INIS Repository Search - Results for the query - DEI:"NEUTRON TRANSPORT"
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http://inis.iaea.org/Search/search.aspx?orig_q=RN:21004861
Comparisons of some neutron transport critical calculationsColombo, V.; Ravetto, P. (Dipartimento di Energetica, Politecnico di Torino, 10129 Torino (Italy))</br>Critical calculations can constitute a good test for the comparisons of the performances of numerical methods to solve the neutron transport equation for multiplying systems. For some paradigmatic calculations, physically significant...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:28075567
On the reciprocity-like relations in linear neutron transport theoryModak, R.S.; Sahni, D.C. (Bhabha Atomic Research Centre, Theoretical Physics Div., Bombay (India))</br>The existence of certain reciprocity-like relations in neutron transport theory was shown earlier under some quite restrictive conditions. Here, these relations are shown to be valid in more general situations by using a different...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:9390674
Neutron spectra associated with a fast pulse assemblyHarvey, J.T.</br>Arkansas Univ., Little Rock (USA)</br>Neutron spectra have been measured by the threshold foil activation technique for the White Sands Missile Range Fast Burst Reactor. The neutron spectra for free-field, free-field with the experimenters table in place, for the in-core...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:6173324
Properties of neutron migration operators for fast multiplying assembliesOrechwa, Y.; Dorning, J. (Univ. of Illinois, Urbana)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:20071087
Iteration and extrapolation methods for the approximate solution of the even-parity transport equation for systems with voidsAckroyd, R.T. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering); Riyait, N.S. (Imperial Coll. of Science and Technology, London (UK). Nuclear Power Group)</br>Conventional finite-element solutions of the even-parity transport equation for systems with voids treat the void as a region of low absorption. This treatment tends to give physically-unacceptable solutions to void problems as the...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17011626
Elementary calculation of the extrapolationBoffi, V.C.; Haggag, M.H.; Spiga, G. (Bologna Univ. (Italy). Lab. di Ingegneria Nucleare)</br>A simple projectional technique combined with an equally simple parametric representation of the transient part of the neutron total flux is proposed for an elementary straightforward calculation of the extrapolation distance in diffusing...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:38031135
Computing time-eigenvalues using the even-parity transport formLathouwers, D. (Department of Radiation, Radionuclides and Reactors, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)), E-mail: D.Lathouwers@tudelft.nl</br>An adaption is presented to an earlier method for the prediction of time-eigenvalues of the neutron transport equation. Instead of using the full-parity first-order equation, the algorithm uses the even-parity formalism in the most...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:28032418
A generic method of analytical solution for approximations of the linear transport equationCardona, Augusto V.</br>Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Escola de Engenharia</br>In this work it is presented a generic method of analytical solution to the one-dimensional S<sub>N</sub>, P<sub>N</sub>, W<sub>N</sub>, Ch<sub>N</sub>, A<sub>N</sub> and L D<sub>N</sub> approximations of the linear transport equation....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18063720
The criticality problems with the Fsub(N) method for the FBIS modelTezcan, C. (Maden Tetkik ve Arama Enstitusu, Ankara (Turkey)); Yildiz, C. (Istanbul Technical Univ., (Turkey). Dept. of Physics)</br>In one-speed, time-independent, neutron transport theory, the Fsub(N) method is used for the FBIS (forward-backward-isotropic scattering) model to reinvestigate the behaviour of the critical size in plane and spherical geometries....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17011623
Special eigenvalues of the time-dependent one-speed neutron-transport equation for a homogeneous sphere with anisotropic scatteringDahl, E.B. (Chalmers Univ. of Tech., Goeteborg (Sweden). Dept. of Reactor Physics)</br>The time dependence of a neutron population in a homogeneous sphere has been studied. The neutrons are of one speed and are assumed to be scattered with linear anisotropy. Vacuum boundary conditions are used. It is shown that the...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:20045841
A new application of space asymptotic transport theoryColombo, V.; Corno, S.E.; Ravetto, P. (Politecnico di Torino (Italy))</br>The present paper concerns a proper statement and a new application of the so-called space asymptotic neutron transport theory in stationary reactor physics. In the first part we formulate the essential assumptions of space asymptotic...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17011624
The energy-dependent backward-forward-isotropic scattering model with some applications to the neutron transport equationWilliams, M.M.R. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering)</br>A multigroup formalism is developed for the backward-forward-isotropic scattering model of neutron transport. Some exact solutions are obtained in two-group theory for slab and spherical geometry. The results are useful for benchmark...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:35051185
Summary of neutronicsReuss, P.</br>This book takes stock on the last knowledge concerning the neutronics, part of the nuclear physique, which studies the neutrons transport in matter and the corresponding reactions. (A.L.B.)
http://inis.iaea.org/Search/search.aspx?orig_q=RN:16056471
The albedo problem in the case of multiple synthetic scattering taking place in a plane-symmetric slabShafiq, A.; Meyer, H.E. de; Grosjean, C.C. (Ghent Rijksuniversiteit (Belgium). Seminarie voor Wiskundige Natuurkunde)</br>An approximate model based on an improved diffusion-type theory is established for treating multiple synthetic scattering in a homogeneous slab of finite thickness. As in the case of the exact treatment given in the preceding paper...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:15013957
A finite element method for neutron transportAckroyd, R.T. (UKAEA Risley Nuclear Power Development Establishment. Process Technology and Safety Directorate)</br>A completely boundary-free maximum principle for the first-order Boltzmann equation is derived from the completely boundary-free maximum principle for the mixed-parity Boltzmann equation. When continuity is imposed on the trial function...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:8306936
Variational coarse-mesh transport method based on constrained angular trial functionsBriggs, L.L.; Lewis, E.E. (Northwestern Univ., Evanston, IL)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:16033015
Finite element method for solving neutron transport problemsFerguson, J.M.; Greenbaum, A.</br>Lawrence Livermore National Lab., CA (USA)</br>A finite element method is introduced for solving the neutron transport equations. Our method falls into the category of Petrov-Galerkin solution, since the trial space differs from the test space. The close relationship between this...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:14744884
Multigroup transport theory - 2. numerical resultsGarcia, R.D.M.; Siewert, C.E. (NC State Univ, Raleigh, USA)</br>The F/sub N/ method is used to establish particularly accurate solutions, at modest cost, for the emerging angular fluxes basic to a class of multigroup particle-transport problems. A study of the fundamental computational aspects...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:16056467
Explicit expressions for the moments of the monoenergetic neutron-transport equation for isotropic scattering in an infinite plane mediumGanapol, B.D. (Arizona Univ., Tucson (USA). Dept. of Nuclear and Energy Engineering)</br>The analytical determination of the spatial and angular moments of the monoenergetic neutron-transport equation in an infinite medium initiated by Case, de Hoffmann and Placzek 30 years ago is completed. Analytical expressions for...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:8306948
State-space splitting in Monte Carlo calculationsMacdonald, J.L.; Cashwell, E.D. (Los Alamos Scientific Lab., NM)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:10428618
Criticality of neutron transport in a slab with finite reflectorsPao, C.V. (Department of Mathematics, North Carolina State University, Raleigh, North Carolina 27607)</br>The purpose of this paper is to investigate the subcriticality and the supercriticality for the neutron transport in a slab which is surrounded by two finite reflectors. The mathematical problem is to determine when the coupled boundary-value...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:40034723
Tracking techniques for the characteristics method applied to the resolution of the neutrons transport equation in multi scale domainsFevotte, F. (CEA Saclay, Dept. Modelisation de Systemes et Structures (DEN/DANS/DM2S/SERMA), 91 - Gif sur Yvette (France))</br>CEA Saclay (DEN/DANS), 91 - Gif sur Yvette (France)</br>At the various stages of a nuclear reactor's life, numerous studies are needed to guaranty the safety and efficiency of the design, analyse the fuel cycle, prepare the dismantlement, and so on. Due to the extreme difficulty to take...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18097411
Legendre expansion of neutron flux over entire multilayer slab geometryRaghav, H.P. (Bhabha Atomic Research Centre, 5th floor, Central Complex, Bombay 400 085 (India))</br>The monoenergetic integral transport equation for a multilayer slab geometry has been solved by Legendre expansion method. The method utilizes an expansion of the neutron flux over entire multilayer slab geometry in Legendre polynomials...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17053536
The asymptotic behaviour of neutron waves in small systems and its application to the investigation of solution of the neutron transport equationPerez, R.B. (Tennessee Univ., Knoxville (USA). Dept. of Nuclear Engineering); Meade, D. (Universidad Nacional Autonoma de Mexico, Mexico City. Centro de Estudios Nucleares); Ohanian, M.J. (Florida Univ., Gainesville (USA). Dept. of...</br>The neutron wave propagation method has been applied in small graphite blocks to study the axial propagation of the neutron disturbance and the transverse wave propagation. No evidence of non-asymptotic behaviour within the limits...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:11564025
An existence and uniqueness theorem for generalized birth and death processesCapasso, V. (Bari Univ. (Italy)); Paveri-Fontana, S.L. (State Univ. of New York, Buffalo (USA))</br>A generalized version of the continuous-time birth-and-death stochastic model is formulated. Employing semigroup methods, it is shown that the initial value problem subject to appropriate regularity requirements admits a unique solution...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18017137
Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutronsStevenson, B. (New Jersey Institute of Technology, Newark)</br>Literature in the area of anisotropic neutron scattering is by no means lacking. Attention, however, is usually devoted to solution of some particular neutron transport problem and the model employed is at best approximate. The present...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:16056466
The albedo problem in the case of multiple synthetic scattering taking place in a plane-symmetric slabShafiq, A.; Meyer, H.E. de; Grosjean, C.C. (Ghent Rijksuniversiteit (Belgium). Seminarie voor Wiskundige Natuurkunde)</br>The albedo problem for a slab in the presence of synthetic scattering as a substitute for anisotropic scattering is treated within the framework of time-independent one-velocity transport theory. The transport equation is transformed...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:51027608
Computation of fundamental time-eigenvalue of the neutron transport equationShen, Huayun; Zhong, Bin; Liu, Huipo (Institute of Applied Physics and Computational Mathematics, Department 6, No. 2. Fenghaodong Road, Haidian District, Beijing, 100094 (China)), E-mail: shen_huayun@iapcm.ac.cn</br>Highlights: •A modified α − k power iteration method is presented for computing the time-eigenvalue. •It is not required to provide the initial values of α for the modified method. •Computational experiences validate the validity...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:38027900
Reactor lattice transport calculationsKulikowska, T. (lnstitute of Atomic Energy, Swierk (Poland)), E-mail: eo2kul@cx1.cyf.gov.pl</br>The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:9416225
Application of invariant imbedding to neutron thermalizationLois, L.; McCreless, T.G. (Nuclear Regulatory Commission, Washington, DC)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:10473769
Three basic neutron-transport problems in spherical geometrySiewert, C.E.; Grandjean, P. (Centre d'Etudes Nucleaires de Saclay, Gif-sur-Yvette, France)</br>The method of elementary solutions and the F/sub N/ method are used to solve three basic problems concerning neutron diffusion in spheres
http://inis.iaea.org/Search/search.aspx?orig_q=RN:29041838
Parallel solution of the neutron multigroup transport equationN'kaoua, T.; Chaigneau, C.; Coulomb, F. (CEA Centre d'Etudes de Limeil, 94 - Villeneuve-Saint-Georges (France))</br>We are interested in the solution of the Multigroup Neutron Transport Equation. After the presentation of the multigroup treatment, the angular, time and space discretization, we expose the modifications that have been made in order...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:28032432
Anisotropic scattering in the variational nodal simplified spherical harmonics formulationLewis, E.E. (Northwestern Univ., Evanston, IL (United States))</br>Under the assumption of isotropic scattering, the simplified spherical harmonics method (SP<sub>N</sub>) was recently formulated in variational nodal form and implemented successfully as an option of the VARIANT code. In this paper,...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18049818
Progress in nodal methods for the solution of the neutron diffusion and transport equationsLawrence, R.D. (Argonne National Lab., IL (USA))</br>Recent progress in the development of coarse-mesh nodal methods for the numerical solution of the neutron diffusion and transport equations is reviewed. In contrast with earlier nodal simulators, more recent nodal diffusion methods...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17038310
Exact time-dependent moments for the monoenergetic neutron transport equation with isotropic scatteringGanapol, B.D. (University of Arizona, Department of Nuclear and Energy Engineering, Tuscon, Arizona)</br>Exact expressions for the moments of the time-dependent angular flux describing the transport of monoenergetic neutrons are derived. The moments are then used to reconstruct the angular flux which is compared to diffusion theory
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18000888
Neutron transport with backward-forward scatteringCassell, J.S. (City of London Polytechnic (UK))</br>An explicit expression is obtained for the neutron flux in the one-speed backward-forward scattering model in a general homogeneous convex medium. The method can be extended to energy-dependent problems via the multigroup procedure....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:38036269
Phenomenology of using semi groups for neutron transport process representationStancic, V. (Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia))</br>Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia)</br>In order to make it easier to use the semigroup tool for describing the neutron filed transport process, compared to the standard one in neutron physics, a phenomenological approach of the model is presented here. (author)
http://inis.iaea.org/Search/search.aspx?orig_q=RN:38054808
Calculation of neutron transport in plane geometry by invariant imbedding methodSimovic, R. (Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia))</br>A practical combination of invariant imbedding and transfer matrix methods was displayed in this paper. A very simple scheme for neutron transport analysis was obtained for slab materials and some results of numerical calculations...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:38108595
On the Use of Importance Sampling in Particle Transport ProblemsEriksson, B.</br>AB Atomenergi, Nykoeping (Sweden)</br>The idea of importance sampling is applied to the problem of solving integral equations of Fredholm's type. Especially Bolzmann's neutron transport equation is taken into consideration. For the solution of the latter equation, an...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:13711333
An approximate method for the calculation of transmission probabilities through hollow cylindersGilai, D. (Israel Atomic Energy Commission, Beersheba. Nuclear Research Center-Negev)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:8338091
Physical and unphysical neutron distributions in reactor calculationsShalitin, D.; Wagschal, J.J.; Yeivin, Y. (Hebrew Univ., Jerusalem (Israel). Racah Inst. of Physics)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:12640310
The isotope density inverse problem in multigroup neutron transportZazula, J.M.</br>Institute of Nuclear Physics, Krakow (Poland)</br>The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:19025873
Neutron stochastic transport theory with delayed neutronsMunoz-Cobo, J.L.; Verdu, G. (Universidad Politecnica de Valencia (Spain))</br>From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:8288832
Comparison of measurements and calculations for ORNL integral neutron scattering experiment for ironCramer, S.N.; Oblow, E.M. (Oak Ridge National Lab., TN)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:43042248
Neutron-based techniques for detection of explosives and drugsKiraly, B.; Olah, L.; Csikai, J., E-mail: csikai@falcon.phys.klte.hu</br>Systematic measurements were carried out on the possible use of elastically backscattered Pu-Be neutrons combined with the thermal neutron reflection method for the identification of land mines and illicit drugs via he detection of...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:49082851
Use of 3-D neutron-gamma transport code ATES3 for radiation shielding analysisGupta, Anurag (Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai (India)); Modak, R.S., E-mail: anurag@barc.gov.in</br>The 3-D steady-state S<sub>N</sub> neutron-gamma transport theory code ATES3 developed at BARC can be utilized for external source problems such as shielding analysis. A brief description on the use of ATES3 and its validation for...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:7252117
Generalized dispersion theory for bounded fast assembliesKraft, T.; Dorning, J. (Univ. of Illinois, Urbana)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:11554176
Escape and transmission probabilities in cylindrical geometryBjerke, M.A.</br>Oak Ridge National Lab., TN (USA)</br>An improved technique for the generation of escape and transmission probabilities in cylindrical geometry was applied to the existing resonance cross section processing code ROLAIDS. The algorithm of Hwang and Toppel, [ANL-FRA-TM-118]...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:14748697
Quadrature sums of highest algebraic degree of precision for neutron transport integralsGallone, S. (Milan Univ. (Italy). Ist. di Scienze Fisiche)</br>A Gaussian-type quadrature formula for neutron transport integrals is here re-established according to the orthogonal-polynomial method. Limit properties of the asymptotic and nonasymptotic parts of the quadrature sum are also obtained,...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:44032662
Application of MCNP in multigroup neutron transport calculationHuang Weibing; Chen Lunshou (Sanmen Nuclear Power Company, Taizhou (China)); Wu Hongchun (Xi'an Jiaotong Univ., Xi'an (China))</br>The details about how to use MCNP to solve multigroup transport problem is introduced. Some critical and fixed source cases are calculated by using MCNP, the results are in good agreement with the reference values. The successful...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:14733043
A finite element method for neutron transport - VI. Upper and lower bounds for local characteristics of solutionsAckroyd, R.T. (UKAEA Risley Nuclear Power Development Establishment. Technical Services and Planning Directorate); Splawski, B.A. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering)</br>It is shown that the finite element method also shares with Monte Carlo the capability to bracket local characteristics of a solution, such as the reaction rate for a small locality. The bracketing bounds for the Monte Carlo method...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:28028954
Improved iteration strategy for the simplified P<sub>N</sub> equationsBrantley, P.S.; Larsen, E.W. (Univ. of Michigan, Ann Arbor, MI (United States))</br>The P<sub>N</sub> approximation to the neutron transport equation has been well established for decades. Gelbard et al. have proposed the FLIP formulation of the planar P<sub>N</sub> equations, consisting of a system of coupled diffusion...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18063719
Analytical solution to the hybrid diffusion-transport equationNanneh, M.M.; Williams, M.M.R. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering)</br>A special integral equation was derived in previous work using a hybrid diffusion-transport theory method for calculating the flux distribution in slab lattices. In this paper an analytical solution of this equation has been carried...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:33034651
P<sub>N</sub> solutions of the time-dependent neutron transport equation with anisotropic scattering in a homogeneous sphereYildiz, Cemal (Department of Physics, Istanbul Technical University, Maslak, Istanbul (Turkey))</br>In an earlier paper, the spherical harmonics method for the solution of the time-dependent transport equation with the Marshak boundary conditions was presented in order to investigate the effect of a strongly anisotropic scattering...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:42004132
The wavelet function expansion method for solving the neutron transport equation - 027Hongchun, Wu; Liangzhi, Cao; Youqi, Zheng; Weiyan, Yang (School of Nuclear Science and Technology, Xi'an Jiaotong University, Shaanxi, P.R. (China))</br>American Nuclear Society, Inc., 555 N. Kensington Avenue, La Grange Park, Illinois 60526 (United States)</br>A coupled method of using discrete ordinate discretization for polar angle and wavelet expansion for azimuthal angle in calculating neutron transport equation is developed. And another synthetic method of using multi-group discretization...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:41119277
Computational methods and data for neutron transport: recent advances and perspectivesRavetto, Piero (Politecnico di Torino, Dipartimento di Energetica, Corso Duca degli Abruzzi, 24 - 10129 Torino (Italy))</br>Paul Scherrer Institut - PSI, 5232 Villigen PSI (Switzerland)</br>This paper presents a short review of the current status of the field of deterministic computational methods for the solution of the neutron transport equation for reactor physics applications. Some recent developments are briefly...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:25067038
Biases in Monte Carlo eigenvalue calculationsGelbard, E.M. (Argonne National Lab., IL (United States). Reactor Analysis Division); Gu, A.G. (Mississippi State Univ., Mississippi State, MS (United States). Mechanical and Nuclear Engineering)</br>The derivation of the standard expression for the Monte Carlo eigenvalue bias is reviewed. It is noted that the bias is due to the repeated normalization of the fission source by the eigenvalue. This normalization can be partially...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:12603830
Angular current approximations in neutron transport calculations using interface currentsMohanakrishnan, P. (Bhabha Atomic Research Centre, Bombay (India). Theoretical Reactor Physics Section)</br>It has been recognized recently that accurate numerical solutions of the neutron transport equation in two dimensions can be obtained by interface current approach provided the angular dependence of these currents is properly accounted...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:20043557
A nodal solution of the multigroup neutron transport equation using spherical harmonicsInanc, F.; Rohach, A.F. (Iowa State Univ. of Science and Technology, Ames (USA). Dept. of Nuclear Engineering)</br>A nodal method has been developed for solving the neutron transport equation using spherical harmonics in 2-D Cartesian coordinates for rectangular nodes. The first-order differential equations resulting from the expansion of the...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:13711351
Choice of angular current approximations for solving neutron transport problems in 2-D by interface current approachMohanakrishnan, P. (Bhabha Atomic Research Centre, Bombay (India). Theoretical Reactor Physics Section)</br>A computer programme, RICANT, for the solution of 2-D neutron transport problems is described. This uses the interface current approach, where the angular neutron currents crossing region surfaces are expanded in terms of associated...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:40005517
Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCADChen Qichang; Wu Hongchun (School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an Shaanxi 710049 (China)); Cao Liangzhi (School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an Shaanxi 710049...</br>A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:11500160
Investigation of new estimation approaches for nuclear reactor computations by Monte CarloRagheb, M.M.H.</br>Wisconsin Univ., Madison (USA)</br>New estimation approaches for nuclear reactor calculations by Monte Carlo are developed and investigated, with the purpose of surmounting or alleviating existing difficulties facing the application of the Monte Carlo Method. Bias-free...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:5150810
Additonal few-group and multigroup calculations of neutron penetrationShure, K. (Westinghouse Electric Corp. West Mifflin, PA)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:20046195
On the neutron-noise transmission studies for non-multiplying media using transport theoryJena, A.K.; Singh, O.P. (Indira Gandhi Centre for Atomic Research, Kalpakkam (India))</br>This paper reports the results of our investigations on the neutron-noise transmission characteristics of non-multiplying media using transport theory. The study has been carried out systematically by first considering the infinite...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:19091101
Discrete ordinates scheme for quadrilateral meshesKumar, V.; Menon, S.V.G. (Bhabha Atomic Research Centre, Bombay (India). Theoretical Physics Div.)</br>A simple extension of the discrete ordinates method for solving the transport equation with quadrilateral meshes in X-Y and R-Z geometry is described. Numerical results of some benchmark problems are presented for showing the adequacy...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:13649561
Complex eigenvalues for neutron transport equation with quadratically anisotropic scatteringSjoestrand, N.G. (Chalmers Tekniska Hoegskola, Goeteborg (Sweden). Institutionen fore Reaktorfysik)</br>Complex eigenvalues for the monoenergetic neutron transport equation in the buckling approximation have been calculated for various combinations of linearly and quadratically anisotropic scattering. The results are discussed in terms...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17067160
A group by group upscattering scaling methodSbaffoni, M.M.; Abbatte, M.J. (Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche; Universidad Nacional de Cuyo, Mendoza (Argentina))</br>A new scaling method is presented, based on a group by group rebalance of the upscattering part of the source term, that can be used to accelerate convergence in discrete-ordinate transport codes. It assures a positive correction...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18067600
A multigroup finite-element solution of the neutron transport equationWood, J.; Williams, M.M.R. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering)</br>The extension of the theory of a multigroup finite-element method of solving the neutron transport equation, to include general anisotropy of scattering and an anisotropic spatially-dependent source, is described. The method, based...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:27015034
Noise analysis using chaos theoryMontesino, Maria Elena; Valero, Esbel T. (Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)); Aguilar, Omar (Centro de Informacion de la Energia Nuclear, La Habana (Cuba))</br>Chaos theory finds application in noise signal analysis in systems of information acquisition with diagnosis value. In the present article it is examined the theoretical foundation of noise discrimination problem from a fractal dynamics...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:26078606
Asymptotic solution of a mixed problem for a neutron transport kinetic equationKucherenko, V.V. (Moskovskij Inzhenerno-Stroitel'nyj Inst., Moscow (Russian Federation))</br>Asymptotic solution of a mixed problem for the neutron transfer kinetic equation in case of smoothly changing equation coefficients and in case of fast-oscillating ones is studied.4 refs
http://inis.iaea.org/Search/search.aspx?orig_q=RN:9416218
Computational scheme for energy group boundary selection using sensitivity theoryHerrnberger, V. (Swiss Federal Inst. for Reactor Research, Wuerenlingen)</br>On the basis of sensitivity profiles, of required target accuracies and of the cost-benefit relationship of the iterative transport codes to be employed, a computational scheme for energy group boundary selection is presented. Initial...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:14783917
Mathematical foundations of the quasistatic approximation in reactor physicsMika, J. (Institute of Nuclear Research, Warsaw (Poland))</br>The quasistatic method used in reactor-physics calculations is applied to a model integro-differential equation in a Banach space. The resulting quasistatic equations are shown to be locally consistent with the original initial-value...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17011628
A multigroup finite-element solution of the neutron transport equationWood, J. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering)</br>The extension of a variational finite-element method of solving the neutron transport equation, to include multigroup-energy dependence in R-Z geometry, is evaluated. The method is implemented in a computer program called FELICIT....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:33040241
The additive angular rebalance acceleration method for solving neutron transport equations in X-Y geometryPark, Chang Je; Cho, Nam Zin (Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of))</br>The additive angular dependent rebalance (AADR) factor acceleration method proposed by the authors previously is an effective acceleration method for the the discrete ordinates neutron transport equation. For slab geometry problems,...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:14753412
Three-dimensional neutron streaming calculations using the Monte Carlo coupling techniqueUeki, K. (Univ Res Inst, Tokyo, Jpn)</br>Two three-dimensional neutron streaming problems are analyzed by using the Monte Carlo coupling technique. A description of the two three-dimensional neutron streaming problems, the procedures, and the performance of the coupling...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:7252137
Invariant imbedding analysis of particle transport incorporating orders-of-scatteringWoolf, S. (Arcon Corp., Wakefield, MA); Garth, J.C.; Filippone, W.L.</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:41030781
An improved three-dimensional wavelet-based method for solving the first-order Boltzmann transport equationZheng Youqi; Wu Hongchun (School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi 710049 (China)); Cao Liangzhi (School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi 710049...</br>A new angular discretization scheme based on the Daubechies' wavelets has been developed in recent studies. A decoupled S<sub>N</sub> and wavelet expansion method was proposed. This paper discusses the limitations and improvements...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:29002889
New algorithms for the 2D transport equationAkesbi, S.; Nicolet, M. (Haute-Alsace Univ., 68 - Mulhouse (France))</br>The aim of this work is to introduce and analyse the new algorithms for solving the transport equation in two-dimensional plane geometry. These schemes are based on a splitting method for the collision operator. Theoretical and numerical...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:35004546
Discrete version of the SHE asymptotics: multigroup neutron transport equationsGoudon, Thierry; Mellet, Antoine (Laboratoire J.A. Dieudonne, UMR 6621, Universite Nice-Sophia Antipolis, Parc Valrose, F-06108 Nice Cedex 02 (France); INRIA-Sophia, project CAIMAN (France); Mathematiques pour l'Industrie et la Physique,...</br>This paper is devoted to the derivation of multigroup diffusion equations from the Boltzmann equation. The limit system couples the energy levels from both zeroth-order term and diffusion currents
http://inis.iaea.org/Search/search.aspx?orig_q=RN:6181794
The WIMS-E module W-SMEARRoth, M.J.</br>UKAEA Reactor Group, Winfrith. Atomic Energy Establishment</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:51086586
KBA parallelization method for computing neutron transport equations on unstructured gridsNikolaeva, O.V.; Gajfulin, S.A.; Bass, L.P. (IPM im. M.V. Keldysha RAN, Moscow (Russian Federation))</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:32061665
Deterministic Transport Methods for Reactor AnalysisAdams, M.L.</br>Texas A and M University, College Station, TX (United States)</br>We discuss deterministic transport methods used today in neutronic analysis of nuclear reactors. Our goal is to provide an overview of the methods that are most widely used for analyzing light water reactors (LWRs) and that (in our...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:51003865
Optimization of neutron transport methods using parallel computingGarcía, M. (Comisión Nacional de Energía Atómica (CNEA), San Carlos de Bariloche, Río Negro (Argentina) Centro Atómico Bariloche (CAB), Dept. de Física de Reactores); Villarino, E. (INVAP S. E., San Carlos de Bariloche, Río Negro...</br>Nuclear reactor physics deals with the solution of the neutron transport equation, for which several numerical methods and resulting calculation codes exist. Using parallel programming to multiply the resources available to these...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:17089222
The heterogeneous response method applied to couple the average pin cell and bulk moderator in cluster geometryLerner, A.M. (Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes); Stamm'ler, R.J.J. (ASEA-ATOM AB, Vaesteraas (Sweden))</br>The first step towards evaluation of the neutron flux throughout a fuel cluster usually consists of obtaining the multigroup flux distribution in the average pin cell and in the circular outside system of shroud and bulk moderator....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:5096084
Singular eigenfunction expansions in neutron transport theoryMcCormick, N.J.; Kuscer, I.</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:10443574
Singularities of the solution to the neutron transport equation in cylindrical symmetryKulikowska, T. (Institute of Nuclear Research, Warsaw (Poland))</br>The singularities of the solution to the neutron transport equation for systems with cylindrical symmetry are discussed in detail. It is shown that the case of spherical symmetry can be considered as a special case of cylindrical...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:28015856
Singular perturbation analysis of the neutron transport equationLosey, D.C. (Westinghouse Savannah River Co., Aiken, SC (United States)); Lee, J.C. (Michigan Univ., Ann Arbor, MI (United States))</br>Westinghouse Savannah River Co., Aiken, SC (United States). Funding organisation: USDOE, Washington, DC (United States)</br>A singular perturbation technique is applied to the one-speed, one- dimensional neutron transport equation with isotropic scattering. Our technique extends previous singular perturbation applications to higher-order and reduces the...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:32065806
Error Analysis of Variations on Larsen's Benchmark ProblemAzmy, YY</br>Oak Ridge National Laboratory (United States). Funding organisation: US Department of Energy (United States)</br>Error norms for three variants of Larsen's benchmark problem are evaluated using three numerical methods for solving the discrete ordinates approximation of the neutron transport equation in multidimensional Cartesian geometry. The...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:8306958
Local separation of angle and energy in one-dimensional transportRakavy, G.; Tikochinsky, Y.; Wagschal, J.J. (Hebrew Univ., Jerusalem)</br>No abstract available.
http://inis.iaea.org/Search/search.aspx?orig_q=RN:23026988
Numerical solution of the equation of neutrons transport on plane geometry by analytical schemes using acceleration by synthetic diffusionAlonso-Vargas, G.</br>Instituto Politecnico Nacional, Mexico City (Mexico). Escuela Superior de Fisica y Matematicas</br>A computer program has been developed which uses a technique of synthetic acceleration by diffusion by analytical schemes. Both in the diffusion equation as in that of transport, analytical schemes were used which allowed a substantial...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:37106928
Neutron Flux Perturbations due to Infinite Plane Absorbers IV: The Exponential Flux RevisitedWilliams, M.M.R.</br>Flux depression factors and measures of asymmetry are presented for an absorbing and scattering slab in an infinite medium in which there is an overall exponential flux. One speed transport theory is employed. The effect of the slab...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:39036523
The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA versionAhnert, C.; Aragones, J. M.</br>Junta de Energia Nuclear (JEN), Madrid (Spain)</br>This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:10447475
Inverse problem for a finite slabSiewert, C.E. (Centre d'Etudes Nucleaires de Saclay, Gif-sur-Yvette, France)</br>The finite-slab inverse problem for multigroup neutron transport theory is solved
http://inis.iaea.org/Search/search.aspx?orig_q=RN:18052309
Assembly homogenization techniques for light water reactor analysisSmith, K.S. (Studsvik of America, Newton, MA)</br>Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:47001731
Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codesHussein, M.S (Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)); Lewis, B.J. (Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario...</br>The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:11500150
Buckled spherical harmonics solutions of neutron transport problemsBlomquist, R.N.; Lewis, E.E.</br>Northwestern Univ., Evanston, IL (USA). Technological Inst</br>The within-group even-parity neutron transport equation is formulated with complex angular and spatial trial functions and with a complex buckling approximation. Three angular trial functions are compared; finite elements, discrete...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:19078348
A perturbation theory for the non-linear system of neutron transport and burnup equationsAlp, C. (Istanbul Technical Univ. (Turkey). Inst. for Nuclear Energy)</br>A perturbation theory for use in nuclear reactor burnup analysis is derived. An important characteristic function is defined, and the related adjoint equation is obtained simply by using the variational principle. The adjoint matrix...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:37039231
On the solution of time-dependent transport equation with time-varying cross sectionsAbdou, M.A. (Theoretical Research Group, Physics Department, Faculty of Science, Mansoura University, Mansoura 35516 (Egypt)), E-mail: m_abdou_eg@yahoo.com</br>The time-dependent neutron transport equation in an infinite medium with time-varying cross sections has been solved by means of two techniques namely, flux-limited approach and maximum entropy method. The behaviour of the distribution...
http://inis.iaea.org/Search/search.aspx?orig_q=RN:45100448
Solutions of one-speed neutron transport equation for spherical geometrySharma, Anuradha Jagmohan (Theoretical Physics Division, Bhabha Atomic Research Centre, Mumbai (India))</br>University of Mumbai, Mumbai (India)</br>The transport equation is used for radiation shielding and reactor core calculations in reactor physics, radiative transfer analysis of stellar and planetary atmospheres, the theory of plasmas, the theory of sound propagation etc....
http://inis.iaea.org/Search/search.aspx?orig_q=RN:12603875
On the solution of the dispersion equation for monoenergetic neutron transport with linearly anisotropic scatteringProtopopescu, V.; Sjoestrand, N.G. (Chalmers Tekniska Hoegskola, Goeteborg (Sweden). Institutionen foer Reaktorfysik)</br>The eigenvalues occurring in the stationary or time dependent, monoenergetic Boltzmann equation with linearly anisotropic scattering are investigated. A detailed analysis is made of the number, nature and behaviour of the eigenvalues...