. 24, 2024-2033 (1997)] was employed. The lateral profile was approximated by a depth dependent Gaussian distribution. The model included the spreads of the Bragg peak and the lateral profiles due to multiple Coulomb scattering. The dependence of the maximum acceptable dose grid size on the

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Science.gov (United States)

Töre, Candan; Ortego, Pedro

2005-01-01

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

Monte Carlo surface flux tallies

International Nuclear Information System (INIS)

Favorite, Jeffrey A.

2010-01-01

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

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

International Nuclear Information System (INIS)

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

1980-03-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Boosted Fast Flux Loop Alternative Cooling Assessment

Energy Technology Data Exchange (ETDEWEB)

Glen R. Longhurst; Donna Post Guillen; James R. Parry; Douglas L. Porter; Bruce W. Wallace

2007-08-01

The Gas Test Loop (GTL) Project was instituted to develop the means for conducting fast neutron irradiation tests in a domestic radiation facility. It made use of booster fuel to achieve the high neutron flux, a hafnium thermal neutron absorber to attain the high fast-to-thermal flux ratio, a mixed gas temperature control system for maintaining experiment temperatures, and a compressed gas cooling system to remove heat from the experiment capsules and the hafnium thermal neutron absorber. This GTL system was determined to provide a fast (E > 0.1 MeV) flux greater than 1.0E+15 n/cm2-s with a fast-to-thermal flux ratio in the vicinity of 40. However, the estimated system acquisition cost from earlier studies was deemed to be high. That cost was strongly influenced by the compressed gas cooling system for experiment heat removal. Designers were challenged to find a less expensive way to achieve the required cooling. This report documents the results of the investigation leading to an alternatively cooled configuration, referred to now as the Boosted Fast Flux Loop (BFFL). This configuration relies on a composite material comprised of hafnium aluminide (Al3Hf) in an aluminum matrix to transfer heat from the experiment to pressurized water cooling channels while at the same time providing absorption of thermal neutrons. Investigations into the performance this configuration might achieve showed that it should perform at least as well as its gas-cooled predecessor. Physics calculations indicated that the fast neutron flux averaged over the central 40 cm (16 inches) relative to ATR core mid-plane in irradiation spaces would be about 1.04E+15 n/cm2-s. The fast-to-thermal flux ratio would be in excess of 40. Further, the particular configuration of cooling channels was relatively unimportant compared with the total amount of water in the apparatus in determining performance. Thermal analyses conducted on a candidate configuration showed the design of the water coolant and

REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

International Nuclear Information System (INIS)

Boettcher, W.; Schmidt, E.

1969-01-01

1 - Nature of physical problem solved: REFLOS is a programme for the evaluation of fuel-loading schemes in heavy water moderated reactors. The problems involved in this study are: a) Burn-up calculation for the reactor cell. b) Determination of reactivity behaviour, power distribution, attainable burn-up for both the running-in period and the equilibrium of a 3-dimensional heterogeneous reactor model; investigation of radial fuel movement schemes. c) Evaluation of mass flows of heavy atoms through the reactor and fuel cycle costs for the running-in, the equilibrium, and the shut down of a power reactor. If the subroutine for treating the reactor cell were replaced by a suitable routine, other reactors with weakly absorbing moderators could be analyzed. 2 - Method of solution: Nuclear constants and isotopic compositions of the different fuels in the reactor are calculated by the cell-burn-up programme and tabulated as functions of the burn-up rate (MWD/T). Starting from a known state of the reactor, the 3-dimensional heterogeneous reactor programme (applying an extension of the technique of Feinberg and Galanin) calculates reactivity and neutron flux distribution using one thermal and one or two fast neutron groups. After a given irradiation time, the new state of the reactor is determined, and new nuclear constants are assigned to the various defined locations in the reactor. Reloading of fuel may occur if the prescribed life of the reactor is reached or if the effective multiplication factor or the power form factor falls below a specified level. The scheme of reloading to be carried out is specified by a load vector, giving the number of channels to be discharged, the kind of movement from one to another channel and the type of fresh fuel to be charged for each single reloading event. After having determined the core states characterizing the equilibrium period, and having decided the fuel reloading scheme for the running-in period of the reactor life, the fuel

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

NARCIS (Netherlands)

Velthof, G.L.; Oenema, O.

1995-01-01

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

Neutron Flux and Activation Calculations for a High Current Deuteron Accelerator

CERN Document Server

Coniglio, Angela; Sandri, Sandro

2005-01-01

Neutron analysis of the first Neutral Beam (NB) for the International Thermonuclear Experimental Reactor (ITER) was performed to provide the basis for the study of the following main aspects: personnel safety during normal operation and maintenance, radiation shielding design, transportability of the NB components in the European countries. The first ITER NB is a medium energy light particle accelerator. In the scenario considered for the calculation the accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The average beam current is 13.3 A. To assess neutron transport in the ITER NB structure a mathematical model of the components geometry was implemented into MCNP computer code (MCNP version 4c2. "Monte Carlo N-Particle Transport Code System." RSICC Computer Code Collection. June 2001). The neutron source definition was outlined considering both D-D and D-T neutron production. FISPACT code (R.A. Forrest, FISPACT-2003. EURATOM/UKAEA Fusion, December 2002) was used to assess neutron...

Validation of calculated tissue maximum ratio obtained from measured percentage depth dose (PPD) data for high energy photon beam ( 6 MV and 15 MV)

International Nuclear Information System (INIS)

Osei, J.E.

2014-07-01

During external beam radiotherapy treatments, high doses are delivered to the cancerous cell. Accuracy and precision of dose delivery are primary requirements for effective and efficiency in treatment. This leads to the consideration of treatment parameters such as percentage depth dose (PDD), tissue air ratio (TAR) and tissue phantom ratio (TPR), which show the dose distribution in the patient. Nevertheless, tissue air ratio (TAR) for treatment time calculation, calls for the need to measure in-air-dose rate. For lower energies, measurement is not a problem but for higher energies, in-air measurement is not attainable due to the large build-up material required for the measurement. Tissue maximum ratio (TMR) is the quantity required to replace tissue air ratio (TAR) for high energy photon beam. It is known that tissue maximum ratio (TMR) is an important dosimetric function in radiotherapy treatment. As the calculation methods used to determine tissue maximum ratio (TMR) from percentage depth dose (PDD) were derived by considering the differences between TMR and PDD such as geometry and field size, where phantom scatter or peak scatter factors are used to correct dosimetric variation due to field size difference. The purpose of this study is to examine the accuracy of calculated tissue maximum ratio (TMR) data with measured TMR values for 6 MV and 15 MV photon beam at Sweden Ghana Medical Centre. With the help of the Blue motorize water phantom and the Omni pro-Accept software, Pdd values from which TMRs are calculated were measured at 100 cm source-to-surface distance (SSD) for various square field sizes from 5x5 cm to 40x40 cm and depth of 1.5 cm to 25 cm for 6 MV and 15 MV x-ray beam. With the same field sizes, depths and energies, the TMR values were measured. The validity of the calculated data was determined by making a comparison with values measured experimentally at some selected field sizes and depths. The results show that; the reference depth of maximum

Dissecting Reactor Antineutrino Flux Calculations

Science.gov (United States)

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

2017-09-01

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

Global view of F-region electron density and temperature at solar maximum

International Nuclear Information System (INIS)

Brace, L.H.; Theis, R.F.; Hoegy, W.R.

1982-01-01

Dynamics Explorer-2 is permitting the first measurements of the global structure of the F-regions at very high levels of solar activity (S>200). Selected full orbits of Langmuir probe measurements of electron temperature, T/sub e/, and density, N/sub e/, are shown to illustrate this global structure and some of the ionospheric features that are the topic of other papers in this issue. The ionospheric thermal structure is of particular interest because T/sub e/ is a sensitive indicator of the coupling of magnetospheric energy into the upper atmosphere. A comparison of these heating effects with those observed at solar minimum shows that the magnetospheric sources are more important at solar maximum, as might have been expected. Heating at the cusp, the auroral oval and the plasma-pause is generally both greater and more variable. Electron cooling rate calculations employing low latitude measurements indicate that solar extreme ultraviolet heating of the F region at solar maximum is enhanced by a factor that is greater than the increase in solar flux. Some of this enhanced electron heating arises from the increase in electron heating efficiency at the higher N/sub e/ of solar maximum, but this appears insufficient to completely resolve the discrepancy

Two-dimensional sensitivity calculation code: SENSETWO

International Nuclear Information System (INIS)

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

1979-05-01

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

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

Science.gov (United States)

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

2017-12-01

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

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

Science.gov (United States)

Basiri, H.; Tavakoli-Anbaran, H.

2018-01-01

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

CrossRef Antiproton Flux, Antiproton-to-Proton Flux Ratio, and Properties of Elementary Particle Fluxes in Primary Cosmic Rays Measured with the Alpha Magnetic Spectrometer on the International Space Station

CERN Document Server

Aguilar, M; Alpat, B; Ambrosi, G; Arruda, L; Attig, N; Aupetit, S; Azzarello, P; Bachlechner, A; Barao, F; Barrau, A; Barrin, L; Bartoloni, A; Basara, L; Başeǧmez-du Pree, S; Battarbee, M; Battiston, R; Bazo, J; Becker, U; Behlmann, M; Beischer, B; Berdugo, J; Bertucci, B; Bindi, V; Boella, G; de Boer, W; Bollweg, K; Bonnivard, V; Borgia, B; Boschini, M  J; Bourquin, M; Bueno, E  F; Burger, J; Cadoux, F; Cai, X  D; Capell, M; Caroff, S; Casaus, J; Castellini, G; Cernuda, I; Cervelli, F; Chae, M  J; Chang, Y  H; Chen, A  I; Chen, G  M; Chen, H  S; Cheng, L; Chou, H  Y; Choumilov, E; Choutko, V; Chung, C  H; Clark, C; Clavero, R; Coignet, G; Consolandi, C; Contin, A; Corti, C; Coste, B; Creus, W; Crispoltoni, M; Cui, Z; Dai, Y  M; Delgado, C; Della Torre, S; Demirköz, M  B; Derome, L; Di Falco, S; Dimiccoli, F; Díaz, C; von Doetinchem, P; Dong, F; Donnini, F; Duranti, M; D'Urso, D; Egorov, A; Eline, A; Eronen, T; Feng, J; Fiandrini, E; Finch, E; Fisher, P; Formato, V; Galaktionov, Y; Gallucci, G; García, B; García-López, R  J; Gargiulo, C; Gast, H; Gebauer, I; Gervasi, M; Ghelfi, A; Giovacchini, F; Goglov, P; Gómez-Coral, D  M; Gong, J; Goy, C; Grabski, V; Grandi, D; Graziani, M; Guerri, I; Guo, K  H; Habiby, M; Haino, S; Han, K  C; He, Z  H; Heil, M; Hoffman, J; Hsieh, T  H; Huang, H; Huang, Z  C; Huh, C; Incagli, M; Ionica, M; Jang, W  Y; Jinchi, H; Kang, S  C; Kanishev, K; Kim, G  N; Kim, K  S; Kirn, Th; Konak, C; Kounina, O; Kounine, A; Koutsenko, V; Krafczyk, M  S; La Vacca, G; Laudi, E; Laurenti, G; Lazzizzera, I; Lebedev, A; Lee, H  T; Lee, S  C; Leluc, C; Li, H  S; Li, J  Q; Li, Q; Li, T  X; Li, W; Li, Z  H; Li, Z  Y; Lim, S; Lin, C  H; Lipari, P; Lippert, T; Liu, D; Liu, Hu; Lu, S  Q; Lu, Y  S; Luebelsmeyer, K; Luo, F; Luo, J  Z; Lv, S  S; Majka, R; Mañá, C; Marín, J; Martin, T; Martínez, G; Masi, N; Maurin, D; Menchaca-Rocha, A; Meng, Q; Mo, D  C; Morescalchi, L; Mott, P; Nelson, T; Ni, J  Q; Nikonov, N; Nozzoli, F; Nunes, P; Oliva, A; Orcinha, M; Palmonari, F; Palomares, C; Paniccia, M; Pauluzzi, M; Pensotti, S; Pereira, R; Picot-Clemente, N; Pilo, F; Pizzolotto, C; Plyaskin, V; Pohl, M; Poireau, V; Putze, A; Quadrani, L; Qi, X  M; Qin, X; Qu, Z  Y; Räihä, T; Rancoita, P  G; Rapin, D; Ricol, J  S; Rodríguez, I; Rosier-Lees, S; Rozhkov, A; Rozza, D; Sagdeev, R; Sandweiss, J; Saouter, P; Schael, S; Schmidt, S  M; Schulz von Dratzig, A; Schwering, G; Seo, E  S; Shan, B  S; Shi, J  Y; Siedenburg, T; Son, D; Song, J  W; Sun, W  H; Tacconi, M; Tang, X  W; Tang, Z  C; Tao, L; Tescaro, D; Ting, Samuel C  C; Ting, S  M; Tomassetti, N; Torsti, J; Türkoğlu, C; Urban, T; Vagelli, V; Valente, E; Vannini, C; Valtonen, E; Vázquez Acosta, M; Vecchi, M; Velasco, M; Vialle, J  P; Vitale, V; Vitillo, S; Wang, L  Q; Wang, N  H; Wang, Q  L; Wang, X; Wang, X  Q; Wang, Z  X; Wei, C  C; Weng, Z  L; Whitman, K; Wienkenhöver, J; Willenbrock, M; Wu, H; Wu, X; Xia, X; Xiong, R  Q; Xu, W; Yan, Q; Yang, J; Yang, M; Yang, Y; Yi, H; Yu, Y  J; Yu, Z  Q; Zeissler, S; Zhang, C; Zhang, J; Zhang, J  H; Zhang, S  D; Zhang, S  W; Zhang, Z; Zheng, Z  M; Zhu, Z  Q; Zhuang, H  L; Zhukov, V; Zichichi, A; Zimmermann, N; Zuccon, P

2016-01-01

A precision measurement by AMS of the antiproton flux and the antiproton-to-proton flux ratio in primary cosmic rays in the absolute rigidity range from 1 to 450 GV is presented based on 3.49×105 antiproton events and 2.42×109 proton events. The fluxes and flux ratios of charged elementary particles in cosmic rays are also presented. In the absolute rigidity range ∼60 to ∼500  GV, the antiproton p¯, proton p, and positron e+ fluxes are found to have nearly identical rigidity dependence and the electron e− flux exhibits a different rigidity dependence. Below 60 GV, the (p¯/p), (p¯/e+), and (p/e+) flux ratios each reaches a maximum. From ∼60 to ∼500  GV, the (p¯/p), (p¯/e+), and (p/e+) flux ratios show no rigidity dependence. These are new observations of the properties of elementary particles in the cosmos.

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

Directory of Open Access Journals (Sweden)

Dong Li

2018-05-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Calculations of shape and stability of menisci in Czochralski growth with tables to determine meniscus heights, maximum heights and capillary constants

International Nuclear Information System (INIS)

Uelhoff, W.; Mika, K.

1975-05-01

The shape and stability of menisci occurring during Czochralski growth have been studied by means of numerical methods for the case of the free surface. The existence of minimal joining angles is shown, beyond which the growing crystal will separate from the melt. The dependence of the interface height on the joining angle for different crystal diameters was calculated. The maximum stable heights and the corresponding joining angles were determined as a function of crystal diameter. A method for measuring the capillary constant of the melt during Czochralski growth is proposed. The results are compared with known analytical approximations. Limitations of the applications caused by a finite crucible radius or low g values are pointed out. For practical use the following functions have been tabulated: 1) meniscus height in dependence on joining angle and crystal radius, 2) the radius-height-ratio in dependence on radius and angle for the calculation of the capillary constant, 3) the maximum stable height and the corresponding growth angle as a function of crystal radius. (orig.) [de

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

International Nuclear Information System (INIS)

Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.

2010-01-01

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are

The study of heat flux for disruption on experimental advanced superconducting tokamak

Science.gov (United States)

Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

2016-05-01

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

Vermeeren, L.

2001-01-01

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

Application of the Bowring correlation for calculating the critical heat flux

International Nuclear Information System (INIS)

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

1986-01-01

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

Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry

International Nuclear Information System (INIS)

Yang, W.S.; Finck, P.J.; Khalil, H.S.

1990-01-01

A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs

Fast neutron fluxes distribution in Egyptian ilmenite concrete

International Nuclear Information System (INIS)

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

1978-01-01

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

Comparison of the 1D flux theory with a 2D hydrodynamic secondary settling tank model.

Science.gov (United States)

Ekama, G A; Marais, P

2004-01-01

The applicability of the 1D idealized flux theory (1DFT) for design of secondary settling tanks (SSTs) is evaluated by comparing its predicted maximum surface overflow (SOR) and solids loading (SLR) rates with that calculated from the 2D hydrodynamic model SettlerCAD using as a basis 35 full scale SST stress tests conducted on different SSTs with diameters from 30 to 45m and 2.25 to 4.1 m side water depth, with and without Stamford baffles. From the simulations, a relatively consistent pattern appeared, i.e. that the 1DFT can be used for design but its predicted maximum SLR needs to be reduced by an appropriate flux rating, the magnitude of which depends mainly on SST depth and hydraulic loading rate (HLR). Simulations of the sloping bottom shallow (1.5-2.5 m SWD) Dutch SSTs tested by STOWa and the Watts et al. SST, all with doubled SWDs, and the Darvill new (4.1 m) and old (2.5 m) SSTs with interchanged depths, were run to confirm the sensitivity of the flux rating to depth and HLR. Simulations with and without a Stamford baffle were also done. While the design of the internal features of the SST, such as baffling, have a marked influence on the effluent SS concentration for underloaded SSTs, these features appeared to have only a small influence on the flux rating, i.e. capacity, of the SST, In the meantime until more information is obtained, it would appear that from the simulations so far that the flux rating of 0.80 of the 1DFT maximum SLR recommended by Ekama and Marais remains a reasonable value to apply in the design of full scale SSTs--for deep SSTs (4 m SWD) the flux rating could be increased to 0.85 and for shallow SSTs (2.5 m SWD) decreased to 0.75. It is recommended that (i) while the apparent interrelationship between SST flux rating and depth suggests some optimization of the volume of the SST, that this be avoided and that (ii) the depth of the SST be designed independently of the surface area as is usually the practice and once selected, the

Flux and energy dependence of methane production from graphite due to H+ impact

International Nuclear Information System (INIS)

Davis, J.W.; Haasz, A.A.; Stangeby, P.C.

1986-06-01

Carbon is in widespread use for limiter surfaces, as well as first wall coatings in current tokamaks. Chemical erosion via methane formation, due to energetic H + impact, is expected to contribute to the total erosion rate of carbon from these surfaces. Experimental results are presented for the methane yield from pyrolytic graphite due to H + exposure, using a mass analyzed ion beam. H + energies of 0.1-3 keV and flux densities of ∼ 5x10 13 to l0 16 H + /cm 2 s were used. The measured methane yield (CH 4 /H + ) initially increases with flux density, then reaches a maximum, which is followed by a gradual decrease. The magnitude of the maximum yield and the flux density at which it occurs depends on the graphite temperature. The yields obtained at temperatures corresponding to yield maxima at specific flux densities also show an initial increase, followed by a shallow maximum and a gradual decrease as a function of flux density; the maximum occurs at ∼10 15 H + /cm 2 s. Also presented are results on the methane production dependence on ion energy over the range 0.1 to 3 keV, and graphite temperature dependence measurements

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

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Calculation and experimental study of the RBMK-1500 reactor emergency cooling at maximum designed accident

International Nuclear Information System (INIS)

Cherkashov, Yu.M.; Vasilevskij, V.P.; Labazov, V.H.; Loninov, A.Ya.; Molochnikov, Yu.S.; Novosel'skij, O.Yu.; Podlazov, L.N.; Pavlov, V.B.; Pushkarev, V.I.

1981-01-01

The analysis of thermohydraulic and neutron-physical processes occurring in the RBMK-1500 reactor during the reactor emergency cooling system triggering (RECS) after the maximum designed accident (MDA) is conducted. The MDA means hypothetical instant hilliotine break of the main circulating pump head collector. During the whole cooling down period the RECS should provide the temperature level of the fuel elements not exceeding 1200 deg C and the channel pipe temperature - 600 deg C. The principal flowsheet of the balloon type RECS is described. Calculations of the valve fast response effect on the RECS productivity are carried out. It is concluded that the chosen balloon RECS provides reliable temperature modes of fuel elements naand channel pipes under the MDA conditions. At the same time a momentary splash of neutron power by the value not more than 10% can take place [ru

Maximum Power from a Solar Panel

Directory of Open Access Journals (Sweden)

Michael Miller

2010-01-01

Full Text Available Solar energy has become a promising alternative to conventional fossil fuel sources. Solar panels are used to collect solar radiation and convert it into electricity. One of the techniques used to maximize the effectiveness of this energy alternative is to maximize the power output of the solar collector. In this project the maximum power is calculated by determining the voltage and the current of maximum power. These quantities are determined by finding the maximum value for the equation for power using differentiation. After the maximum values are found for each time of day, each individual quantity, voltage of maximum power, current of maximum power, and maximum power is plotted as a function of the time of day.

The power and robustness of maximum LOD score statistics.

Science.gov (United States)

Yoo, Y J; Mendell, N R

2008-07-01

The maximum LOD score statistic is extremely powerful for gene mapping when calculated using the correct genetic parameter value. When the mode of genetic transmission is unknown, the maximum of the LOD scores obtained using several genetic parameter values is reported. This latter statistic requires higher critical value than the maximum LOD score statistic calculated from a single genetic parameter value. In this paper, we compare the power of maximum LOD scores based on three fixed sets of genetic parameter values with the power of the LOD score obtained after maximizing over the entire range of genetic parameter values. We simulate family data under nine generating models. For generating models with non-zero phenocopy rates, LOD scores maximized over the entire range of genetic parameters yielded greater power than maximum LOD scores for fixed sets of parameter values with zero phenocopy rates. No maximum LOD score was consistently more powerful than the others for generating models with a zero phenocopy rate. The power loss of the LOD score maximized over the entire range of genetic parameters, relative to the maximum LOD score calculated using the correct genetic parameter value, appeared to be robust to the generating models.

Surface fluxes in heterogeneous landscape

Energy Technology Data Exchange (ETDEWEB)

Bay Hasager, C

1997-01-01

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

The Effects of Data Gaps on the Calculated Monthly Mean Maximum and Minimum Temperatures in the Continental United States: A Spatial and Temporal Study.

Science.gov (United States)

Stooksbury, David E.; Idso, Craig D.; Hubbard, Kenneth G.

1999-05-01

Gaps in otherwise regularly scheduled observations are often referred to as missing data. This paper explores the spatial and temporal impacts that data gaps in the recorded daily maximum and minimum temperatures have on the calculated monthly mean maximum and minimum temperatures. For this analysis 138 climate stations from the United States Historical Climatology Network Daily Temperature and Precipitation Data set were selected. The selected stations had no missing maximum or minimum temperature values during the period 1951-80. The monthly mean maximum and minimum temperatures were calculated for each station for each month. For each month 1-10 consecutive days of data from each station were randomly removed. This was performed 30 times for each simulated gap period. The spatial and temporal impact of the 1-10-day data gaps were compared. The influence of data gaps is most pronounced in the continental regions during the winter and least pronounced in the southeast during the summer. In the north central plains, 10-day data gaps during January produce a standard deviation value greater than 2°C about the `true' mean. In the southeast, 10-day data gaps in July produce a standard deviation value less than 0.5°C about the mean. The results of this study will be of value in climate variability and climate trend research as well as climate assessment and impact studies.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

Science.gov (United States)

Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

2014-08-01

The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

Using sonic anemometer temperature to measure sensible heat flux in strong winds

Directory of Open Access Journals (Sweden)

S. P. Burns

2012-09-01

Full Text Available Sonic anemometers simultaneously measure the turbulent fluctuations of vertical wind (w' and sonic temperature (T_s', and are commonly used to measure sensible heat flux (H. Our study examines 30-min heat fluxes measured with a Campbell Scientific CSAT3 sonic anemometer above a subalpine forest. We compared H calculated with T_s to H calculated with a co-located thermocouple and found that, for horizontal wind speed (U less than 8 m s⁻¹, the agreement was around ±30 W m⁻². However, for U ≈ 8 m s⁻¹, the CSAT H had a generally positive deviation from H calculated with the thermocouple, reaching a maximum difference of ≈250 W m⁻² at U ≈ 18 m s⁻¹. With version 4 of the CSAT firmware, we found significant underestimation of the speed of sound and thus T_s in high winds (due to a delayed detection of the sonic pulse, which resulted in the large CSAT heat flux errors. Although this T_s error is qualitatively similar to the well-known fundamental correction for the crosswind component, it is quantitatively different and directly related to the firmware estimation of the pulse arrival time. For a CSAT running version 3 of the firmware, there does not appear to be a significant underestimation of T_s; however, a T_s error similar to that of version 4 may occur if the CSAT is sufficiently out of calibration. An empirical correction to the CSAT heat flux that is consistent with our conceptual understanding of the T_s error is presented. Within a broader context, the surface energy balance is used to evaluate the heat flux measurements, and the usefulness of side-by-side instrument comparisons is discussed.

Unfolded ANO-1 fluxes using the LEPRICON methodology

International Nuclear Information System (INIS)

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

1985-01-01

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

Plasma crowbars in cylindrical flux compression experiments

International Nuclear Information System (INIS)

Suter, L.J.

1979-01-01

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

Numerical Calculation of the Peaking Factor of a Water-Cooled W/Cu Monoblock for a Divertor

Science.gov (United States)

Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

2015-09-01

In order to accurately predict the incident critical heat flux (ICHF, the heat flux at the heated surface when CHF occurs) of a water-cooled W/Cu monoblock for a divertor, the exact knowledge of its peaking factors (fp) under one-sided heating conditions with different design parameters is a key issue. In this paper, the heat conduction in the solid domain of a water-cooled W/Cu monoblock is calculated numerically by assuming the local heat transfer coefficients (HTC) of the cooling wall to be functions of the local wall temperature, so as to obtain fp. The reliability of the calculation method is validated by an experimental example result, with the maximum error of 2.1% only. The effects of geometric and flow parameters on the fp of a water-cooled W/Cu monoblock are investigated. Within the scope of this study, it is shown that the fp increases with increasing dimensionless W/Cu monoblock width and armour thickness (the shortest distance between the heated surface and Cu layer), and the maximum increases are 43.8% and 22.4% respectively. The dimensionless W/Cu monoblock height and Cu thickness have little effect on fp. The increase of Reynolds number and Jakob number causes the increase of fp, and the maximum increases are 6.8% and 9.6% respectively. Based on the calculated results, an empirical correlation on peaking factor is obtained via regression. These results provide a valuable reference for the thermal-hydraulic design of water-cooled divertors. supported by National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005) and Funding of Jiangsu Innovation Program for Graduate Education, China (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

Simplified Methodology to Estimate the Maximum Liquid Helium (LHe) Cryostat Pressure from a Vacuum Jacket Failure

Science.gov (United States)

Ungar, Eugene K.; Richards, W. Lance

2015-01-01

The aircraft-based Stratospheric Observatory for Infrared Astronomy (SOFIA) is a platform for multiple infrared astronomical observation experiments. These experiments carry sensors cooled to liquid helium temperatures. The liquid helium supply is contained in large (i.e., 10 liters or more) vacuum-insulated dewars. Should the dewar vacuum insulation fail, the inrushing air will condense and freeze on the dewar wall, resulting in a large heat flux on the dewar's contents. The heat flux results in a rise in pressure and the actuation of the dewar pressure relief system. A previous NASA Engineering and Safety Center (NESC) assessment provided recommendations for the wall heat flux that would be expected from a loss of vacuum and detailed an appropriate method to use in calculating the maximum pressure that would occur in a loss of vacuum event. This method involved building a detailed supercritical helium compressible flow thermal/fluid model of the vent stack and exercising the model over the appropriate range of parameters. The experimenters designing science instruments for SOFIA are not experts in compressible supercritical flows and do not generally have access to the thermal/fluid modeling packages that are required to build detailed models of the vent stacks. Therefore, the SOFIA Program engaged the NESC to develop a simplified methodology to estimate the maximum pressure in a liquid helium dewar after the loss of vacuum insulation. The method would allow the university-based science instrument development teams to conservatively determine the cryostat's vent neck sizing during preliminary design of new SOFIA Science Instruments. This report details the development of the simplified method, the method itself, and the limits of its applicability. The simplified methodology provides an estimate of the dewar pressure after a loss of vacuum insulation that can be used for the initial design of the liquid helium dewar vent stacks. However, since it is not an exact

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1977-01-01

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

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

Science.gov (United States)

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

2018-01-28

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

Measurement of the gamma ray flux between 50 and 350 GeV from the Mrk 501 Blazar with the experiment CELESTE

International Nuclear Information System (INIS)

Brion, E.

2005-10-01

The blazar Mrk 501 has a non-thermal emission spectrum with 2 components. The first one, located between radio waves and X-rays, is due to the synchrotron emission of the magnetized jet, while the second one, emitted in the high energy gamma-ray domain, is still not fully understood. Until 1999, this last domain had only been covered between 100 MeV and 4 GeV as well as above 300 GeV. This energy gap was filled by the creation of the CELESTE experiment, recording Cherenkov emission produced by gamma-rays between 50 and 350 GeV penetrating the atmosphere. Mrk 501, which has a variable emission, was observed in 2000 and 2001, and was detected in 2000. A flux has been calculated which constrains the high energy emission models, presented in this thesis. Crab nebula flux measurements validate the method since this source is the standard candle for atmospheric Cherenkov telescopes. Analysis cuts for Mrk 501 are determined using data from the blazar Mrk 421, which has nearly the same declination as Mrk 501. Finally, improved detector simulations were used to calculate the effective area of the instrument, taking the atmosphere quality into account, yielding the flux for Mrk 501 during observations taken between April and June 2000. This flux was compared with a synchrotron self-Compton emission model and with data taken in X-rays. It shows that Mrk 501 was slightly more active during this period compared to the remainder of the year and to the year 2001. A flux upper limit is calculated for other measurements. This is the first measurement in the energy range 50 - 350 GeV (this range represents the limits in energy for which the trigger rate, that is the convolution between the source spectrum and the effective area of the instrument, is higher than 20% of the trigger maximum). It helps to constrain the position of the inverse Compton emission maximum and tends to favor, in this particular case, X- and gamma-ray emission processes from 2 different electron populations

Analytic flux formulas and tables of shielding functions

International Nuclear Information System (INIS)

Wallace, O.J.

1981-06-01

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

Calculation of fluxes through a repository caused by a local well

International Nuclear Information System (INIS)

Thunvik, R.

1983-05-01

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

Evaluation of NASA's Carbon Monitoring System (CMS) Flux Pilot: Terrestrial CO2 Fluxes

Science.gov (United States)

Fisher, J. B.; Polhamus, A.; Bowman, K. W.; Collatz, G. J.; Potter, C. S.; Lee, M.; Liu, J.; Jung, M.; Reichstein, M.

2011-12-01

NASA's Carbon Monitoring System (CMS) flux pilot project combines NASA's Earth System models in land, ocean and atmosphere to track surface CO2 fluxes. The system is constrained by atmospheric measurements of XCO2 from the Japanese GOSAT satellite, giving a "big picture" view of total CO2 in Earth's atmosphere. Combining two land models (CASA-Ames and CASA-GFED), two ocean models (ECCO2 and NOBM) and two atmospheric chemistry and inversion models (GEOS-5 and GEOS-Chem), the system brings together the stand-alone component models of the Earth System, all of which are run diagnostically constrained by a multitude of other remotely sensed data. Here, we evaluate the biospheric land surface CO2 fluxes (i.e., net ecosystem exchange, NEE) as estimated from the atmospheric flux inversion. We compare against the prior bottom-up estimates (e.g., the CASA models) as well. Our evaluation dataset is the independently derived global wall-to-wall MPI-BGC product, which uses a machine learning algorithm and model tree ensemble to "scale-up" a network of in situ CO2 flux measurements from 253 globally-distributed sites in the FLUXNET network. The measurements are based on the eddy covariance method, which uses observations of co-varying fluxes of CO2 (and water and energy) from instruments on towers extending above ecosystem canopies; the towers integrate fluxes over large spatial areas (~1 km2). We present global maps of CO2 fluxes and differences between products, summaries of fluxes by TRANSCOM region, country, latitude, and biome type, and assess the time series, including timing of minimum and maximum fluxes. This evaluation shows both where the CMS is performing well, and where improvements should be directed in further work.

Solar maximum observatory

International Nuclear Information System (INIS)

Rust, D.M.

1984-01-01

The successful retrieval and repair of the Solar Maximum Mission (SMM) satellite by Shuttle astronauts in April 1984 permitted continuance of solar flare observations that began in 1980. The SMM carries a soft X ray polychromator, gamma ray, UV and hard X ray imaging spectrometers, a coronagraph/polarimeter and particle counters. The data gathered thus far indicated that electrical potentials of 25 MeV develop in flares within 2 sec of onset. X ray data show that flares are composed of compressed magnetic loops that have come too close together. Other data have been taken on mass ejection, impacts of electron beams and conduction fronts with the chromosphere and changes in the solar radiant flux due to sunspots. 13 references

Asymmetric flux generation and its relaxation in reversed field pinch

International Nuclear Information System (INIS)

Arimoto, H.; Masamune, S.; Nagata, A.

1985-02-01

The toroidally asymmetric flux enhancement [''dynamo effect''] and the axisymmetrization of the enhanced fluxes that follows in the setting up phase of Reversed Field Pinch are investigated on the STP-3[M] device. A rapid increase in the toroidal flux generated by the dynamo effect is first observed near the poloidal and toroidal current feeders. Then, this inhomogeneity of the flux propagates toroidally towards the plasma current. The axisymmetrization of the flux is attained just after the maximum of plasma current. The MHD activities decrease significantly after this axisymmetrization and the quiescent period is obtained. (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Improved HOR fuel management by flux measurement data feedback

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

Improved HOR fuel management by flux measurement data feedback

International Nuclear Information System (INIS)

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

1997-01-01

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

The study of heat flux for disruption on experimental advanced superconducting tokamak

International Nuclear Information System (INIS)

Yang, Zhendong; Fang, Jianan; Luo, Jiarong; Cui, Zhixue; Gong, Xianzu; Gan, Kaifu; Zhao, Hailin; Zhang, Bin; Chen, Meiwen

2016-01-01

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dR_s_e_p = −2 cm, while it changes to upper single null (dR_s_e_p = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m"2.

Solar neutrino flux at keV energies

Science.gov (United States)

Vitagliano, Edoardo; Redondo, Javier; Raffelt, Georg

2017-12-01

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

Predicting radon flux from uranium mill tailings

International Nuclear Information System (INIS)

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

1983-11-01

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

Foil activation detectors - some remarks on the choice of detectors, the adjustment of cross-sections and the unfolding of flux spectra

International Nuclear Information System (INIS)

McCracken, A.K.; Packwood, A.

1978-01-01

Neutron spectroscopy in a favourable environment can yield without supporting calculations a wealth of spectral detail which cannot be approached by the multiple foil analysis (MFA) method. On the other hand in hostile environments only MFA methods are available and they require validation and/or improvement by exposing them to comparison with other types of measurement and definitive calculation in tightly controlled test neutron spectra. This paper considers some problems related to MFA unfolding of flux spectra, systematic and random errors in detector measurements and the choice of detectors which will be of maximum use in all environments of current interest

Gradient heat flux measurement as monitoring method for the diesel engine

Science.gov (United States)

Sapozhnikov, S. Z.; Mityakov, V. Yu; Mityakov, A. V.; Vintsarevich, A. V.; Pavlov, A. V.; Nalyotov, I. D.

2017-11-01

The usage of gradient heat flux measurement for monitoring of heat flux on combustion chamber surface and optimization of diesel work process is proposed. Heterogeneous gradient heat flux sensors can be used at various regimes for an appreciable length of time. Fuel injection timing is set by the position of the maximum point on the angular heat flux diagram however, the value itself of the heat flux may not be considered. The development of such an approach can be productive for remote monitoring of work process in the cylinders of high-power marine engines.

Radiation dosimetry at the BNL High Flux Beam Reactor

International Nuclear Information System (INIS)

Holden, N.E.; Hu, J.P.; Reciniello, R.N.

1998-02-01

The HFBR is a heavy water, D 2 O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of 235 U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR's D 2 O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles

Biogenic emissions of volatile organic compounds from gorse (Ulex europaeus): Diurnal emission fluxes at Kelling Heath, England

Science.gov (United States)

Cao, X.-L.; Boissard, C.; Juan, A. J.; Hewitt, C. N.; Gallagher, M.

1997-08-01

Volatile organic compound (VOC) emission fluxes from Gorse (Ulex europaeus) were measured during May 30-31, 1995 at Kelling Heath in eastern England by using bag enclosure and gradient methods simultaneously. The enclosure measurements were made from branches at different stages of physiological development (flowering, after flowering, and mixed). Isoprene was found to represent 90% of the total VOC emissions, and its emission rates fluctuated from 6 ng (g dwt)-1 h-1 in the early morning to about 9700 ng(g dwt)-1 h-1 at midday. Averaged emission rates standardized to 20°C were 1625, 2120, and 3700 ng (g dwt)-1 h-1 for the new grown, "mixed," and flowering branch, respectively. Trans-ocimene and α-pinene were the main monoterpenes emitted and represented, on average, 47.6% and 36.9% of the total monoterpenes. Other monoterpenes, camphene, sabinene, β-pinene, myrcene, limonene and γ-terpinene, were positively identified but together represented less than 1.5% of the total VOC emissions from gorse. Maximum isoprene concentrations in air at the site were measured around midday at 2 m (174 parts per trillion by volume, or pptv) and 6 m (149 pptv), and minimum concentrations were measured during the night (8 pptv at both heights). Mean daytime α-pinene air concentrations of 141 and 60 pptv at 2 and 6 m height were determined, but trans-ocimene concentrations were less than the analytical detection limit (4 pptv), suggesting rapid chemical removal of this compound from air. The isoprene fluxes calculated by the micrometeorological gradient method showed a pattern similar to that of those calculated by the enclosure method, with isoprene emission rates maximum at midday (100 μg m-2 h-1) and not detectable during the nighttime. Assessment of the fraction of the site covered by gorse plants enabled an extrapolation of emission fluxes from the enclosure measurements. When averaged over the 2 day experiment, isoprene fluxes of 29.8 and 27.8 μg m-2 h-1 were obtained from

Iron fluxes to Talos Dome, Antarctica, over the past 200 kyr

Directory of Open Access Journals (Sweden)

P. Vallelonga

2013-03-01

Full Text Available Atmospheric fluxes of iron (Fe over the past 200 kyr are reported for the coastal Antarctic Talos Dome ice core, based on acid leachable Fe concentrations. Fluxes of Fe to Talos Dome were consistently greater than those at Dome C, with the greatest difference observed during interglacial climates. We observe different Fe flux trends at Dome C and Talos Dome during the deglaciation and early Holocene, attributed to a combination of deglacial activation of dust sources local to Talos Dome and the reorganisation of atmospheric transport pathways with the retreat of the Ross Sea ice shelf. This supports similar findings based on dust particle sizes and fluxes and Rare Earth Element fluxes. We show that Ca and Fe should not be used as quantitative proxies for mineral dust, as they all demonstrate different deglacial trends at Talos Dome and Dome C. Considering that a 20 ppmv decrease in atmospheric CO2 at the coldest part of the last glacial maximum occurs contemporaneously with the period of greatest Fe and dust flux to Antarctica, we confirm that the maximum contribution of aeolian dust deposition to Southern Ocean sequestration of atmospheric CO2 is approximately 20 ppmv.

Appropriateness of dynamical systems for the comparison of different embedding methods via calculation of the maximum Lyapunov exponent

International Nuclear Information System (INIS)

Franchi, M; Ricci, L

2014-01-01

The embedding of time series provides a valuable, and sometimes indispensable, tool in order to analyze the dynamical properties of a chaotic system. To this purpose, the choice of the embedding dimension and lag is decisive. The scientific literature describes several methods for selecting the most appropriate parameter pairs. Unfortunately, no conclusive criterion to decide which method – and thus which embedding pair – is the best has been so far devised. A widely employed quantity to compare different methods is the maximum Lyapunov exponent (MLE) because, for chaotic systems that have explicit analytic representations, MLE can be numerically evaluated independently of the embedding dimension and lag. Within this framework, we investigated the dependence on the calculated MLE on the embedding dimension and lag in the case of three dynamical systems that are also widespreadly used as reference systems, namely the Lorenz, Rössler and Mackey-Glass attractors. By also taking into account the statistical fluctuations of the calculated MLE, we propose a new method to assess which systems provide suitable test benches for the comparison of different embedding methods via MLE calculation. For example we found that, despite of its popularity in this scientific context, the Rössler attractor is not a reliable workbench to test the validity of an embedding method

One dimensional benchmark calculations using diffusion theory

International Nuclear Information System (INIS)

Ustun, G.; Turgut, M.H.

1986-01-01

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

Receiver function estimated by maximum entropy deconvolution

Institute of Scientific and Technical Information of China (English)

吴庆举; 田小波; 张乃铃; 李卫平; 曾融生

2003-01-01

Maximum entropy deconvolution is presented to estimate receiver function, with the maximum entropy as the rule to determine auto-correlation and cross-correlation functions. The Toeplitz equation and Levinson algorithm are used to calculate the iterative formula of error-predicting filter, and receiver function is then estimated. During extrapolation, reflective coefficient is always less than 1, which keeps maximum entropy deconvolution stable. The maximum entropy of the data outside window increases the resolution of receiver function. Both synthetic and real seismograms show that maximum entropy deconvolution is an effective method to measure receiver function in time-domain.

Vertical Josephson Interferometer for Tunable Flux Qubit

Energy Technology Data Exchange (ETDEWEB)

Granata, C [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Vettoliere, A [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Lisitskiy, M [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Rombetto, S [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Russo, M [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Ruggiero, B [Istituto di Cibernetica ' E. Caianiello' del Consiglio Nazionale delle Ricerche, I- 80078, Pozzuoli (Italy); Corato, V [Dipartimento di Ingegneria dell' Informazione, Seconda Universita di Napoli, I-8 1031, Aversa (Italy) and Istituto di Cibernetica ' E. Caianiello' del CNR, I-80078, Pozzuoli (Italy); Russo, R [Dipartimento di Ingegneria dell' Informazione, Seconda Universita di Napoli, I-8 1031, Aversa (Italy) and Istituto di Cibernetica ' E. Caianiello' del CNR, I-80078, Pozzuoli (Italy); Silvestrini, P [Dipartimento di Ingegneria dell' Informazione, Seconda Universita di Napoli, I-8 1031, Aversa (Italy) and Istituto di Cibernetica ' E. Caianiello' del CNR, I-80078, Pozzuoli (Italy)

2006-06-01

We present a niobium-based Josephson device as prototype for quantum computation with flux qubits. The most interesting feature of this device is the use of a Josephson vertical interferometer to tune the flux qubit allowing the control of the off-diagonal Hamiltonian terms of the system. In the vertical interferometer, the Josephson current is precisely modulated from a maximum to zero with fine control by a small transversal magnetic field parallel to the rf superconducting loop plane.

Actinide and Xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D.L.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and xenon reactivity effects in ATW high flux systems

International Nuclear Information System (INIS)

Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C.

1995-01-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

Actinide and Xenon reactivity effects in ATW high flux systems

Energy Technology Data Exchange (ETDEWEB)

Woosley, M. [Univ. of Virginia, Charlottesville, VA (United States); Olson, K.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)] [and others

1995-10-01

In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

Improved HOR fuel management by flux measurement data feedback

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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

International Nuclear Information System (INIS)

Clem, John R.

2011-01-01

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

Reactor calculation benchmark PCA blind test results

International Nuclear Information System (INIS)

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

1980-01-01

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

Reactor calculation benchmark PCA blind test results

Energy Technology Data Exchange (ETDEWEB)

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

1980-01-01

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

Comparison of calculational methods for liquid metal reactor shields

International Nuclear Information System (INIS)

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

1985-09-01

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

Rotating flux compressor for energy conversion

International Nuclear Information System (INIS)

Chowdhuri, P.; Linton, T.W.; Phillips, J.A.

1983-01-01

The rotating flux compressor (RFC) converts rotational kinetic energy into an electrical output pulse which would have higher energy than the electrical energy initially stored in the compressor. An RFC has been designed in which wedge-shaped rotor blades pass through the air gaps between successive turns of a solenoid, the stator. Magnetic flux is generated by pulsing the stator solenoids when the inductance is a maximum, i.e., when the flux fills the stator-solenoid volume. Connecting the solenoid across a load conserves the flux which is compressed within the small volume surrounding the stator periphery when the rotor blades cut into the free space between the stator plates, creating a minimum-inductance condition. The unique features of this design are: (1) no electrical connections (brushes) to the rotor; (2) no conventional windings; and (3) no maintenance. The device has been tested up to 5000 rpm of rotor speed

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

International Nuclear Information System (INIS)

Stepanek, J.

1980-01-01

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

BRIGITTE, Dose Rate and Heat Source and Energy Flux for Self-Absorbing Rods

International Nuclear Information System (INIS)

Jegu, M.; Clement, M.

1978-01-01

1 - Nature of physical problem solved: Calculation of dose rate, heat sources or energy flux. The sources are self-absorbing radioactive rods. The shielding consists of blocks of which the cross section can be defined. 2 - Method of solution: Exponential attenuation and build-up factor between source points and detector points. Source integration with error estimate. Automatic or controlled build-up with monitor print-out. 3 - Restrictions on the complexity of the problem: Number of energy points, regions, detector points, abscissa points of the rod, vertical position of the rod, are all limited to ten. The maximum total number of vertical steps is 124

Extreme Maximum Land Surface Temperatures.

Science.gov (United States)

Garratt, J. R.

1992-09-01

There are numerous reports in the literature of observations of land surface temperatures. Some of these, almost all made in situ, reveal maximum values in the 50°-70°C range, with a few, made in desert regions, near 80°C. Consideration of a simplified form of the surface energy balance equation, utilizing likely upper values of absorbed shortwave flux (1000 W m2) and screen air temperature (55°C), that surface temperatures in the vicinity of 90°-100°C may occur for dry, darkish soils of low thermal conductivity (0.1-0.2 W m1 K1). Numerical simulations confirm this and suggest that temperature gradients in the first few centimeters of soil may reach 0.5°-1°C mm1 under these extreme conditions. The study bears upon the intrinsic interest of identifying extreme maximum temperatures and yields interesting information regarding the comfort zone of animals (including man).

Seasonal and interannual variability in deep ocean particle fluxes at the Oceanic Flux Program (OFP)/Bermuda Atlantic Time Series (BATS) site in the western Sargasso Sea near Bermuda

Science.gov (United States)

Conte, Maureen H.; Ralph, Nate; Ross, Edith H.

Since 1978, the Oceanic Flux Program (OFP) time-series sediment traps have measured particle fluxes in the deep Sargasso Sea near Bermuda. There is currently a 20+yr flux record at 3200-m depth, a 12+yr flux at 1500-m depth, and a 9+yr record at 500-m depth. Strong seasonality is observed in mass flux at all depths, with a flux maximum in February-March and a smaller maximum in December-January. There is also significant interannual variability in the flux, especially with respect to the presence/absence of the December-January flux maximum and in the duration of the high flux period in the spring. The flux records at the three depths are surprisingly coherent, with no statistically significant temporal lag between 500 and 3200-m fluxes at our biweekly sample resolution. Bulk compositional data indicate an extremely rapid decrease in the flux of organic constituents with depth between 500 and 1500-m, and a smaller decrease with depth between 1500 and 3200-m depth. In contrast, carbonate flux is uniform or increases slightly between 500 and 1500-m, possibly reflecting deep secondary calcification by foraminifera. The lithogenic flux increases by over 50% between 500 and 3200-m depth, indicating strong deep water scavenging/repackaging of suspended lithogenic material. Concurrent with the rapid changes in flux composition, there is a marked reduction in the heterogeneity of the sinking particle pool with depth, especially within the mesopelagic zone. By 3200-m depth, the bulk composition of the sinking particle pool is strikingly uniform, both seasonally and over variations in mass flux of more than an order of magnitude. These OFP results provide strong indirect evidence for the intensity of reprocessing of the particle pool by resident zooplankton within mesopelagic and bathypelagic waters. The rapid loss of organic components, the marked reduction in the heterogeneity of the bulk composition of the flux, and the increase in terrigenous fluxes with depth are most

Assessing the applicability of the 1D flux theory to full-scale secondary settling tank design with a 2D hydrodynamic model.

Science.gov (United States)

Ekama, G A; Marais, P

2004-02-01

The applicability of the one-dimensional idealized flux theory (1DFT) for the design of secondary settling tanks (SSTs) is evaluated by comparing its predicted maximum surface overflow (SOR) and solids loading (SLR) rates with that calculated with the two-dimensional computational fluid dynamics model SettlerCAD using as a basis 35 full-scale SST stress tests conducted on different SSTs with diameters from 30 to 45m and 2.25-4.1m side water depth (SWD), with and without Stamford baffles. From the simulations, a relatively consistent pattern appeared, i.e. that the 1DFT can be used for design but its predicted maximum SLR needs to be reduced by an appropriate flux rating, the magnitude of which depends mainly on SST depth and hydraulic loading rate (HLR). Simulations of the Watts et al. (Water Res. 30(9)(1996)2112) SST, with doubled SWDs and the Darvill new (4.1m) and old (2.5m) SSTs with interchanged depths, were run to confirm the sensitivity of the flux rating to depth and HLR. Simulations with and without a Stamford baffle were also performed. While the design of the internal features of the SST, such as baffling, has a marked influence on the effluent SS concentration while the SST is underloaded, these features appeared to have only a small influence on the flux rating, i.e. capacity, of the SST. Until more information is obtained, it would appear from the simulations that the flux rating of 0.80 of the 1DFT maximum SLR recommended by Ekama and Marais (Water Pollut. Control 85(1)(1986)101) remains a reasonable value to apply in the design of full-scale SSTs-for deep SSTs (4m SWD) the flux rating could be increased to 0.85 and for shallow SSTs (2.5m SWD) decreased to 0.75. It is recommended that (i) while the apparent interrelationship between SST flux rating and depth suggests some optimization of the volume of the SST, this be avoided and (ii) the depth of the SST be designed independently of the surface area as is usually the practice and once selected, the

Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

Directory of Open Access Journals (Sweden)

Yang Seong Woo

2016-01-01

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

Concentrations and fluxes of isoprene and oxygenated VOCs at a French Mediterranean oak forest

International Nuclear Information System (INIS)

Kalogridis, C.; Gros, V.; Sarda-Esteve, R.; Bonsang, B.; Bonnaire, N.; Boissard, C.; Baisnee, D.; Lathiere, J.

2014-01-01

The CANOPEE project aims to better understand the biosphere-atmosphere exchanges of biogenic volatile organic compounds (BVOCs) in the case of Mediterranean ecosystems and the impact of in-canopy processes on the atmospheric chemical composition above the canopy. Based on an intensive field campaign, the objective of our work was to determine the chemical composition of the air inside a canopy as well as the net fluxes of reactive species between the canopy and the boundary layer. Measurements were carried out during spring 2012 at the field site of the Oak Observatory of the Observatoire de Haute Provence (O3HP) located in the southeast of France. The site is a forest ecosystem dominated by downy oak, Quercus pubescens Willd., a typical Mediterranean species which features large isoprene emission rates. Mixing ratios of isoprene, its degradation products methylvinylketone (MVK) and methacrolein (MACR) and several other oxygenated VOC (OxVOC) were measured above the canopy using an online proton transfer reaction mass spectrometer (PTR-MS), and fluxes were calculated by the disjunct eddy covariance approach. The O3HP site was found to be a very significant source of isoprene emissions, with daily maximum ambient concentrations ranging between 2-16 ppbv inside and 2-5 ppbv just above the top of the forest canopy. Significant isoprene fluxes were observed only during daytime, following diurnal cycles with midday net emission fluxes from the canopy ranging between 2.0 and 9.7 mgm -2 h -1 . Net isoprene normalized flux (at 30 C, 1000 μmol quantam -2 s -1 ) was estimated at 7.4 mgm -2 h -1 . Evidence of direct emission of methanol was also found exhibiting maximum daytime fluxes ranging between 0.2 and 0.6 mgm -2 h -1 , whereas flux values for monoterpenes and others OxVOC such as acetone and acetaldehyde were below the detection limit. The MVK+MACR-to-isoprene ratio provided useful information on the oxidation of isoprene, and is in agreement with recent findings

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1976-01-01

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

Flux-limited diffusion models in radiation hydrodynamics

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Pawel, R.E.

1988-01-01

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

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

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

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

Atmospheric neutrino fluxes

International Nuclear Information System (INIS)

Perkins, D.H.

1984-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Munkejord, Svend Tollak

2006-05-11

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

Flux measurement in ZBR at the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Dauke, M.

2005-01-01

The determination of the neutron flux in the TRIGA-2-Vienna reactor was the objective of this research. The theory of the method (4π-β detectors) is presented as well as the determination of the maximum flux, gold-cadmium differential measurement, cobalt-wire measurement, finally a comparison of all results was made and interpreted. (nevyjel)

CONDOR: neutronic code for fuel elements calculation with rods

International Nuclear Information System (INIS)

Villarino, E.A.

1990-01-01

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

Atmosphere–Surface Fluxes of CO2 using Spectral Techniques

DEFF Research Database (Denmark)

Sørensen, Lise Lotte; Larsen, Søren Ejling

2010-01-01

Different flux estimation techniques are compared here in order to evaluate air–sea exchange measurement methods used on moving platforms. Techniques using power spectra and cospectra to estimate fluxes are presented and applied to measurements of wind speed and sensible heat, latent heat and CO2...... fluxes. Momentum and scalar fluxes are calculated from the dissipation technique utilizing the inertial subrange of the power spectra and from estimation of the cospectral amplitude, and both flux estimates are compared to covariance derived fluxes. It is shown how even data having a poor signal......-to-noise ratio can be used for flux estimations....

Flow-excursion-induced dryout at low-heat-flux

International Nuclear Information System (INIS)

Khatib-Rahbar, M.; Cazzoli, E.G.

1983-01-01

Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures

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

Quantifying fluxes and characterizing compositional changes of dissolved organic matter in aquatic systems in situ using combined acoustic and optical measurements

Science.gov (United States)

Downing, B.D.; Boss, E.; Bergamaschi, B.A.; Fleck, J.A.; Lionberger, M.A.; Ganju, N.K.; Schoellhamer, D.H.; Fujii, R.

2009-01-01

Studying the dynamics and geochemical behavior of dissolved and particulate organic material is difficult because concentration and composition may rapidly change in response to aperiodic as well as periodic physical and biological forcing. Here we describe a method useful for quantifying fluxes and analyzing dissolved organic matter (DOM) dynamics. The method uses coupled optical and acoustic measurements that provide robust quantitative estimates of concentrations and constituent characteristics needed to investigate processes and calculate fluxes of DOM in tidal and other lotic environments. Data were collected several times per hour for 2 weeks or more, with the frequency and duration limited only by power consumption and data storage capacity. We assessed the capabilities and limitations of the method using data from a winter deployment in a natural tidal wetland of the San Francisco Bay estuary. We used statistical correlation of in situ optical data with traditional laboratory analyses of discrete water samples to calibrate optical properties suited as proxies for DOM concentrations and characterizations. Coupled with measurements of flow velocity, we calculated long-term residual horizontal fluxes of DOC into and out from a tidal wetland. Subsampling the dataset provides an estimate for the maximum sampling interval beyond which the error in flux estimate is significantly increased.?? 2009, by the American Society of Limnology and Oceanography, Inc.

Notes on neutron flux measurement

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

The main purpose of this work is to get an useful guide to carry out topical neutron flux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutron fluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutron fluxes, which are used in neutron calculations. (Author) 16 refs

Tabular method of critical heat flux description in square packing rod bundles

International Nuclear Information System (INIS)

Bobkov, V.P.; Smogalev, I.P.

2003-01-01

Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

Unconventional application of the two-flux approximation for the calculation of the Ambartsumyan-Chandrasekhar function and the angular spectrum of the backward-scattered radiation for a semi-infinite isotropically scattering medium

Science.gov (United States)

Remizovich, V. S.

2010-06-01

It is commonly accepted that the Schwarzschild-Schuster two-flux approximation (1905, 1914) can be employed only for the calculation of the energy characteristics of the radiation field (energy density and energy flux density) and cannot be used to characterize the angular distribution of radiation field. However, such an inference is not valid. In several cases, one can calculate the radiation intensity inside matter and the reflected radiation with the aid of this simplest approximation in the transport theory. In this work, we use the results of the simplest one-parameter variant of the two-flux approximation to calculate the angular distribution (reflection function) of the radiation reflected by a semi-infinite isotropically scattering dissipative medium when a relatively broad beam is incident on the medium at an arbitrary angle relative to the surface. We do not employ the invariance principle and demonstrate that the reflection function exhibits the multiplicative property. It can be represented as a product of three functions: the reflection function corresponding to the single scattering and two identical h functions, which have the same physical meaning as the Ambartsumyan-Chandrasekhar function ( H) has. This circumstance allows a relatively easy derivation of simple analytical expressions for the H function, total reflectance, and reflection function. We can easily determine the relative contribution of the true single scattering in the photon backscattering at an arbitrary probability of photon survival Λ. We compare all of the parameters of the backscattered radiation with the data resulting from the calculations using the exact theory of Ambartsumyan, Chandrasekhar, et al., which was developed decades after the two-flux approximation. Thus, we avoid the application of fine mathematical methods (the Wiener-Hopf method, the Case method of singular functions, etc.) and obtain simple analytical expressions for the parameters of the scattered radiation

Spectral maximum entropy hydrodynamics of fermionic radiation: a three-moment system for one-dimensional flows

International Nuclear Information System (INIS)

Banach, Zbigniew; Larecki, Wieslaw

2013-01-01

The spectral formulation of the nine-moment radiation hydrodynamics resulting from using the Boltzmann entropy maximization procedure is considered. The analysis is restricted to the one-dimensional flows of a gas of massless fermions. The objective of the paper is to demonstrate that, for such flows, the spectral nine-moment maximum entropy hydrodynamics of fermionic radiation is not a purely formal theory. We first determine the domains of admissible values of the spectral moments and of the Lagrange multipliers corresponding to them. We then prove the existence of a solution to the constrained entropy optimization problem. Due to the strict concavity of the entropy functional defined on the space of distribution functions, there exists a one-to-one correspondence between the Lagrange multipliers and the moments. The maximum entropy closure of moment equations results in the symmetric conservative system of first-order partial differential equations for the Lagrange multipliers. However, this system can be transformed into the equivalent system of conservation equations for the moments. These two systems are consistent with the additional conservation equation interpreted as the balance of entropy. Exploiting the above facts, we arrive at the differential relations satisfied by the entropy function and the additional function required to close the system of moment equations. We refer to this additional function as the moment closure function. In general, the moment closure and entropy–entropy flux functions cannot be explicitly calculated in terms of the moments determining the state of a gas. Therefore, we develop a perturbation method of calculating these functions. Some additional analytical (and also numerical) results are obtained, assuming that the maximum entropy distribution function tends to the Maxwell–Boltzmann limit. (paper)

Statistical mechanics of flux lines in high-temperature superconductors

International Nuclear Information System (INIS)

Dasgupta, C.

1992-01-01

The shortness of the low temperature coherence lengths of high T c materials leads to new mechanisms of pinning of flux lines. Lattice periodic modulations of the order parameters itself acts to pin vortex lines in regions of the unit cell were the order parameter is small. A presentation of flux creep and flux noise at low temperature and magnetic fields in terms of motion of simple metastable defects on flux lines is made, with a calculation of flux lattice melting. 12 refs

Cool seafloor hydrothermal springs reveal global geochemical fluxes

Science.gov (United States)

Wheat, C. Geoffrey; Fisher, Andrew T.; McManus, James; Hulme, Samuel M.; Orcutt, Beth N.

2017-10-01

We present geochemical data from the first samples of spring fluids from Dorado Outcrop, a basaltic edifice on 23 M.y. old seafloor of the Cocos Plate, eastern Pacific Ocean. These samples were collected from the discharge of a cool hydrothermal system (CHS) on a ridge flank, where typical reaction temperatures in the volcanic crust are low (2-20 °C) and fluid residence times are short. Ridge-flank hydrothermal systems extract 25% of Earth's lithospheric heat, with a global discharge rate equivalent to that of Earth's river discharge to the ocean; CHSs comprise a significant fraction of this global flow. Upper crustal temperatures around Dorado Outcrop are ∼15 °C, the calculated residence time is V, U, Mg, phosphate, Si and Li are different. Applying these observed differences to calculated global CHS fluxes results in chemical fluxes for these ions that are ≥15% of riverine fluxes. Fluxes of K and B also may be significant, but better analytical resolution is required to confirm this result. Spring fluids also have ∼50% less dissolved oxygen (DO) than bottom seawater. Calculations of an analytical model suggest that the loss of DO occurs primarily (>80%) within the upper basaltic crust by biotic and/or abiotic consumption. This calculation demonstrates that permeable pathways within the upper crust can support oxic water-rock interactions for millions of years.

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

Research on cooling of ultra high critical heat flux with external flow boiling of water. Challenge to achieve ultra high critical heat flux and improvement in estimation of critical heat flux. JAERI's nuclear research promotion program, H11-004 (Contract research)

International Nuclear Information System (INIS)

Monde, Masanori; Mitsutake, Yuichi; Ishida, Kenji; Hino, Ryutaro

2003-03-01

An ultra high critical heat flux (CHF) has been challenged with a highly subcooled water jet impinging on a small rectangular heated surface. Major objective of the study is to achieve an ultra high heat flux cooling as large as 100 MW/m 2 and to establish an accurate estimation method of the CHF. The experiments were carried out over the experimental range; a fixed jet diameter of 2 mm, jet velocity of 5 - 35 m/s, degree of subcooling of 80 - 170 K and system pressure of 0.1 - 1.0 MPa. The rectangular heated surface with a thin nickel foil of 0.03 - 0.3 mm in thickness, 5 and 10 mm in length, and 4 mm in width and heated by a direct current. Effects of thickness of heater wall, jet velocity and subcooling on the CHF were experimentally elucidated. The experimental results show that the CHF decreases about 50% as the heater thickness, namely heat capacity of heater decreases. Characteristics of the CHF with heater length of 10 mm are correlated within ±20% by the generalized correlation of subcooled CHF proposed by the authors. However, the CHF with the shorter heater length of 5 mm shows large deviation of -40% especially at lower subcooling and higher velocity. The maximum CHF of 212 MW/m 2 was achieved at the subcooling of 151 K, the jet velocity of 35 m/s and system pressure of 0.5 MPa. The maximum CHF under atmospheric pressure approaches to 48% of the ultimate maximum heat flux given by the assumptions that vapor molecules leave a liquid-vapor interface at the average speed of a Boltzman-Maxwellian gas and any molecules returning to the interface are not permitted. The ratio of the CHF and ultimate maximum heat flux was considerably enhanced from the existing record of 30%. This study can give the feasibility of ultra high heat flux removal facing in a development of components such as a diverter of a fusion reactor. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1966-07-01

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

Approximation for maximum pressure calculation in containment of PWR reactors

International Nuclear Information System (INIS)

Souza, A.L. de

1989-01-01

A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

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)

Variable Eddington factors and flux-limiting diffusion coefficients

International Nuclear Information System (INIS)

Whalen, P.P.

1982-01-01

Variable Eddington factors and flux limiting diffusion coefficients arise in two common techniques of closing the moment equations of transport. The first two moment equations of the full transport equation are still frequently used to solve many problems of radiative or particle transport. An approximate analysis, developed by Levermore, exhibits the relation between the coefficients of the two different techniques. This analysis is described and then used to test the validity of several commonly used flux limiters and Eddington factors. All of the ad-hoc flux limiters have limited validity. All of the variable Eddington factors derived from some underlying description of the angular distribution function are generally valid. The use of coefficients from Minerbo's elegant maximum entropy Eddington factor analysis is suggested for use in either flux limited diffusion or variable Eddington factor equations

Flux at a point in MCNP

International Nuclear Information System (INIS)

Cashwell, E.D.; Schrandt, R.G.

1980-01-01

The current state of the art of calculating flux at a point with MCNP is discussed. Various techniques are touched upon, but the main emphasis is on the fast improved version of the once-more-collided flux estimator, which has been modified to treat neutrons thermalized by the free gas model. The method is tested on several problems on interest and the results are presented

Total maximum allocated load calculation of nitrogen pollutants by linking a 3D biogeochemical-hydrodynamic model with a programming model in Bohai Sea

Science.gov (United States)

Dai, Aiquan; Li, Keqiang; Ding, Dongsheng; Li, Yan; Liang, Shengkang; Li, Yanbin; Su, Ying; Wang, Xiulin

2015-12-01

The equal percent removal (EPR) method, in which pollutant reduction ratio was set as the same in all administrative regions, failed to satisfy the requirement for water quality improvement in the Bohai Sea. Such requirement was imposed by the developed Coastal Pollution Total Load Control Management. The total maximum allocated load (TMAL) of nitrogen pollutants in the sea-sink source regions (SSRs) around the Bohai Rim, which is the maximum pollutant load of every outlet under the limitation of water quality criteria, was estimated by optimization-simulation method (OSM) combined with loop approximation calculation. In OSM, water quality is simulated using a water quality model and pollutant load is calculated with a programming model. The effect of changes in pollutant loads on TMAL was discussed. Results showed that the TMAL of nitrogen pollutants in 34 SSRs was 1.49×105 ton/year. The highest TMAL was observed in summer, whereas the lowest in winter. TMAL was also higher in the Bohai Strait and central Bohai Sea and lower in the inner area of the Liaodong Bay, Bohai Bay and Laizhou Bay. In loop approximation calculation, the TMAL obtained was considered satisfactory for water quality criteria as fluctuation of concentration response matrix with pollutant loads was eliminated. Results of numerical experiment further showed that water quality improved faster and were more evident under TMAL input than that when using the EPR method

Marginal thermal-neutron peak fluxes in systems with modulation of reactivity

International Nuclear Information System (INIS)

Alekseev, N.I.; Stolypin, V.S.

1978-01-01

A possibility of obtaining high (including marginal) thermal neutron peak fluxes PHIsub(m) in a light water trap of a pulsed fast reactor with modulation of reactivity has been studied. The dependences of sub(m) on the subcriticality and supercriticality as well as on the supercritical state duration have been calculated on stepped variations of the reactivity. The calculations show that PHIsub(m) of about 7.3x10 18 neutron/cm 2 xs with the effective pulse duration of approximately 150 μc, pulse frequency of approximately 1 Hz and at fuel temperature of approximately 1300 deg C can be obtained with the reactor. The comparative calculations show that sub(m) is 1.5 times higher than that of a booster obtained using a ''meson plant'' (designs of the booster and the reactor are equivalent). The neutron background between pulses in the reactor is much lower than in the booster, and there is no need for a power injector in the reactor altogether. Meanwhile the maximum attainable PHIsub(m) for the booster and the reactor are the same and equal approximately 2x10 19 neutron/cm 2 xs

Implementação computacional do modelo carga-fluxo de carga-fluxo de dipolo para cálculo e interpretação das intensidades do espectro infravermelho Computational implementation of the model charge-charge flux-dipole flux for calculation and analysis of infrared intensities

Directory of Open Access Journals (Sweden)

Thiago C. F. Gomes

2008-01-01

Full Text Available The first computational implementation that automates the procedures involved in the calculation of infrared intensities using the charge-charge flux-dipole flux model is presented. The atomic charges and dipoles from the Quantum Theory of Atoms in Molecules (QTAIM model was programmed for Morphy98, Gaussian98 and Gaussian03 programs outputs, but for the ChelpG parameters only the Gaussian programs are supported. Results of illustrative but new calculations for the water, ammonia and methane molecules at the MP2/6-311++G(3d,3p theoretical level, using the ChelpG and QTAIM/Morphy charges and dipoles are presented. These results showed excellent agreement with analytical results obtained directly at the MP2/6-311++G(3d,3p level of theory.

Neutrino fluxes produced by high energy solar flare particles

International Nuclear Information System (INIS)

Kolomeets, E.V.; Shmonin, V.L.

1975-01-01

In this work the calculated differential energy spectra of neutrinos poduced by high energy protons accelerated during 'small' solar flares are presented. The muon flux produced by neutrino interactions with the matter at large depths under the ground is calculated. The obtained flux of muons for the total number of solar flare accelerated protons of 10 28 - 10 32 is within 10 9 - 10 13 particles/cm 2 X s x ster. (orig.) [de

Monitoring to assess progress toward meeting the Assabet River, Massachusetts, phosphorus total maximum daily load - Aquatic macrophyte biomass and sediment-phosphorus flux

Science.gov (United States)

Zimmerman, Marc J.; Qian, Yu; Yong Q., Tian

2011-01-01

In 2004, the Total Maximum Daily Load (TMDL) for Total Phosphorus in the Assabet River, Massachusetts, was approved by the U.S. Environmental Protection Agency. The goal of the TMDL was to decrease the concentrations of the nutrient phosphorus to mitigate some of the instream ecological effects of eutrophication on the river; these effects were, for the most part, direct consequences of the excessive growth of aquatic macrophytes. The primary instrument effecting lower concentrations of phosphorus was to be strict control of phosphorus releases from four major wastewatertreatment plants in Westborough, Marlborough, Hudson, and Maynard, Massachusetts. The improvements to be achieved from implementing this control were lower concentrations of total and dissolved phosphorus in the river, a 50-percent reduction in aquatic-plant biomass, a 30-percent reduction in episodes of dissolved oxygen supersaturation, no low-flow dissolved oxygen concentrations less than 5.0 milligrams per liter, and a 90-percent reduction in sediment releases of phosphorus to the overlying water. In 2007, the U.S. Geological Survey, in cooperation with the Massachusetts Department of Environmental Protection, initiated studies to evaluate conditions in the Assabet River prior to the upgrading of wastewater-treatment plants to remove more phosphorus from their effluents. The studies, completed in 2008, implemented a visual monitoring plan to evaluate the extent and biomass of the floating macrophyte Lemna minor (commonly known as lesser duckweed) in five impoundments and evaluated the potential for phosphorus flux from sediments in impounded and free-flowing reaches of the river. Hydrologically, the two study years 2007 and 2008 were quite different. In 2007, summer streamflows, although low, were higher than average, and in 2008, the flows were generally higher than in 2007. Visually, the effects of these streamflow differences on the distribution of Lemna were obvious. In 2007, large amounts of

The Net Carbon Flux due to Deforestation and Forest Re-Growth in the Brazilian Amazon: Analysis using a Process-Based Model

Science.gov (United States)

Hirsch, A. I.; Little, W. S.; Houghton, R. A.; Scott, N. A.; White, J. D.

2004-01-01

We developed a process-based model of forest growth, carbon cycling, and land cover dynamics named CARLUC (for CARbon and Land Use Change) to estimate the size of terrestrial carbon pools in terra firme (non-flooded) forests across the Brazilian Legal Amazon and the net flux of carbon resulting from forest disturbance and forest recovery from disturbance. Our goal in building the model was to construct a relatively simple ecosystem model that would respond to soil and climatic heterogeneity that allows us to study of the impact of Amazonian deforestation, selective logging, and accidental fire on the global carbon cycle. This paper focuses on the net flux caused by deforestation and forest re-growth over the period from 1970-1998. We calculate that the net flux to the atmosphere during this period reached a maximum of approx. 0.35 PgC/yr (1PgC = 1 x 10(exp I5) gC) in 1990, with a cumulative release of approx. 7 PgC from 1970- 1998. The net flux is higher than predicted by an earlier study by a total of 1 PgC over the period 1989-1 998 mainly because CARLUC predicts relatively high mature forest carbon storage compared to the datasets used in the earlier study. Incorporating the dynamics of litter and soil carbon pools into the model increases the cumulative net flux by approx. 1 PgC from 1970-1998, while different assumptions about land cover dynamics only caused small changes. The uncertainty of the net flux, calculated with a Monte-Carlo approach, is roughly 35% of the mean value (1 SD).

From COS ecosystem fluxes to GPP: integrating soil, branch and ecosystem fluxes.

Science.gov (United States)

Kooijmans, L.; Maseyk, K. S.; Vesala, T.; Mammarella, I.; Baker, I. T.; Seibt, U.; Sun, W.; Aalto, J.; Franchin, A.; Kolari, P.; Keskinen, H.; Levula, J.; Chen, H.

2016-12-01

The close coupling of Carbonyl Sulfide (COS) and CO2 due to a similar uptake pathway into plant stomata makes COS a promising new tracer that can potentially be used to partition the Net Ecosystem Exchange into gross primary production (GPP) and respiration. Although ecosystem-scale measurements have been made at several sites, the contribution of different ecosystem components to the total COS budget is often unknown. Besides that, the average Leaf Relative Uptake (LRU) ratio needs to be better determined to accurately translate COS ecosystem fluxes into GPP estimates when the simple linear correlation between GPP estimates and COS plant uptake is used. We performed two campaigns in the summer of 2015 and 2016 at the SMEAR II site in Hyytiälä, Finland to provide better constrained COS flux data for boreal forests. A combination of COS measurements were made during both years, i.e. atmospheric profile concentrations up to 125 m, eddy-covariance fluxes and soil chamber fluxes. In addition to these, branch chamber measurements were done in 2016 in an attempt to observe the LRU throughout the whole season. The LRU ratio shows an exponential correlation with photosynthetic active radiation (PAR) but is constant for PAR levels above 500 µmol m-2 s-1. Mid-day LRU values are 1.0 (aspen) and 1.5 (pine). The correlation between LRU and PAR can be explained by the fact that COS is hydrolyzed with the presence of the enzyme carbonic anhydrase, and is not light dependent, whereas the photosynthetic uptake of CO2 is. We observed nighttime fluxes on the order of 25-30 % of the daily maximum COS uptake. Soils are a small sink of COS and contribute to 3 % of the total ecosystem COS flux during daytime. In a comparison between observed and simulated fluxes from the Simple Biosphere (SiB) model, the modelled COS and CO2 ecosystem fluxes are on average 40 % smaller than the observed fluxes, however, the Ecosystem Relative Uptake (ERU) ratios are identical at a value of 1.9 ± 0

Adjusted neutron spectra of STEK cores for reactivity calculations

International Nuclear Information System (INIS)

Dekker, J.W.M.; Dragt, J.B.; Janssen, A.J.; Heijboer, R.J.; Klippel, H.Th.

1978-02-01

Neutron flux and adjoint flux spectra form a pre-requisite in the analysis of reactivity worth data measured in the STEK facility. First, a survey of all available information about these spectra is given. Next a special application of a general adjustment method is described. This method has been used to obtain adjusted STEK group flux and adjoint flux spectra, starting from calculated spectra. These theoretical spectra were adjusted to reactivity worths of natural boron (nat. B) and 235 U as well as a number of fission reaction rates. As a by-product in this adjustment calculation adjusted fission group cross sections of 235 U were obtained. The results, viz. group fluxes and adjoint fluxes and adjusted fission cross sections of 235 U are given. They have been used for the interpretation of fission product reactivity worth measurements made in STEK

Atmospheric dry deposition fluxes of trace elements measured in Bursa, Turkey

International Nuclear Information System (INIS)

Tasdemir, Yuecel; Kural, Can

2005-01-01

Trace element dry deposition fluxes were measured using a smooth, greased, knife-edge surrogate surface (KSS) holding greased Mylar strips in Bursa, Turkey. Sampling program was conducted between October 2002 and June 2003 and 46 dry deposition samples were collected. The average fluxes of crustal metals (Mg, Ca, and Fe) were one to four orders of magnitude higher than the fluxes of anthropogenic metals. Trace element fluxes ranged from 3 (Cd) to 24 230 (Ca) μg m -2 d -1 . The average trace element dry deposition fluxes measured in this study were similar to those measured in other urban areas. In addition, ambient air samples were also collected simultaneously with flux samples and concentrations of trace elements, collected with a TSP sampler, were between 0.7 and 4900 ng m -3 for Cd and Ca, respectively. The overall trace element dry deposition velocities, calculated by dividing the fluxes to the particle phase concentrations ranged from 2.3±1.7 cm s -1 (Pb) to 11.1±6.4 cm s -1 (Ni). These values are in good agreement with the values calculated using similar techniques. The anthropogenic and crustal contributions were estimated by employing enrichment factors (EFs) calculated relative to the average crustal composition. Low EFs for dry deposition samples were calculated. This is probably due to contamination of local dust and its important contribution to the collected samples. - Mechanical turbulence has an important influence on re-suspension and dry deposition of trace elements in an urban area

Theory and application of maximum magnetic energy in toroidal plasmas

International Nuclear Information System (INIS)

Chu, T.K.

1992-02-01

The magnetic energy in an inductively driven steady-state toroidal plasma is a maximum for a given rate of dissipation of energy (Poynting flux). A purely resistive steady state of the piecewise force-free configuration, however, cannot exist, as the periodic removal of the excess poloidal flux and pressure, due to heating, ruptures the static equilibrium of the partitioning rational surfaces intermittently. The rupture necessitates a plasma with a negative q'/q (as in reverse field pinches and spheromaks) to have the same α in all its force-free regions and with a positive q'/q (as in tokamaks) to have centrally peaked α's

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

Energy Technology Data Exchange (ETDEWEB)

Smith, M. A. [TerraPower, Bellevue, WA (United States); Lee, C. H. [TerraPower, Bellevue, WA (United States); Hill, R. N. [TerraPower, Bellevue, WA (United States)

2017-06-28

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.

Wide range neutron flux monitor

International Nuclear Information System (INIS)

Endo, Yorimasa; Fukushima, Toshiki.

1983-01-01

Purpose: To provide a wide range neutron-flux monitor adapted such that the flux monitoring function and alarming function can automatically by shifted from pulse counting system to cambel method system. Constitution: A wide range neutron-flux monitor comprises (la) pulse counting system and (lb) cambel-method system for inputting detection signals from neutron detectors and separating them into signals for the pulse measuring system and the cambel measuring system, (2) overlap detection and calculation circuit for detecting the existence of the overlap of two output signals from the (la) and (lb) systems, and (3) trip circuit for judging the abnormal state of neutron detectors upon input of the detection signals. (Seki, T.)

SR 97 - Radionuclide transport calculations

Energy Technology Data Exchange (ETDEWEB)

Lindgren, Maria [Kemakta Konsult AB, Stockholm (Sweden); Lindstroem, Fredrik [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

1999-12-01

An essential component of a safety assessment is to calculate radionuclide release and dose consequences for different scenarios and cases. The SKB tools for such a quantitative assessment are used to calculate the maximum releases and doses for the hypothetical repository sites Aberg, Beberg and Ceberg for the initial canister defect scenario and also for the glacial melting case for Aberg. The reasonable cases, i.e. all parameters take reasonable values, results in maximum biosphere doses of 5x10{sup -8} Sv/yr for Aberg, 3x10{sup -8} Sv/yr for Beberg and 1x10{sup -8} Sv/yr for Ceberg for peat area. These doses lie significantly below 0.15 mSv/yr. (A dose of 0.15 mSv/yr for unit probability corresponds to the risk limit of 10{sup -5} per year for the most exposed individuals recommended in regulations.) The conclusion that the maximum risk would lie well below 10{sup -5} per year is also demonstrated by results from the probabilistic calculations, which directly assess the resulting risk by combining dose and probability estimates. The analyses indicate that the risk is 2x10{sup -5} Sv/yr for Aberg, 8x10{sup -7} Sv/yr for Beberg and 3x10{sup -8} Sv/yr for Ceberg. The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. The influence from varying one parameter never changes the doses as much as an order of magnitude. In the far field the most important uncertainties affecting release and retention are associated with permeability and connectivity of the fractures in the rock. These properties affect several parameters. Highly permeable and well connected fractures imply high groundwater fluxes and short groundwater travel times. Sparsely connected or highly variable fracture properties implies low flow wetted surface along migration paths. It should, however, be remembered that the far-field parameters have little importance if the near-field parameters take their reasonable

SR 97 - Radionuclide transport calculations

International Nuclear Information System (INIS)

Lindgren, Maria; Lindstroem, Fredrik

1999-12-01

An essential component of a safety assessment is to calculate radionuclide release and dose consequences for different scenarios and cases. The SKB tools for such a quantitative assessment are used to calculate the maximum releases and doses for the hypothetical repository sites Aberg, Beberg and Ceberg for the initial canister defect scenario and also for the glacial melting case for Aberg. The reasonable cases, i.e. all parameters take reasonable values, results in maximum biosphere doses of 5x10 -8 Sv/yr for Aberg, 3x10 -8 Sv/yr for Beberg and 1x10 -8 Sv/yr for Ceberg for peat area. These doses lie significantly below 0.15 mSv/yr. (A dose of 0.15 mSv/yr for unit probability corresponds to the risk limit of 10 -5 per year for the most exposed individuals recommended in regulations.) The conclusion that the maximum risk would lie well below 10 -5 per year is also demonstrated by results from the probabilistic calculations, which directly assess the resulting risk by combining dose and probability estimates. The analyses indicate that the risk is 2x10 -5 Sv/yr for Aberg, 8x10 -7 Sv/yr for Beberg and 3x10 -8 Sv/yr for Ceberg. The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. The influence from varying one parameter never changes the doses as much as an order of magnitude. In the far field the most important uncertainties affecting release and retention are associated with permeability and connectivity of the fractures in the rock. These properties affect several parameters. Highly permeable and well connected fractures imply high groundwater fluxes and short groundwater travel times. Sparsely connected or highly variable fracture properties implies low flow wetted surface along migration paths. It should, however, be remembered that the far-field parameters have little importance if the near-field parameters take their reasonable values. In that case almost all

Calculation of the fine neutron flux distribution in the multi-zone reactor cell by spherical harmonics method; Odredjivanje fine raspodele fluksa u viseregionalnoj celiji reaktora primenom metode sfernih harmonika

Energy Technology Data Exchange (ETDEWEB)

Jockovic, M; Stefanovic, D; Barucija, M [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-07-01

This presentation contains development and application of method for calculating the neutron flux distribution by P{sub 3} approximation. A computer codes was developed for the ZUSE-Z-23 computer. Some typical results are included.

Rocket measurement of auroral electron fluxes associated with field-aligned currents

International Nuclear Information System (INIS)

Pazich, P.M.; Anderson, H.R.

1975-01-01

A Nike-Tomahawk rocket was instrumented with a vector magnetometer and an array of particle detectors including an electron and proton energyspectrometer covering the energy range 0.5-20 keV in seven fixed intervals and measuring the pitch angle distribution from 0degree to 180degree as the rocket spun. The payload was launched from Poker Flat, Alaska, at 0722 UT on February 25, 1972, over a bright auroral band that evidently was the poleward electron aurora, beyond the trapping boundary. An upper limit to the measured proton flux was 10 6 /cm 2 s sr keV. The energy spectrum of the electron flux measured during passage over the visible aurora always exhibited a peak within the measured energy range. During passage over the brighter auroral forms the peak shifted from approx.3 to approx.10 keV, the pitch angle distribution became peaked along B, and the intensity increased. Maximum fluxes of approx.3times10 8 el/cm 2 s sr keV were seen over the aurora, which reached approx.60 kR of lambda5577. The electron flux in regions of maximum flux tended to be the most field-aligned in the energy interval showing the highest intensity

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1957-07-01

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

On standard forms for transport equations and fluxes: Part 2

International Nuclear Information System (INIS)

Ross, D.W.

1990-03-01

Quasilinear expressions for anomalous particle and energy fluxes arising from electrostatic plasma turbulence in a tokamak are reviewed yet again. Further clarifications are made, and the position taken in a previous report is modified. There, the total energy flux, Q j , and the conductive heat flux, q j , were correctly defined, and the anomalous Q j was correctly calculated. It was shown that the anomalous energy transport can be correctly described by ∇·Q* j , where Q* j = 3/5 Q j , with all remaining source terms such as left-angle p j ∇·Vj} cancelling. Here, a revised discussion is given of the identification of the anomalous conductive flux, q j , in which the distinction between Q j and Q* j is reconsidered. It is shown that there is more than one consistent way to define q j . Transport calculations involving only theoretical electrostatic turbulent fluxes are unaffected by these distinctions since Q j or Q* j , rather than q j , is the quantity naturally calculated in the theory. However, an ambiguity remains in experimental transport analysis if the measured particle flux Γ j = n j V j is to be used in the energy equation. This is because we cannot be sure how properly to treat the source terms p j ∇·V j or { p j ∇·V j }. 17 refs

Self-adjointness of the fast flux in a pressurized water reactor

International Nuclear Information System (INIS)

Mosteller, R.D.

1985-01-01

Most computer codes for the analysis of systems transients rely on a simplified representation of the active core, typically employing either a one-dimensional or a point kinetics model. The collapsing of neutronics data from multidimensional steady-state calculations normally employs flux/flux-adjoint weighting. The multidimensional calculations, however, usually are performed only for the forward problem, not the adjoint. The collapsing methodologies employed in generating the neutronics input for transient codes typically construct adjoint fluxes from the assumption that the fast flux is self-adjoint. Until now, no further verification of this assumption has been undertaken for thermal reactors. As part of the verification effort for EPRI's reactor analysis support package, the validity of this assumption now has been investigated for a modern pressurized water reactor (PWR). The PDQ-7 code was employed to perform two-group fine-mesh forward and adjoint calculations for a two-dimensional representation of Zion Unit 2 at beginning of life, based on the standard PWR ARMP model. It has been verified that the fast flux is very nearly self-adjoint in a PWR. However, a significant error can arise during the subsequent construction of the thermal adjoint flux unless allowance is made for the difference between the forward and adjoint thermal buckling terms. When such a difference is included, the thermal adjoint flux can be estimated very accurately

Flux-pinning-induced stresses in a hollow superconducting cylinder with flux creep and viscosity properties

International Nuclear Information System (INIS)

Feng, W.J.; Gao, S.W.

2014-01-01

Highlights: • Magnetoelastic problem for a superconducting cylinder with a hole is investigated. • The effects of both flux creep and viscous flux flow on stresses are analyzed. • For the FC case, the maximal hoop tensile stress always occurs at hole edge. • For the ZFC case, the maximal hoop stress is not certain to occur at hole edge. - Abstract: The magnetoelastic problem for a superconducting cylinder with a concentric hole placed in a magnetic field is investigated, where the flux creep and viscous flux flow have been considered. The stress distributions are derived and numerical calculated for the descending field in both the zero-field cooling (ZFC) and field cooling (FC) processes. The effects of applied magnetic field, flux creep and viscous flux flow on the maximal radial and hoop stresses are discussed in detail, and some novel phenomena are found. Among others, for the FC case, the maximal hoop tensile stress always occurs at the hole edge, whist for the ZFC case, the maximal stresses including both hoop and radial stresses either occur in the vicinity of the hole or occur at the position of flux frontier in the remagnetization process. For the descending field, in general, both the flux creep and viscosity parameters have important effects on the maximal radial and hoop stresses. All these phenomena are perhaps of vital importance for the application of superconductors

Simplified magnetic circuit for the calculation of the stray magnetic flux through the shell gaps

Energy Technology Data Exchange (ETDEWEB)

Collarin, P.; Piovan, R. [Associazioni EURATOM-ENEA-CNR-Univ. di Padova (Italy). Gruppo di Padova per Ricerche sulla Fusione

1995-12-31

Significant toroidal magnetic field perturbations, stray flux at the shell gaps and current mismatching in the coils of the toroidal field winding are measured during the start-up and the flat-top phases of RFX. These phenomena are consistent with large and wall locked MHD modes: at first some m = 1 modes evolve separately one after the other, afterwards they concur to a wide and localized plasma perturbation that persists during the flat-top. These perturbations are heavily influenced by the stray magnetic flux through the shell gaps. Hence a magnetic circuit that mainly considers the magnetic reluctance of the conducting shell gaps has been developed in order to estimate this stray flux and, therefore, to evaluate the stabilizing capability of the shell. The observation of the MHD modes, the description of the equivalent magnetic network, the estimation of the stray flux and the comparison with the experimental measurements are reported in the paper.

Simplified magnetic circuit for the calculation of the stray magnetic flux through the shell gaps

International Nuclear Information System (INIS)

Collarin, P.; Piovan, R.

1995-01-01

Significant toroidal magnetic field perturbations, stray flux at the shell gaps and current mismatching in the coils of the toroidal field winding are measured during the start-up and the flat-top phases of RFX. These phenomena are consistent with large and wall locked MHD modes: at first some m = 1 modes evolve separately one after the other, afterwards they concur to a wide and localized plasma perturbation that persists during the flat-top. These perturbations are heavily influenced by the stray magnetic flux through the shell gaps. Hence a magnetic circuit that mainly considers the magnetic reluctance of the conducting shell gaps has been developed in order to estimate this stray flux and, therefore, to evaluate the stabilizing capability of the shell. The observation of the MHD modes, the description of the equivalent magnetic network, the estimation of the stray flux and the comparison with the experimental measurements are reported in the paper

Predicting radio fluxes of extrasolar planets (Griessmeier+, 2007)

NARCIS (Netherlands)

Griessmeier, J.M.; Zarka, P.; Spreeuw, H.

2007-01-01

Expected radio emission from presently known exoplanets. For each of the currently known exoplanets, we list its estimated magnetic moment, maximum radio emission frequency, plasma frequency in the ambient stellar wind, and radio fluxes according to three different models. (1 data file).

Shielding calculations for a 30 MeV proton accelerator

International Nuclear Information System (INIS)

Nandy, Maitreyee; Sarkar, P.K.

2003-01-01

Full text: The thickness of the shield, made of ordinary concrete, required to reduce the equivalent dose rate below the maximum permissible limit and to ensure safe operation of a 30 MeV proton accelerator has been estimated using the Moyer model. Required double differential neutron yield from thick stopping targets has been calculated for several reactions to be used for production of 67 Ga, 111 In, 123 I and 201 Tl radioisotopes. The neutron emission at 0 deg and 90 deg angles with respect to the incident beam direction is estimated using the hybrid model code ALICE91 which considers preequilibrium and equilibrium emissions from the target+projectile composite system. From this neutron yield the equivalent neutron dose rate at unit distance is determined using the ICRP recommended flux-to-dose conversion factors

Modeling of Drift Effects on Solar Tower Concentrated Flux Distributions

Directory of Open Access Journals (Sweden)

Luis O. Lara-Cerecedo

2016-01-01

Full Text Available A novel modeling tool for calculation of central receiver concentrated flux distributions is presented, which takes into account drift effects. This tool is based on a drift model that includes different geometrical error sources in a rigorous manner and on a simple analytic approximation for the individual flux distribution of a heliostat. The model is applied to a group of heliostats of a real field to obtain the resulting flux distribution and its variation along the day. The distributions differ strongly from those obtained assuming the ideal case without drift or a case with a Gaussian tracking error function. The time evolution of peak flux is also calculated to demonstrate the capabilities of the model. The evolution of this parameter also shows strong differences in comparison to the case without drift.

Neutron flux in a periodical slab geometry (1960); Flux neutronique dans un milieu multiplicateur perturbe par la presence de lames planes (1960)

Energy Technology Data Exchange (ETDEWEB)

Lamare, J de; Mathelot, P; Cadhilac, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

In the present report, we explain an original method to perform exact calculations of neutron flux in either of two geometries: a slab surrounded by an infinite multiplying medium or a periodical, one dimensional array of two different media. (author) [French] La methode exposee dans ce rapport permet de calculer exactement la distribution du flux dans un milieu multiplicateur infini dont l'uniformite est perturbee par l'interposition d'une lame plane de nature differente, ou dont l'heterogeneite est de caractere periodique et unidimensionnel. (auteur)

Heat flux driven ion turbulence

International Nuclear Information System (INIS)

Garbet, X.

1998-01-01

This work is an analysis of an ion turbulence in a tokamak in the case where the thermal flux is fixed and the temperature profile is allowed to fluctuate. The system exhibits some features of Self-Organized Critical systems. In particular, avalanches are observed. Also the frequency spectrum of the thermal flux exhibits a structure similar to the one of a sand pile automaton, including a 1/f behavior. However, the time average temperature profile is found to be supercritical, i.e. the temperature gradient stays above the critical value. Moreover, the heat diffusivity is lower for a turbulence calculated at fixed flux than a fixed temperature gradient, with the same time average temperature. This behavior is attributed to a stabilizing effect of avalanches. (author)

MICROX-2: an improved two-region flux spectrum code for the efficient calculation of group cross sections

International Nuclear Information System (INIS)

Mathews, D.; Koch, P.

1979-12-01

The MICROX-2 code is an improved version of the MICROX code. The improvements allow MICROX-2 to be used for the efficient and rigorous preparation of broad group neutron cross sections for poorly moderated systems such as fast breeder reactors in addition to the well moderated thermal reactors for which MICROX was designed. MICROX-2 is an integral transport theory code which solves the neutron slowing down and thermalization equations on a detailed energy grid for two-region lattice cells. The fluxes in the two regions are coupled by transport corrected collision probabilities. The inner region may include two different types of grains (particles). Neutron leakage effects are treated by performing B 1 slowing down and P 0 plus DB 2 thermalization calculations in each region. Cell averaged diffusion coefficients are prepared with the Benoist cell homogenization prescription

Uncertainties in (E)UV model atmosphere fluxes

Science.gov (United States)

Rauch, T.

2008-04-01

Context: During the comparison of synthetic spectra calculated with two NLTE model atmosphere codes, namely TMAP and TLUSTY, we encounter systematic differences in the EUV fluxes due to the treatment of level dissolution by pressure ionization. Aims: In the case of Sirius B, we demonstrate an uncertainty in modeling the EUV flux reliably in order to challenge theoreticians to improve the theory of level dissolution. Methods: We calculated synthetic spectra for hot, compact stars using state-of-the-art NLTE model-atmosphere techniques. Results: Systematic differences may occur due to a code-specific cutoff frequency of the H I Lyman bound-free opacity. This is the case for TMAP and TLUSTY. Both codes predict the same flux level at wavelengths lower than about 1500 Å for stars with effective temperatures (T_eff) below about 30 000 K only, if the same cutoff frequency is chosen. Conclusions: The theory of level dissolution in high-density plasmas, which is available for hydrogen only should be generalized to all species. Especially, the cutoff frequencies for the bound-free opacities should be defined in order to make predictions of UV fluxes more reliable.

Fluxes of ammonia in the coastal marine boundary layer

DEFF Research Database (Denmark)

Sørensen, L.L.; Hertel, O.; Skjøth, C.A.

2003-01-01

Concentrations of ammonia in air and ammonium in surface water were measured from a platform in the Southern North Sea close to the Dutch coast. Fluxes were derived from the measurements applying Monin-Obukhov similarity theory and exchange velocities calculated. The fluxes and air concentrations...

Configuration of LWR fuel enrichment or burnup yielding maximum power

International Nuclear Information System (INIS)

Bartosek, V.; Zalesky, K.

1976-01-01

An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

Flux line lattice in type II super conductors

International Nuclear Information System (INIS)

Manindra Kumar; Singh, Arun Kumar; Surendra Kumar

2003-01-01

The shear modules C 66 of the flux line lattice in type II super conductors can be obtained from a two body interaction between the flux lines even at large inductions B ∼ HC 2 . The potential is composed of a repulsive and an attractive part and has a range diverging at HC 2 . An explicit expression for the Ginzberg-Landau C 66 is given for arbitrary B and k' (G-L parameter). The graph for C 66 exhibits the expected maximum at a certain value of b. (author)

The Open Flux Problem

Science.gov (United States)

Linker, J. A.; Caplan, R. M.; Downs, C.; Riley, P.; Mikic, Z.; Lionello, R.; Henney, C. J.; Arge, C. N.; Liu, Y.; Derosa, M. L.; Yeates, A.; Owens, M. J.

2017-10-01

The heliospheric magnetic field is of pivotal importance in solar and space physics. The field is rooted in the Sun’s photosphere, where it has been observed for many years. Global maps of the solar magnetic field based on full-disk magnetograms are commonly used as boundary conditions for coronal and solar wind models. Two primary observational constraints on the models are (1) the open field regions in the model should approximately correspond to coronal holes (CHs) observed in emission and (2) the magnitude of the open magnetic flux in the model should match that inferred from in situ spacecraft measurements. In this study, we calculate both magnetohydrodynamic and potential field source surface solutions using 14 different magnetic maps produced from five different types of observatory magnetograms, for the time period surrounding 2010 July. We have found that for all of the model/map combinations, models that have CH areas close to observations underestimate the interplanetary magnetic flux, or, conversely, for models to match the interplanetary flux, the modeled open field regions are larger than CHs observed in EUV emission. In an alternative approach, we estimate the open magnetic flux entirely from solar observations by combining automatically detected CHs for Carrington rotation 2098 with observatory synoptic magnetic maps. This approach also underestimates the interplanetary magnetic flux. Our results imply that either typical observatory maps underestimate the Sun’s magnetic flux, or a significant portion of the open magnetic flux is not rooted in regions that are obviously dark in EUV and X-ray emission.

The Open Flux Problem

Energy Technology Data Exchange (ETDEWEB)

Linker, J. A.; Caplan, R. M.; Downs, C.; Riley, P.; Mikic, Z.; Lionello, R. [Predictive Science Inc., 9990 Mesa Rim Road, Suite 170, San Diego, CA 92121 (United States); Henney, C. J. [Air Force Research Lab/Space Vehicles Directorate, 3550 Aberdeen Avenue SE, Kirtland AFB, NM (United States); Arge, C. N. [Science and Exploration Directorate, NASA/GSFC, Greenbelt, MD 20771 (United States); Liu, Y. [W. W. Hansen Experimental Physics Laboratory, Stanford University, Stanford, CA 94305 (United States); Derosa, M. L. [Lockheed Martin Solar and Astrophysics Laboratory, 3251 Hanover Street B/252, Palo Alto, CA 94304 (United States); Yeates, A. [Department of Mathematical Sciences, Durham University, Durham, DH1 3LE (United Kingdom); Owens, M. J., E-mail: linkerj@predsci.com [Space and Atmospheric Electricity Group, Department of Meteorology, University of Reading, Earley Gate, P.O. Box 243, Reading RG6 6BB (United Kingdom)

2017-10-10

The heliospheric magnetic field is of pivotal importance in solar and space physics. The field is rooted in the Sun’s photosphere, where it has been observed for many years. Global maps of the solar magnetic field based on full-disk magnetograms are commonly used as boundary conditions for coronal and solar wind models. Two primary observational constraints on the models are (1) the open field regions in the model should approximately correspond to coronal holes (CHs) observed in emission and (2) the magnitude of the open magnetic flux in the model should match that inferred from in situ spacecraft measurements. In this study, we calculate both magnetohydrodynamic and potential field source surface solutions using 14 different magnetic maps produced from five different types of observatory magnetograms, for the time period surrounding 2010 July. We have found that for all of the model/map combinations, models that have CH areas close to observations underestimate the interplanetary magnetic flux, or, conversely, for models to match the interplanetary flux, the modeled open field regions are larger than CHs observed in EUV emission. In an alternative approach, we estimate the open magnetic flux entirely from solar observations by combining automatically detected CHs for Carrington rotation 2098 with observatory synoptic magnetic maps. This approach also underestimates the interplanetary magnetic flux. Our results imply that either typical observatory maps underestimate the Sun’s magnetic flux, or a significant portion of the open magnetic flux is not rooted in regions that are obviously dark in EUV and X-ray emission.

Interaction of storage carbohydrates and other cyclic fluxes with central metabolism : A quantitative approach by non-stationary 13C metabolic flux analysis

NARCIS (Netherlands)

Suarez-Mendez, C. A.; Hanemaaijer, M.; ten Pierick, Angela; Wolters, J. C.; Heijnen, J.J.; Wahl, S. A.

2016-01-01

13C labeling experiments in aerobic glucose limited cultures of Saccharomyces cerevisiae at four different growth rates (0.054; 0.101, 0.207, 0.307 h-1) are used for calculating fluxes that include intracellular cycles (e.g., storage carbohydrate cycles, exchange fluxes with amino acids), which are

Third law of thermodynamics in the presence of a heat flux

International Nuclear Information System (INIS)

Camacho, J.

1995-01-01

Following a maximum entropy formalism, we study a one-dimensional crystal under a heat flux. We obtain the phonon distribution function and evaluate the nonequilibrium temperature, the specific heat, and the entropy as functions of the internal energy and the heat flux, in both the quantum and the classical limits. Some analogies between the behavior of equilibrium systems at low absolute temperature and nonequilibrium steady states under high values of the heat flux are shown, which point to a possible generalization of the third law in nonequilibrium situations

Neutron flux monitor

International Nuclear Information System (INIS)

Seki, Eiji; Tai, Ichiro.

1984-01-01

Purpose: To maintain the measuring accuracy and the reponse time within an allowable range in accordance with the change of neutron fluxes in a nuclear reactor pressure vessel. Constitution: Neutron fluxes within a nuclear reactor pressure vessel are detected by detectors, converted into pulse signals and amplified in a range switching amplifier. The amplified signals are further converted through an A/D converter and digital signals from the converter are subjected to a square operation in an square operation circuit. The output from the circuit is inputted into an integration circuit to selectively accumulate the constant of 1/2n, 1 - 1/2n (n is a positive integer) respectively for two continuing signals to perform weighing. Then, the addition is carried out to calculate the integrated value and the addition number is changed by the chane in the number n to vary the integrating time. The integrated value is inputted into a control circuit to control the value of n so that the fluctuation and the calculation time for the integrated value are within a predetermined range and, at the same time, the gain of the range switching amplifier is controlled. (Seki, T.)

Study on minimum heat-flux point during boiling heat transfer on horizontal plates

International Nuclear Information System (INIS)

Nishio, Shigefumi

1985-01-01

The characteristics of boiling heat transfer are usually shown by the boiling curve of N-shape having the maximum and minimum points. As for the limiting heat flux point, that is, the maximum point, there have been many reports so far, as it is related to the physical burn of heat flux-controlling type heating surfaces. But though the minimum heat flux point is related to the quench point as the problems in steel heat treatment, the core safety of LWRs, the operational stability of superconducting magnets, the start-up characteristics of low temperature machinery, the condition of vapor explosion occurrence and so on, the systematic information has been limited. In this study, the effects of transient property and the heat conductivity of heating surfaces on the minimum heat flux condition in the pool boiling on horizontal planes were experimentally examined by using liquid nitrogen. The experimental apparatuses for steady boiling, for unsteady boiling with a copper heating surface, and for unsteady boiling with a heating surface other than copper were employed. The boiling curves obtained with these apparatuses and the minimum heat flux point condition are discussed. (Kako, I.)

KoFlux: Korean Regional Flux Network in AsiaFlux

Science.gov (United States)

Kim, J.

2002-12-01

AsiaFlux, the Asian arm of FLUXNET, held the Second International Workshop on Advanced Flux Network and Flux Evaluation in Jeju Island, Korea on 9-11 January 2002. In order to facilitate comprehensive Asia-wide studies of ecosystem fluxes, the meeting launched KoFlux, a new Korean regional network of long-term micrometeorological flux sites. For a successful assessment of carbon exchange between terrestrial ecosystems and the atmosphere, an accurate measurement of surface fluxes of energy and water is one of the prerequisites. During the 7th Global Energy and Water Cycle Experiment (GEWEX) Asian Monsoon Experiment (GAME) held in Nagoya, Japan on 1-2 October 2001, the Implementation Committee of the Coordinated Enhanced Observing Period (CEOP) was established. One of the immediate tasks of CEOP was and is to identify the reference sites to monitor energy and water fluxes over the Asian continent. Subsequently, to advance the regional and global network of these reference sites in the context of both FLUXNET and CEOP, the Korean flux community has re-organized the available resources to establish a new regional network, KoFlux. We have built up domestic network sites (equipped with wind profiler and radiosonde measurements) over deciduous and coniferous forests, urban and rural rice paddies and coastal farmland. As an outreach through collaborations with research groups in Japan, China and Thailand, we also proposed international flux sites at ecologically and climatologically important locations such as a prairie on the Tibetan plateau, tropical forest with mixed and rapid land use change in northern Thailand. Several sites in KoFlux already begun to accumulate interesting data and some highlights are presented at the meeting. The sciences generated by flux networks in other continents have proven the worthiness of a global array of micrometeorological flux towers. It is our intent that the launch of KoFlux would encourage other scientists to initiate and

Natural Elemental Concentrations and Fluxes: Their Use as Indicators of Repository Safety

Energy Technology Data Exchange (ETDEWEB)

Miller, Bill; Lind, Andy; Savage, Dave; Maul, Philip; Robinson, Peter [EnvirosQuantisci, Melton Mowbray (United Kingdom)

2002-03-01

The calculated post-closure performance of a radioactive waste repository is generally quantified in terms of radiological dose or risk to humans, with safety being determined by whether the calculated exposure values are consistent with predetermined target criteria which are deemed to represent acceptable radiological hazards. Despite their general acceptance, however, dose and risk are not perfect measures of repository safety because, in order to calculate them, gross assumptions must be made for future human behaviour patterns. Such predictions clearly become increasingly uncertain as forecasts are made further into the future. As a consequence, there has been a growing interest in developing other ways of assessing repository safety which do not require assumptions to be made for future human behaviour. One proposed assessment method is to use the distributions of naturally-occurring chemical species in the environment, expressed either as concentrations or fluxes of elements, radionuclides or radioactivity, as natural safety indicators which may be compared with the PA predictions of repository releases. Numerous comparisons are possible between the repository and natural systems. The primary objective is to use the natural system to provide context to the hazard presented by the repository releases. Put simply, if it can be demonstrated that the flux to the biosphere from the repository is not significant compared with the natural flux from the geosphere, then its radiological significance should not be of great or priority concern. Natural safety indicators may be quantified on a site specific basis, using information derived from a repository site characterisation programme, and can be compared to the outputs from the associated site specific PAs. Such calculations and comparisons may be very detailed and might examine, for example, the spatial and temporal variations in the distributions and fluxes of naturally-occurring chemical species arising from

SREM - WRS system module number 3348 for calculating the removal flux due to point, line or disc sources

International Nuclear Information System (INIS)

Grimstone, M.J.

1978-06-01

The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the uncollided flux and first-collision source from a disc source in a slab geometry system, a line source at the centre of a cylindrical system or a point source at the centre of a spherical system. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

Energy Technology Data Exchange (ETDEWEB)

Smith, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. N. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-15

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further

An evaluation of multigroup flux predictions in the EBR-II core

International Nuclear Information System (INIS)

Hill, R.N.; Fanning, T.H.; Finck, P.J.

1991-01-01

The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required

An evaluation of multigroup flux predictions in the EBR-II core

Energy Technology Data Exchange (ETDEWEB)

Hill, R.N.; Fanning, T.H.; Finck, P.J.

1991-12-31

The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

An evaluation of multigroup flux predictions in the EBR-II core

Energy Technology Data Exchange (ETDEWEB)

Hill, R.N.; Fanning, T.H.; Finck, P.J.

1991-01-01

The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.

Calculation system for physical analysis of boiling water reactors

International Nuclear Information System (INIS)

Bouveret, F.

2001-01-01

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)

Dependence of Core and Extended Flux on Core Dominance ...

Indian Academy of Sciences (India)

Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

Dynamical Processes in Flux Tubes and their Role in ...

Indian Academy of Sciences (India)

We model the dynamical interaction between magnetic flux tubes and granules in the solar photosphere which leads to the excitation of transverse (kink) and longitudinal (sausage) tube waves. The investigation is motivated by the interpretation of network oscillations in terms of flux tube waves. The calculations show that ...

Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

International Nuclear Information System (INIS)

Bittelli, U.D.

1988-01-01

This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

Determination of the Atmospheric Neutrino Fluxes from Atmospheric Neutrino Data

NARCIS (Netherlands)

Gonzalez-Garcia, M. C.; Maltoni, M.; Rojo, J.

2006-01-01

The precise knowledge of the atmospheric neutrino fluxes is a key ingredient in the interpretation of the results from any atmospheric neutrino experiment. In the standard atmospheric neutrino data analysis, these fluxes are theoretical inputs obtained from sophisticated numerical calculations based

Maximum Efficiency per Torque Control of Permanent-Magnet Synchronous Machines

Directory of Open Access Journals (Sweden)

Qingbo Guo

2016-12-01

Full Text Available High-efficiency permanent-magnet synchronous machine (PMSM drive systems need not only optimally designed motors but also efficiency-oriented control strategies. However, the existing control strategies only focus on partial loss optimization. This paper proposes a novel analytic loss model of PMSM in either sine-wave pulse-width modulation (SPWM or space vector pulse width modulation (SVPWM which can take into account both the fundamental loss and harmonic loss. The fundamental loss is divided into fundamental copper loss and fundamental iron loss which is estimated by the average flux density in the stator tooth and yoke. In addition, the harmonic loss is obtained from the Bertotti iron loss formula by the harmonic voltages of the three-phase inverter in either SPWM or SVPWM which are calculated by double Fourier integral analysis. Based on the analytic loss model, this paper proposes a maximum efficiency per torque (MEPT control strategy which can minimize the electromagnetic loss of PMSM in the whole operation range. As the loss model of PMSM is too complicated to obtain the analytical solution of optimal loss, a golden section method is applied to achieve the optimal operation point accurately, which can make PMSM work at maximum efficiency. The optimized results between SPWM and SVPWM show that the MEPT in SVPWM has a better effect on the optimization performance. Both the theory analysis and experiment results show that the MEPT control can significantly improve the efficiency performance of the PMSM in each operation condition with a satisfied dynamic performance.

Experimental study of flux depressions and anti-reactivities created by irradiation loops; Etude experimentale des depressions de flux et antireactivites creees par les dispositifs d'irradiation

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

Methods for fast computing of thermal flux depressions and reactivities created by irradiation-loops in natural water reactors are studied in this report. The classical methods of approximation which have been used are: diffusion theory or absorption-probability calculations for the flux-depression and perturbation theory for the anti-reactivities. Pertinent formulae are compiled together with graphs from theoretical calculations. These formulae and graphs have been checked from numerous experiments which show that the approximations used here are quite close to the actual physical situation, even when the theories are based from assumptions which cannot be verified here. (author) [French] Ce rapport propose aux experimentateurs des piles a eau legere des methodes de determination rapide des depressions de flux thermique et antireactivites creees par les dispositifs d'irradiation. Les methodes classiques d'approximation sont utilisees, a savoir: theorie de diffusion ou calcul de probabilites d'absorption pour les depressions de flux, theorie des perturbations pour les antireactivites. Un formulaire pratique, accompagne d'abaques est deduit des calculs theoriques et verifie par de nombreuses experiences qui montrent que les evaluations faites sont tres proches de la realite, meme dans le cas ou les hypotheses relatives aux theories utilisees ne sont pas respectees. (auteur)

Experimental study of flux pinning in NbN films and multilayers: Ultimate limits on critical currents in superconductors

International Nuclear Information System (INIS)

Gray, K.E.; Kampwirth, R.T.; Capone, D.W. II; Murduck, J.M.

1988-08-01

A flux pinning model is presented which predicts the maximum critical current density attainable in superconductors. That such a limit must exist comes from the realization that flux pinning is strongest in regions of weak superconductivity, but these regions cannot carry a large supercurrent. Since the same regions within the superconductor cannot be used for both pinning and supercurrent conductions, there must be an optimum mix, leading to a maximum J/sub c/. Measurements on films and multilayers of NbN have verified many details of the model including anisotropy effects and a strong reduction in J/sub c/ for defect spacings smaller than the flux core diameter. In an optimized multilayer the pinning force reached /approximately/22% of the theoretical maximum. The implications of these results on the practical applications of NbN films and on the maximum critical current density in the new high temperature superconductors are also discussed. 24 refs., 4 figs

Photoneutron Flux Measurement via Neutron Activation Analysis in a Radiotherapy Bunker with an 18 MV Linear Accelerator

Science.gov (United States)

Çeçen, Yiğit; Gülümser, Tuğçe; Yazgan, Çağrı; Dapo, Haris; Üstün, Mahmut; Boztosun, Ismail

2017-09-01

In cancer treatment, high energy X-rays are used which are produced by linear accelerators (LINACs). If the energy of these beams is over 8 MeV, photonuclear reactions occur between the bremsstrahlung photons and the metallic parts of the LINAC. As a result of these interactions, neutrons are also produced as secondary radiation products (γ,n) which are called photoneutrons. The study aims to map the photoneutron flux distribution within the LINAC bunker via neutron activation analysis (NAA) using indium-cadmium foils. Irradiations made at different gantry angles (0°, 90°, 180° and 270°) with a total of 91 positions in the Philips SLI-25 linear accelerator treatment room and location-based distribution of thermal neutron flux was obtained. Gamma spectrum analysis was carried out with high purity germanium (HPGe) detector. Results of the analysis showed that the maximum neutron flux in the room occurred at just above of the LINAC head (1.2x105 neutrons/cm2.s) which is compatible with an americium-beryllium (Am-Be) neutron source. There was a 90% decrease of flux at the walls and at the start of the maze with respect to the maximum neutron flux. And, just in front of the LINAC door, inside the room, neutron flux was measured less than 1% of the maximum.

Photoneutron Flux Measurement via Neutron Activation Analysis in a Radiotherapy Bunker with an 18 MV Linear Accelerator

Directory of Open Access Journals (Sweden)

Çeçen Yiğit

2017-01-01

Full Text Available In cancer treatment, high energy X-rays are used which are produced by linear accelerators (LINACs. If the energy of these beams is over 8 MeV, photonuclear reactions occur between the bremsstrahlung photons and the metallic parts of the LINAC. As a result of these interactions, neutrons are also produced as secondary radiation products (γ,n which are called photoneutrons. The study aims to map the photoneutron flux distribution within the LINAC bunker via neutron activation analysis (NAA using indium-cadmium foils. Irradiations made at different gantry angles (0°, 90°, 180° and 270° with a total of 91 positions in the Philips SLI-25 linear accelerator treatment room and location-based distribution of thermal neutron flux was obtained. Gamma spectrum analysis was carried out with high purity germanium (HPGe detector. Results of the analysis showed that the maximum neutron flux in the room occurred at just above of the LINAC head (1.2x105 neutrons/cm2.s which is compatible with an americium-beryllium (Am-Be neutron source. There was a 90% decrease of flux at the walls and at the start of the maze with respect to the maximum neutron flux. And, just in front of the LINAC door, inside the room, neutron flux was measured less than 1% of the maximum.

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)

The calculation of Tritium burnup in Tokamaks

International Nuclear Information System (INIS)

Bittoni, E.; Haegi, M.

1987-01-01

In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code

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

Energy Technology Data Exchange (ETDEWEB)

Authier, N

1998-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Authier, N

1998-12-01

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

Flux-line-cutting losses in type-II superconductors

International Nuclear Information System (INIS)

Clem, J.R.

1982-01-01

Energy dissipation associated with flux-line cutting (intersection and cross-joining of adjacent nonparallel vortices) is considered theoretically. The flux-line-cutting contribution to the dissipation per unit volume, arising from mutual annihilation of transverse magnetic flux, is identified as J/sub parallel/xE/sub parallel/, where J/sub parallel/ and E/sub parallel/ are the components of the current density and the electric field parallel to the magnetic induction. The dynamical behavior of the magnetic structure at the flux-line-cutting threshold is shown to be governed by a special critical-state model similar to that proposed by previous authors. The resulting flux-line-cutting critical-state model, characterized in planar geometry by a parallel critical current density J/sub c/parallel or a critical angle gradient k/sub c/, is used to calculate predicted hysteretic ac flux-line-cutting losses in type-II superconductors in which the flux pinning is weak. The relation of the theory to previous experiments is discussed

VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

International Nuclear Information System (INIS)

Teuchert, E.; Haas, K.A.

1995-01-01

flux for 1515 compositions in 2-D cases, r-z (9999 compositions in 3-D cases, x-y-z). The burnup scheme has been developed from the FEVER code. The build-up history of up to 49 fission product nuclides in the compositions is followed explicitly. The diffusion part of the program system can be repeated at many short burnup time steps, and the spectrum module can be repeated at larger time steps, when some significant change in the spectrum is expected. The fuel management and cost module performs the fuel shuffling and general evaluations of the reactor and fuel element life history. The fuel management simulates the currently known shuffling and out of pile routes for various reactors. It has further been extended to include the typical features of the pebble bed reactor such as burnup dependent optional reloading of elements, separated treatment of different fuel streams, and recycling in new fuel element types according to a consistent mass balance and timing. Optionally, several different types of data files can be set up with characteristic data of the reactor life. These are used for more detailed investigations and display programs. The restart option allows the study of special phases of the reactor life, e.g. changes of the fueling scheme, of the burnup, of the power output, of the coolant temperature, and of the corresponding reactivity effects. The fuel cycle cost data set is made for the present worth KPD code. Two-dimensional thermal hydraulics studies for operating and emergency conditions can be performed with the THERMIX code. The averaged temperatures of the different spectrum zones in the core are returned from the thermal hydraulics to the subsequent step of the reactor history. 3 - Restrictions on the complexity of the problem: In epithermal energy range the cell spectrum calculation is missing. If needed, it must be simulated by disadvantage factors being obtained in other codes. Further, dynamic common must be defined for the commons VARDIM, COCI

Device for investigation of magnetic flux jumps in ribbon superconductors

International Nuclear Information System (INIS)

Andrianov, A.V.; Bashkirov, Yu.A.; Kremlev, M.G.

1986-01-01

A device for simulation of magnetic flux jumps in superconductors of conducting magnet sandwich-type windings super-applyed of a ribbon conductor is described. A superconducting magnet with a measuring cassetter are the main elements of the device. An external magnetic field is generated by a two-sectional superconducting magnet permitting to simulate the shape of the magnetic field characteristic for sandwich-type windings. Maximum radial component of the magnetic field is 2 T. Jumps of the magnetic flux are recorded by induction transducers and the magnetic field-by Hall trasducer. The effect of coating of standard metal on magnetic flux jumps in Nb 3 Sn base superconducting ribbon is considered

Study on radiation flux of the receiver with a parabolic solar concentrator system

International Nuclear Information System (INIS)

Mao, Qianjun; Shuai, Yong; Yuan, Yuan

2014-01-01

Highlights: • The idea of integral dish and multi-dishes in a parabolic solar collector has been proposed. • The impacts of three factors of the receiver have been investigated. • The radiation flux distribution can benefit from a large system error. - Abstract: The solar receiver plays a key role in the performance of a solar dish electric generator. Its radiation flux distribution can directly affect the efficiency of the parabolic solar concentrator system. In this paper, radiation flux distribution of the receiver is simulated successfully using MCRT method. The impacts of incident solar irradiation, aspect ratio (the ratio of the receiver height to the receiver diameter), and system error on the radiation flux of the receiver are investigated. The parameters are studied in the following ranges: incident solar irradiation from 100 to 1100 W/m 2 , receiver aspect ratio from 0.5 to 1.5, and the system error from 0 to 10 mrad. A non-dimensional parameter Θ is defined to represent the ratio of radiation flux to incident solar irradiation. The results show that the maximum of Θ is about 200 in simulation conditions. The aspect ratio and system error have a significant impact on the radiation flux. The optimal receiver aspect ratio is 1.5 at a constant incident solar irradiation, and the maximum of radiation flux increases with decreasing system error, however, the radiation flux distribution can benefit from a large system error. Meanwhile, effects of integral dish and multi-dishes on the radiation flux distribution have been investigated. The results show that the accuracy of two cases can be ignored within the same parameters

Calculation device for amount of heavy element nuclide in reactor fuels and calculation method therefor

International Nuclear Information System (INIS)

Naka, Takafumi; Yamamoto, Munenari.

1995-01-01

When there are two or more origins of deuterium nuclides in reactor fuels, there are disposed a memory device for an amount of deuterium nuclides for every origin in a noted fuel segment at a certain time point, a device for calculating the amount of nuclides for every origin and current neutron fluxes in the noted fuel segment, and a device for separating and then displaying the amount of deuterium nuclides for every origin. Equations for combustion are dissolved for every origin of the deuterium nuclides based on the amount of the deuterium nuclides for every origin and neutron fluxes, to calculate the current amount of deuterium nuclides for every origin. The amount of deuterium nuclides originated from uranium is calculated ignoring α-decay of curium, while the amount of deuterium nuclides originated from plutonium is calculated ignoring the generation of plutonium formed from neptunium. Deuterium nuclides can be measured and controlled accurately for every origin of the reactor fuels. Even when nuclear fuel materials have two or more nationalities, the measurement and control thereof can be conducted for every country. (N.H.)

Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

International Nuclear Information System (INIS)

Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

1987-01-01

A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

Experimental study of flux depressions and anti-reactivities created by irradiation loops; Etude experimentale des depressions de flux et antireactivites creees par les dispositifs d'irradiation

Energy Technology Data Exchange (ETDEWEB)

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

1967-07-01

Methods for fast computing of thermal flux depressions and reactivities created by irradiation-loops in natural water reactors are studied in this report. The classical methods of approximation which have been used are: diffusion theory or absorption-probability calculations for the flux-depression and perturbation theory for the anti-reactivities. Pertinent formulae are compiled together with graphs from theoretical calculations. These formulae and graphs have been checked from numerous experiments which show that the approximations used here are quite close to the actual physical situation, even when the theories are based from assumptions which cannot be verified here. (author) [French] Ce rapport propose aux experimentateurs des piles a eau legere des methodes de determination rapide des depressions de flux thermique et antireactivites creees par les dispositifs d'irradiation. Les methodes classiques d'approximation sont utilisees, a savoir: theorie de diffusion ou calcul de probabilites d'absorption pour les depressions de flux, theorie des perturbations pour les antireactivites. Un formulaire pratique, accompagne d'abaques est deduit des calculs theoriques et verifie par de nombreuses experiences qui montrent que les evaluations faites sont tres proches de la realite, meme dans le cas ou les hypotheses relatives aux theories utilisees ne sont pas respectees. (auteur)

Flux driven turbulence in tokamaks

International Nuclear Information System (INIS)

Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

1999-01-01

This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

Dissolved methane in the Beaufort Sea and the Arctic Ocean, 1992-2009; sources and atmospheric flux

Science.gov (United States)

Lorenson, Thomas D.; Greinert, Jens; Coffin, Richard B.

2016-01-01

Methane concentration and isotopic composition was measured in ice-covered and ice-free waters of the Arctic Ocean during eleven surveys spanning the years of 1992-1995 and 2009. During ice-free periods, methane flux from the Beaufort shelf varies from 0.14 to 0.43 mg CH4 m-2 day-1. Maximum fluxes from localized areas of high methane concentration are up to 1.52 mg CH4 m-2 day-1. Seasonal buildup of methane under ice can produce short-term fluxes of methane from the Beaufort shelf that varies from 0.28 to 1.01 to mg CH4 m-2 day-1. Scaled-up estimates of minimum methane flux from the Beaufort Sea and pan-Arctic shelf for both ice-free and ice-covered periods range from 0.02 Tg CH4 yr-1 and 0.30 Tg CH4 yr-1 respectively to maximum fluxes of 0.18 Tg CH4 yr-1 and 2.2 Tg CH4 yr-1 respectively. A methane flux of 0.36 Tg CH4 yr-1from the deep Arctic Ocean was estimated using data from 1993-94. The flux can be as much as 2.35 Tg CH4 yr-1 estimated from maximum methane concentrations and wind speeds of 12 m/s, representing only 0.42% of the annual atmospheric methane budget of ~560 Tg CH4 yr-1. There were no significant changes in methane fluxes during the time period of this study. Microbial methane sources predominate with minor influxes from thermogenic methane offshore Prudhoe Bay and the Mackenzie River delta and may include methane from gas hydrate. Methane oxidation is locally important on the shelf and is a methane sink in the deep Arctic Ocean.

Factors controlling vertical fluxes of prrticles in the Arabian Sea

Digital Repository Service at National Institute of Oceanography (India)

Nair, T.M.B.; Ramaswamy, V.; Parthiban, G.; Shankar, R.

)) in the western Arabian Sea. Carbonate contributed mainly by foraminifers and coccolithophorids, are the dominant component in all the traps. Opal fluxes were maximum in the western Arabian Sea. At all the locations, lithogenic percentages increased with depth...

Updated thermal model using simplified short-wave radiosity calculations

International Nuclear Information System (INIS)

Smith, J.A.; Goltz, S.M.

1994-01-01

An extension to a forest canopy thermal radiance model is described that computes the short-wave energy flux absorbed within the canopy by solving simplified radiosity equations describing flux transfers between canopy ensemble classes partitioned by vegetation layer and leaf slope. Integrated short-wave reflectance and transmittance-factors obtained from measured leaf optical properties were found to be nearly equal for the canopy studied. Short-wave view factor matrices were approximated by combining the average leaf scattering coefficient with the long-wave view factor matrices already incorporated in the model. Both the updated and original models were evaluated for a dense spruce fir forest study site in Central Maine. Canopy short-wave absorption coefficients estimated from detailed Monte Carlo ray tracing calculations were 0.60, 0.04, and 0.03 for the top, middle, and lower canopy layers corresponding to leaf area indices of 4.0, 1.05, and 0.25. The simplified radiosity technique yielded analogous absorption values of 0.55, 0.03, and 0.01. The resulting root mean square error in modeled versus measured canopy temperatures for all layers was less than 1°C with either technique. Maximum error in predicted temperature using the simplified radiosity technique was approximately 2°C during peak solar heating. (author)

Updated thermal model using simplified short-wave radiosity calculations

Energy Technology Data Exchange (ETDEWEB)

Smith, J. A.; Goltz, S. M.

1994-02-15

An extension to a forest canopy thermal radiance model is described that computes the short-wave energy flux absorbed within the canopy by solving simplified radiosity equations describing flux transfers between canopy ensemble classes partitioned by vegetation layer and leaf slope. Integrated short-wave reflectance and transmittance-factors obtained from measured leaf optical properties were found to be nearly equal for the canopy studied. Short-wave view factor matrices were approximated by combining the average leaf scattering coefficient with the long-wave view factor matrices already incorporated in the model. Both the updated and original models were evaluated for a dense spruce fir forest study site in Central Maine. Canopy short-wave absorption coefficients estimated from detailed Monte Carlo ray tracing calculations were 0.60, 0.04, and 0.03 for the top, middle, and lower canopy layers corresponding to leaf area indices of 4.0, 1.05, and 0.25. The simplified radiosity technique yielded analogous absorption values of 0.55, 0.03, and 0.01. The resulting root mean square error in modeled versus measured canopy temperatures for all layers was less than 1°C with either technique. Maximum error in predicted temperature using the simplified radiosity technique was approximately 2°C during peak solar heating. (author)

Installation of the MAXIMUM microscope at the ALS

International Nuclear Information System (INIS)

Ng, W.; Perera, R.C.C.; Underwood, J.H.; Singh, S.; Solak, H.; Cerrina, F.

1995-10-01

The MAXIMUM scanning x-ray microscope, developed at the Synchrotron Radiation Center (SRC) at the University of Wisconsin, Madison was implemented on the Advanced Light Source in August of 1995. The microscope's initial operation at SRC successfully demonstrated the use of multilayer coated Schwarzschild objective for focusing 130 eV x-rays to a spot size of better than 0.1 micron with an electron energy resolution of 250meV. The performance of the microscope was severely limited, because of the relatively low brightness of SRC, which limits the available flux at the focus of the microscope. The high brightness of the ALS is expected to increase the usable flux at the sample by a factor of 1,000. The authors will report on the installation of the microscope on bending magnet beamline 6.3.2 at the ALS and the initial measurement of optical performance on the new source, and preliminary experiments with surface chemistry of HF etched Si will be described

Fluxes of chemically reactive species inferred from mean concentration measurements

NARCIS (Netherlands)

Galmarini, S.; Vilà-Guerau De Arellano, J.; Duyzer, J.H.

1997-01-01

A method is presented for the calculation of the fluxes of chemically reactive species on the basis of routine measurements of meteorological variables and chemical species. The method takes explicity into account the influence of chemical reactions on the fluxes of the species. As a demonstration

Temporal and spatial changes in mixed layer properties and atmospheric net heat flux in the Nordic Seas

International Nuclear Information System (INIS)

Smirnov, A; Alekseev, G; Korablev, A; Esau, I

2010-01-01

The Nordic Seas are an important area of the World Ocean where warm Atlantic waters penetrate far north forming the mild climate of Northern Europe. These waters represent the northern rim of the global thermohaline circulation. Estimates of the relationships between the net heat flux and mixed layer properties in the Nordic Seas are examined. Oceanographic data are derived from the Oceanographic Data Base (ODB) compiled in the Arctic and Antarctic Research Institute. Ocean weather ship 'Mike' (OWS) data are used to calculate radiative and turbulent components of the net heat flux. The net shortwave flux was calculated using a satellite albedo dataset and the EPA model. The net longwave flux was estimated by Southampton Oceanography Centre (SOC) method. Turbulent fluxes at the air-sea interface were calculated using the COARE 3.0 algorithm. The net heat flux was calculated by using oceanographic and meteorological data of the OWS 'Mike'. The mixed layer depth was estimated for the period since 2002 until 2009 by the 'Mike' data as well. A good correlation between these two parameters has been found. Sensible and latent heat fluxes controlled by surface air temperature/sea surface temperature gradient are the main contributors into net heat flux. Significant correlation was found between heat fluxes variations at the OWS 'Mike' location and sea ice export from the Arctic Ocean.

Temporal and spatial changes in mixed layer properties and atmospheric net heat flux in the Nordic Seas

Energy Technology Data Exchange (ETDEWEB)

Smirnov, A; Alekseev, G [SI ' Arctic and Antarctic Research Institute' , St. Petersburg (Russian Federation); Korablev, A; Esau, I, E-mail: avsmir@aari.nw.r [Nansen Environmental and Remote Sensing Centre, Bergen (Norway)

2010-08-15

The Nordic Seas are an important area of the World Ocean where warm Atlantic waters penetrate far north forming the mild climate of Northern Europe. These waters represent the northern rim of the global thermohaline circulation. Estimates of the relationships between the net heat flux and mixed layer properties in the Nordic Seas are examined. Oceanographic data are derived from the Oceanographic Data Base (ODB) compiled in the Arctic and Antarctic Research Institute. Ocean weather ship 'Mike' (OWS) data are used to calculate radiative and turbulent components of the net heat flux. The net shortwave flux was calculated using a satellite albedo dataset and the EPA model. The net longwave flux was estimated by Southampton Oceanography Centre (SOC) method. Turbulent fluxes at the air-sea interface were calculated using the COARE 3.0 algorithm. The net heat flux was calculated by using oceanographic and meteorological data of the OWS 'Mike'. The mixed layer depth was estimated for the period since 2002 until 2009 by the 'Mike' data as well. A good correlation between these two parameters has been found. Sensible and latent heat fluxes controlled by surface air temperature/sea surface temperature gradient are the main contributors into net heat flux. Significant correlation was found between heat fluxes variations at the OWS 'Mike' location and sea ice export from the Arctic Ocean.

Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

International Nuclear Information System (INIS)