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Harvey, J.T.

Arkansas Univ., Little Rock (USA)

Arkansas Univ., Little Rock (USA)

AbstractAbstract

[en] 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 irradiation port (Glory Hole) and with a one-half inch section of aluminum in place are reported. Neutron transport calculations are also performed for the above mentioned spectra and for six other geometrics reported elsewhere. The absolute values of the spectral parameters derived from calculations do not agree with the spectral parameters obtained by the foil activation method, but the same trends are observed. Transport calculation results are combined with the foil activation results to give ''best value'' spectral parameters. It was found that the free-field spectrum was moderately hard and the Glory Hole spectra to be softer by approximately 18 percent when compared to the free-field spectrum. The experimenters table and two sections of aluminum of differing thicknesses were found to have no major effect on the neutron spectrum when comparing spectral parameters. The neutron spectrum behind a section on plexiglas was found to be considerably harder than the free-field spectrum. The differences between the results of the neutron transport calculations and the foil activation method are discussed

Primary Subject

Source

1977; 152 p; University Microfilms Order No. 77-23,339; Thesis (Ph. D.).

Record Type

Report

Literature Type

Thesis/Dissertation

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

Cardona, Augusto V.

Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Escola de Engenharia

Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Escola de Engenharia

AbstractAbstract

[en] In this work it is presented a generic method of analytical solution to the one-dimensional S

_{N}, P_{N}, W_{N}, Ch_{N}, A_{N}and L D_{N}approximations of the linear transport equation. The main idea of this method consists in the application of the Laplace transform to solve the differential equation system related to the considered approximations and solution of the resultant algebraic system by Trzaska's algorithm. (author). 46 refs., 3 figs., 15 tabsOriginal Title

Metodo generico de solucao analitica para aproximacoes da equacao linear de transporte

Primary Subject

Source

May 1996; 96 p; Available from the Library of Brazilian Nuclear Energy Commission, Rio de Janeiro; Tese (Ph.D.).

Record Type

Miscellaneous

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Thesis/Dissertation

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LanguageLanguage

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INIS VolumeINIS Volume

INIS IssueINIS Issue

Ragheb, M.M.H.

Wisconsin Univ., Madison (USA)

Wisconsin Univ., Madison (USA)

AbstractAbstract

[en] 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 estimators which extract more information on a given particle tract than currently used estimators are deduced. These depend on estimating separately the different Neumann series terms of the solution to the Boltzmann Transport Equation, and are based on underlying absorbing or nonabsorbing Markov Chain random walk models. They are applied to representative slowing-down and deep penetration problems. Effective biases arising from application of currently used methods are highly suppressed. Comparison to analytical solutions and to currently used methods shows their distinctive advantages. An approach for the systematic determination of optimal biasing parameters in importance sampling calculations, depending on particle tracks scaling; which avoids the effective biases and infinite variances in these calculations, is developed and applied to the deep penetration problem of particle transport. The use of stationary functionals from variational theory for variance reduction in Monte Carlo calculations is discussed. Suggestions for alternative error estimation methods in Monte Carlo calculations, based on functionals of the second moment for the collision and last collision estimators, or on reconstructing the sample distributions from sample moments, are exposed. Implementation of the suggested new approaches in future Monte Carlo calculations is discussed

Primary Subject

Secondary Subject

Source

1978; 503 p; University Microfilms Order No. 78-23,084; Thesis (Ph. D.).

Record Type

Report

Literature Type

Thesis/Dissertation

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

Alonso-Vargas, G.

Instituto Politecnico Nacional, Mexico City (Mexico). Escuela Superior de Fisica y Matematicas

Instituto Politecnico Nacional, Mexico City (Mexico). Escuela Superior de Fisica y Matematicas

AbstractAbstract

[en] A computer program has been developed which uses a technique of synthetic acceleration by diffusion by analytical schemes. Both in the diffusion equation as in that of transport, analytical schemes were used which allowed a substantial time saving in the number of iterations required by source iteration method to obtain the K

_{e}ff. The program developed ASD (Synthetic Diffusion Acceleration) by diffusion was written in FORTRAN and can be executed on a personal computer with a hard disc and mathematical O-processor. The program is unlimited as to the number of regions and energy groups. The results obtained by the ASD program for K_{e}ff is nearly completely concordant with those of obtained utilizing the ANISN-PC code for different analytical type problems in this work. The ASD program allowed obtention of an approximate solution of the neutron transport equation with a relatively low number of internal reiterations with good precision. One of its applications would be in the direct determinations of axial distribution neutronic flow in a fuel assembly as well as in the obtention of the effective multiplication factor. (Author)Original Title

Solucion numerica de la ecuacion de transporte de neutrones en geometria plana mediante esquemas analiticos empleando aceleracion por difusion sintetica

Primary Subject

Source

1991; 136 p; Thesis (M. Sc.).

Record Type

Miscellaneous

Literature Type

Thesis/Dissertation

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Country of publication

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INIS VolumeINIS Volume

INIS IssueINIS Issue

Sharma, Anuradha Jagmohan

University of Mumbai, Mumbai (India)

University of Mumbai, Mumbai (India)

AbstractAbstract

[en] 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. The earliest investigations were carried out in the theory of radiative transfer. Then with the advent of neutron fission as a source of energy, this theory was applied for reactor physics calculation

Primary Subject

Source

Feb 2000; 153 p; University of Mumbai; Mumbai (India); Available from University of Mumbai, Mumbai (IN); 58 refs.; Thesis (Ph.D.)

Record Type

Miscellaneous

Literature Type

Thesis/Dissertation

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

Lieberoth, J.

Technische Hochschule Aachen (F.R. Germany). Fakultaet fuer Maschinenwesen

Technische Hochschule Aachen (F.R. Germany). Fakultaet fuer Maschinenwesen

AbstractAbstract

No abstract available

Original Title

Ein Beitrag zur Monte Carlo-Methode in der Reaktortheorie

Primary Subject

Source

30 Jun 1972; 183 p; 19 figs. With abstract. Available from the library of TH Aachen.; Habil. Schr.

Record Type

Report

Literature Type

Thesis/Dissertation

Country of publication

Publication YearPublication Year

Reference NumberReference Number

INIS VolumeINIS Volume

INIS IssueINIS Issue

Song, Jae Seung

Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

AbstractAbstract

[en] Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

Primary Subject

Secondary Subject

Source

Feb 1992; 45 p; Available from Korea Advanced Institute of Science and Technology, Daejeon (KR); Thesis (Mr. Eng.)

Record Type

Miscellaneous

Literature Type

Thesis/Dissertation

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INIS VolumeINIS Volume

INIS IssueINIS Issue

Flores Calderon, J.E.

Instituto Politecnico Nacional, Mexico City. Escuela Superior de Fisica y Matematicas

Instituto Politecnico Nacional, Mexico City. Escuela Superior de Fisica y Matematicas

AbstractAbstract

[en] Experiments with non-stationary neutron transport in large cavity moderators (l>>Σsub(tr)

^{-1}) (where l is the characteristic cavity length and Σsub(tr)^{-1}the macroscopic transport section of the moderator) led to the method reported in this study which, based on neutron impulses for determining albedo of thermal neutrons, gave a precision greater by an order of magnitude over previous methods. A sufficient time interval after introduction of the neutron flux into the moderator chamber decreased exponentially the decay constant L, which was itself related to albedo by a function called f. Numerical calculations of albedo were assisted. (author)Original Title

Determinacion del albedo por el metodo del impulso neutronico

Primary Subject

Source

1982; 85 p; Tesis (M. in Sci.).

Record Type

Miscellaneous

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Thesis/Dissertation

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INIS VolumeINIS Volume

INIS IssueINIS Issue

AbstractAbstract

[en] The principle of the curved neutron guide is to transport neutrons far away from the reactor core with as minimum particle loss as possible. After a series of total reflection,the neutron beam is no longer visible from the reactor core and consequently, gamma radiations and fast neutrons emitted from the core are scattered by the walls of the guide and absorbed by the biological shielding set around the guide. The curved neutron guide provides a high-quality beam of slow neutrons. The first chapter deals with the theoretical concept of curved guide, we have determined the parameters for the setting of such a guide in the EL3 reactor at Saclay (France). The different tolerances on the state the surface, on the alignment of the different parts of the guide, on the waving of the guide wall have been assessed. The second chapter presents the technical solution chosen that complies to all the required specifications. The curved neutron guide has been designed for neutrons with wavelength of 4 Angstroms, it is 29 m long, has a bending radius of 835 m and is composed of 87 rectangular components made of glass plates on which a 1500 angstrom thick layer of nickel has been deposited. Each component is set with a fixed angle of (4±0.25)*10

^{-4}radians from the previous component in order to form the bending radius. The last chapter is dedicated to the neutron flux measurement made at the end of the neutron guideOriginal Title

Guide courbe conducteur de neutrons

Primary Subject

Source

4 Mar 1969; 110 p; 8 refs.; Available from the INIS Liaison Officer for France, see the 'INIS contacts' section of the INIS website for current contact and E-mail addresses: http://www.iaea.org/inis/Contacts/; These Docteur de l'Universite, Mention Sciences

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Report

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Thesis/Dissertation

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AbstractAbstract

[en] A modular nodal method is developed for solving the neutron transport equation by using the spherical harmonics approximation in two-dimensional Cartesian coordinates. The spherical harmonics approximation is based upon the second order even-parity form of the neutron transport equation. The boundary conditions of the spherical harmonics approximation are manipulated to have the forms analogous to the partial currents in the neutron diffusion equation. The relationships are developed for generating both the second order spherical harmonic equations and the boundary conditions in an automatic manner. The nodal method developed is based upon a least squares minimization technique. In that method, the spherical harmonic moments are expanded into fourth order Legendre polynomials. While some of the unknown coefficients are determined through the equations provided by the minimization scheme, the others are obtained through implementation of the boundary conditions in an integral sense. The order of the P

_{n}approximation in the nodes are determined by the developed scheme in automatic manner. The distribution of the approximations orders may be different in different parts of the problem domainPrimary Subject

Source

1989; 127 p; Iowa State Univ; Ames, IA (USA); University Microfilms, PO Box 1764, Ann Arbor, MI 48106, Order No.89-20,145; Thesis (Ph. D.).

Record Type

Miscellaneous

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Thesis/Dissertation

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