Dissertations / Theses on the topic 'Neutron transport problem'

Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions. Classical tomography fails to reconstruct the optical properties of thick scattering objects because it does not adequately account for the scattering component of the neutron beam intensity exiting the sample. We proposed a new method of computed tomography which employs an inverse problem analysis of both the transmitted and scattered images generated from a beam passing through an optically thick object. This inverse problem makes use of a computationally efficient, two-dimensional forward problem based on neutron transport theory that effectively calculates the detector readings around the edges of an object. The forward problem solution uses a Step-Characteristic (SC) code with known uncollided source per cell, zero boundary flux condition and Sn discretization for the angular dependence. The calculation of the uncollided sources is performed by using an accurate discretization scheme given properties and position of the incoming beam and beam collimator. The detector predictions are obtained considering both the collided and uncollided components of the incoming radiation. The inverse problem is referred as an optimization problem. The function to be minimized, called an objective function, is calculated as the normalized-squared error between predicted and measured data. The predicted data are calculated by assuming a uniform distribution for the optical properties of the object. The objective function depends directly on the optical properties of the object; therefore, by minimizing it, the correct property distribution can be found. The minimization of this multidimensional function is performed with the Polack Ribiere conjugate-gradient technique that makes use of the gradient of the function with respect to the cross sections of the internal cells of the domain. The forward and inverse models have been successfully tested against numerical results obtained with MCNP (Monte Carlo Neutral Particles) showing excellent agreements. The reconstructions of several objects were successful. In the case of a single intrusion, TNTs (Tomography Neutron Transport using Scattering) was always able to detect the intrusion. In the case of the double body object, TNTs was able to reconstruct partially the optical distribution. The most important defect, in terms of gradient, was correctly located and reconstructed. Difficulties were discovered in the location and reconstruction of the second defect. Nevertheless, the results are exceptional considering they were obtained by lightening the object from only one side. The use of multiple beams around the object will significantly improve the capability of TNTs since it increases the number of constraints for the minimization problem.

This thesis studies the so-called “criticality problem”, an important generalised eigenvalue problem arising in neutron transport theory. The smallest positive real eigenvalue of the problem contains valuable information about the status of the fission chain reaction in the nuclear reactor (i.e. the criticality of the reactor), and thus plays an important role in the design and safety of nuclear power stations. Because of the practical importance, efficient numerical methods to solve the criticality problem are needed, and these are the focus of this thesis. In the theory we consider the time-independent neutron transport equation in the monoenergetic homogeneous case with isotropic scattering and vacuum boundary conditions. This is an unsymmetric integro-differential equation in 5 independent variables, modelling transport, scattering, and fission, where the dependent variable is the neutron angular flux. We show that, before discretisation, the nonsymmetric eigenproblem for the angular flux is equivalent to a related eigenproblem for the scalar flux, involving a symmetric positive definite weakly singular integral operator(in space only). Furthermore, we prove the existence of a simple smallest positive real eigenvalue with a corresponding eigenfunction that is strictly positive in the interior of the reactor. We discuss approaches to discretise the problem and present discretisations that preserve the underlying symmetry in the finite dimensional form. The thesis then describes methods for computing the criticality in nuclear reactors, i.e. the smallest positive real eigenvalue, which are applicable for quite general geometries and physics. In engineering practice the criticality problem is often solved iteratively, using some variant of the inverse power method. Because of the high dimension, matrix representations for the operators are often not available and the inner solves needed for the eigenvalue iteration are implemented by matrix-free inneriterations. This leads to inexact iterative methods for criticality computations, for which there appears to be no rigorous convergence theory. The fact that, under appropriate assumptions, the integro-differential eigenvalue problem possesses an underlying symmetry (in a space of reduced dimension) allows us to perform a systematic convergence analysis for inexact inverse iteration and related methods. In particular, this theory provides rather precise criteria on how accurate the inner solves need to be in order for the whole iterative method to converge. The theory is illustrated with numerical examples on several test problems of physical relevance, using GMRES as the inner solver. We also illustrate the use of Monte Carlo methods for the solution of neutron transport source problems as well as for the criticality problem. Links between the steps in the Monte Carlo process and the underlying mathematics are emphasised and numerical examples are given. Finally, we introduce an iterative scheme (the so-called “method of perturbation”) that is based on computing the difference between the solution of the problem of interest and the known solution of a base problem. This situation is very common in the design stages for nuclear reactors when different materials are tested, or the material properties change due to the burn-up of fissile material. We explore the relation ofthe method of perturbation to some variants of inverse iteration, which allows us to give convergence results for the method of perturbation. The theory shows that the method is guaranteed to converge if the perturbations are not too large and the inner problems are solved with sufficiently small tolerances. This helps to explain the divergence of the method of perturbation in some situations which we give numerical examples of. We also identify situations, and present examples, in which the method of perturbation achieves the same convergence rate as standard shifted inverse iteration. Throughout the thesis further numerical results are provided to support the theory.

[ES] Uno de los objetivos más importantes en el análisis de la seguridad en el campo de la ingeniería nuclear es el cálculo, rápido y preciso, de la evolución de la potencia dentro del núcleo del reactor. La distribución de los neutrones se puede describir a través de la ecuación de transporte de Boltzmann. La solución de esta ecuación no puede obtenerse de manera sencilla para reactores realistas, y es por ello que se tienen que considerar aproximaciones numéricas. En primer lugar, esta tesis se centra en obtener la solución para varios problemas estáticos asociados con la ecuación de difusión neutrónica: los modos lambda, los modos gamma y los modos alpha. Para la discretización espacial se ha utilizado un método de elementos finitos de alto orden. Diversas características de cada problema espectral se analizan y se comparan en diferentes reactores. Después, se investigan varios métodos de cálculo para problemas de autovalores y estrategias para calcular los problemas algebraicos obtenidos a partir de la discretización espacial. La mayoría de los trabajos destinados a la resolución de la ecuación de difusión neutrónica están diseñados para la aproximación de dos grupos de energía, sin considerar dispersión de neutrones del grupo térmico al grupo rápido. La principal ventaja de la metodología que se propone es que no depende de la geometría del reactor, del tipo de problema de autovalores ni del número de grupos de energía del problema. Tras esto, se obtiene la solución de las ecuaciones estacionarias de armónicos esféricos. La implementación de estas ecuaciones tiene dos principales diferencias respecto a la ecuación de difusión neutrónica. Primero, la discretización espacial se realiza a nivel de pin. Por tanto, se estudian diferentes tipos de mallas. Segundo, el número de grupos de energía es, generalmente, mayor que dos. De este modo, se desarrollan estrategias a bloques para optimizar el cálculo de los problemas algebraicos asociados. Finalmente, se implementa un método modal actualizado para integrar la ecuación de difusión neutrónica dependiente del tiempo. Se presentan y comparan los métodos modales basados en desarrollos en función de los diferentes modos espaciales para varios tipos de transitorios. Además, también se desarrolla un control de paso de tiempo adaptativo, que evita la actualización de los modos de una manera fija y adapta el paso de tiempo en función de varias estimaciones del error.
[CAT] Un dels objectius més importants per a l'anàlisi de la seguretat en el camp de l'enginyeria nuclear és el càlcul, ràpid i precís, de l'evolució de la potència dins del nucli d'un reactor. La distribució dels neutrons pot modelar-se mitjançant l'equació del transport de Boltzmann. La solució d'aquesta equació per a un reactor realístic no pot obtenir's de manera senzilla. És per això que han de considerar-se aproximacions numèriques. En primer lloc, la tesi se centra en l'obtenció de la solució per a diversos problemes estàtics associats amb l'equació de difusió neutrònica: els modes lambda, els modes gamma i els modes alpha. Per a la discretització espacial s'ha utilitzat un mètode d'elements finits d'alt ordre. Algunes de les característiques dels problemes espectrals s'analitzaran i es compararan per a diferents reactors. Tanmateix, diversos solucionadors de problemes d'autovalors i estratègies es desenvolupen per a calcular els problemes obtinguts de la discretització espacial. La majoria dels treballs per a resoldre l'equació de difusió neutrònica estan dissenyats per a l'aproximació de dos grups d'energia i sense considerar dispersió de neutrons del grup tèrmic al grup ràpid. El principal avantatge de la metodologia exposada és que no depèn de la geometria del reactor, del tipus de problema d'autovalors ni del nombre de grups d'energia del problema. Seguidament, s'obté la solució de les equacions estacionàries d'harmònics esfèrics. La implementació d'aquestes equacions té dues principals diferències respecte a l'equació de difusió. Primer, la discretització espacial es realitza a nivell de pin a partir de l'estudi de diferents malles. Segon, el nombre de grups d'energia és, generalment, major que dos. D'aquesta forma, es desenvolupen estratègies a blocs per a optimitzar el càlcul dels problemes algebraics associats. Finalment, s'implementa un mètode modal amb actualitzacions dels modes per a integrar l'equació de difusió neutrònica dependent del temps. Es presenten i es comparen els mètodes modals basats en l'expansió dels diferents modes espacials per a diversos tipus de transitoris. A més a més, un control de pas de temps adaptatiu es desenvolupa, evitant l'actualització dels modes d'una manera fixa i adaptant el pas de temps en funció de vàries estimacions de l'error.
[EN] One of the most important targets in nuclear safety analyses is the fast and accurate computation of the power evolution inside of the reactor core. The distribution of neutrons can be described by the neutron transport Boltzmann equation. The solution of this equation for realistic nuclear reactors is not straightforward, and therefore, numerical approximations must be considered. First, the thesis is focused on the attainment of the solution for several steady-state problems associated with neutron diffusion problem: the $\lambda$-modes, the $\gamma$-modes and the $\alpha$-modes problems. A high order finite element method is used for the spatial discretization. Several characteristics of each type of spectral problem are compared and analyzed on different reactors. Thereafter, several eigenvalue solvers and strategies are investigated to compute efficiently the algebraic eigenvalue problems obtained from the discretization. Most works devoted to solve the neutron diffusion equation are made for the approximation of two energy groups and without considering up-scattering. The main property of the proposed methodologies is that they depend on neither the reactor geometry, the type of eigenvalue problem nor the number of energy groups. After that, the solution of the steady-state simplified spherical harmonics equations is obtained. The implementation of these equations has two main differences with respect to the neutron diffusion. First, the spatial discretization is made at level of pin. Thus, different meshes are studied. Second, the number of energy groups is commonly bigger than two. Therefore, block strategies are developed to optimize the computation of the algebraic eigenvalue problems associated. Finally, an updated modal method is implemented to integrate the time-dependent neutron diffusion equation. Modal methods based on the expansion of the different spatial modes are presented and compared in several types of transients. Moreover, an adaptive time-step control is developed that avoids setting the time-step with a fixed value and it is adapted according to several error estimations.
Carreño Sánchez, AM. (2020). Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/144771
TESIS

The goal of this thesis is to build hybrid deterministic/Monte Carlo algorithms for solving the neutron transport equation and associated k-eigenvalue problem. We begin by introducing and deriving the transport equation before discussing a series of deterministic methods for solving the transport equation. To begin we consider moment-based acceleration techniques for both the one and two-dimensional fixed source problems. Once this machinery has been developed, we will apply similar techniques for computing the dominant eigenvalue of the neutron transport equation. We'll motivate the development of hybrid methods by describing the deficiencies of deterministic methods before describing Monte Carlo methods and their advantages. We conclude the thesis with a chapter describing the detailed implementation of hybrid methods for both the fixed-source and k-eigenvalue problem in both one and two space dimensions. We'll use a series of test problems to demonstrate the effectiveness of these algorithms before hinting at some possible areas of future work.

In this work, an analytical approach is used along with nodal schemes for the solution of xed source two-dimensional neutron transport problems, in Cartesian geometry, de ned in heterogeneous medium, with anisotropic scattering. The methodology is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric angular quadrature set. One-dimensional equations for the averaged angular uxes are obtained by transverse integration of the original problem. Such equations are solved by the ADO method. Explicit expressions in spatial variables are derived for averaged uxes in each region in which the domain is subdivided. The solution in each region is coupled with that of its neighbouring regions to provide the solution in the whole domain, without resorting to using iterative methods. As usual in nodal schemes, auxiliary equations are needed. Here two di erent treatments were given to this issue: one based on relations between the unknown ows in the contours of the regions and the average angular uxes, and another in which these ows are approximated by polynomials of order zero being in this case, incorporated into the source term. Numerical results were compared with available literature showing the solution preserve the computational e ciency which has been a good feature of the ADO method when applied to different problems.
Neste trabalho, uma abordagem analítica é utilizada juntamente com esquemas nodais na resolução de problemas bidimensionais de transporte de nêutrons de fonte fixa, em geometria cartesiana, definidos em meio heterogêneo, com espalhamento anisotrópico. A metodologia proposta é desenvolvida a partir da versão em ordenadas discretas da equação de transporte bidimensional, juntamente com o esquema de quadratura simétrica de nível. As equações em ordenadas discretas são integradas transversalmente, originando equações unidimensionais para os fluxos angulares médios. Tais equações unidimensionais são resolvidas pelo método ADO (Analytical Discrete Ordinates). Expressões explícitas nas variáveis espaciais são derivadas para os fluxos angulares médios em cada região em que o domínio foi subdividido. A solução em cada região é acoplada às regiões vizinhas, para fornecer a solução no domínio todo, sem a utilização de métodos iterativos. Como usual em esquemas nodais, equações auxiliares são necessárias, recebendo neste estudo dois tratamentos distintos: um em que os fluxos desconhecidos nos contornos das regiões assumem relações de proporcionalidade, com os fluxos angulares médios; e, outro, em que esses fluxos são aproximados por polinômios de ordem zero sendo, nesse caso, incorporados ao termo fonte. Resultados numéricos obtidos e comparados com disponíveis na literatura mostram a viabilidade da formulação, mantendo a eficiência computacional já verificada no tratamento de outros problemas, com o uso do método ADO.

A Variational Transport Theory Method for Two-Dimensional Reactor Core Calculations Scott W. Mosher 110 Pages Directed by Dr. Farzad Rahnema It seems very likely that the next generation of reactor analysis methods will be based largely on neutron transport theory, at both the assembly and core levels. Signifi-cant progress has been made in recent years toward the goal of developing a transport method that is applicable to large, heterogeneous coarse-meshes. Unfortunately, the ma-jor obstacle hindering a more widespread application of transport theory to large-scale calculations is still the computational cost. In this dissertation, a variational heterogeneous coarse-mesh transport method has been extended from one to two-dimensional Cartesian geometry in a practical fashion. A generalization of the angular flux expansion within a coarse-mesh was developed. This allows a far more efficient class of response functions (or basis functions) to be employed within the framework of the original variational principle. New finite element equations were derived that can be used to compute the expansion coefficients for an individual coarse-mesh given the incident fluxes on the boundary. In addition, the non-variational method previously used to converge the expansion coefficients was developed in a new and more thorough manner by considering the implications of the fission source treat-ment imposed by the response expansion. The new coarse-mesh method was implemented for both one and two-dimensional (2-D) problems in the finite-difference, multigroup, discrete ordinates approximation. An efficient set of response functions was generated using orthogonal boundary conditions constructed from the discrete Legendre polynomials. Several one and two-dimensional heterogeneous light water reactor benchmark problems were studied. Relatively low-order response expansions were used to generate highly accurate results using both the variational and non-variational methods. The expansion order was found to have a far more significant impact on the accuracy of the results than the type of method. The varia-tional techniques provide better accuracy, but at substantially higher computational costs. The non-variational method is extremely robust and was shown to achieve accurate re-sults in the 2-D problems, as long as the expansion order was not very low.

A new solution technique is derived for the time-dependent transport equation. This approach extends the steady-state coarse-mesh transport method that is based on global-local decompositions of large (i.e. full-core) neutron transport problems. The new method is based on polynomial expansions of the space, angle and time variables in a response-based formulation of the transport equation. The local problem (coarse mesh) solutions, which are entirely decoupled from each other, are characterized by space-, angle- and time-dependent response functions. These response functions are, in turn, used to couple an arbitrary sequence of local problems to form the solution of a much larger global problem. In the current work, the local problem (response function) computations are performed using the Monte Carlo method, while the global (coupling) problem is solved deterministically. The spatial coupling is performed by orthogonal polynomial expansions of the partial currents on the local problem surfaces, and similarly, the timedependent response of the system (i.e. the time-varying flux) is computed by convolving the time-dependent surface partial currents and time-dependent volumetric sources against pre-computed time-dependent response kernels.

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Nesta dissertação propomos a utilização do método espectro-nodal SGF, cf. spectral Greens function, para transporte SN de partículas neutras, para determinarmos os fluxos angulares nas interfaces das regiões homogêneas do domínio espacial heterogêneo, com espalhamento linearmente anisotrópico usando preferencialmente altas ordens de quadraturas angulares. As reconstruções espaciais analíticas dos fluxos angulares são feitas no interior das regiões homogêneas, determinando as constantes arbitrárias da solução analítica local das equações SN no interior dos nodos espaciais da grade de dicretização. A seguir, utilizando essas constantes, determinamos as expressões do fluxo escalar e da corrente de nêutrons, que são substituídas na equação de transporte unidimensional em geometria retangular Cartesiana no termo de fonte por espalhamento linearmente anisotrópico. Resolvemos analiticamente a equação de transporte com os termos do fluxo escalar e corrente de nêutrons assim aproximados para estimarmos o perfil do fluxo angular de nêutrons no domínio. Esta reconstrução analítica aproximada da solução da equação de transporte de partículas neutras em geometria unidimensional Cartesiana constitui um problema inverso, na medida em que a partir da solução nodal de malha grossa fazemos primeiramente uma reconstrução analítica espacial do fluxo angular nas direções das ordenadas discretas, para em seguida procedermos à reconstrução analítica aproximada do fluxo no domínio angular.
We describe the application of the spectral Greens function SN nodal method for one-speed neutral particle transport calculations to determine the angular fluxes at the homogeneized regions within heterogeneous domains, for linearly anisotropic scattering, using preferably high-order angular quadratures. The reconstruction scheme in the space variable of the angular flux is carried out within the homogenized regions using uniform spatial grid. We determine the arbitrary constants of the analytical SN general solution inside each spatial node. Then, we determine the SN expression for the scalar flux and total current that we substitute into the analytical slab-geometry transport equation, precisely into its linearly anisotropic scattering source term. Further, we solve analytically the slab-geometry transport equation, so approximated, to obtain the flux profile within the space and angular domains. This approximate analytical reconstruction scheme of the solution of the neutral particle transport equation in slab geometry is an inverse problem, in the sense that we use accurate coarse-mesh SN numerical solution, to recover the SN analytical solution in the space variable, and then reconstruct the solution approximately in the angular domain.

Nesta tese mostraremos uma aplicação do método SGF, cf., spectral Greens function, para gerar fluxos angulares nas interfaces dos nodos na formulação de ordenadas discretas (SN) para equação de transporte de partículas neutras em geometria cartesiana unidimensional e a uma velocidade, usando quadraturas de alta ordem. Utilizando o método de malha grossa SGF, primeiramente determinamos as constantes arbitrárias da solução geral analítica das equações SN em cada nodo espacial. Então usamos a fórmula de quadratura angular para estimar uma expressão para o fluxo escalar de nêutrons e substituímos no termo de fonte de espalhamento isotrópico na equação de transporte. Resolvemos analiticamente a equação unidimensional de transporte de nêutrons com a fonte de espalhamento aproximada desta maneira e geramos os valores para o fluxo angular no interior de cada nodo espacial. Como o método SGF gera soluções numéricas de malha grossa completamente livres de erro de truncamento espacial, esperamos que o esquema de reconstrução analítica proposto tenha alta precisão para os fluxos angulares, considerando as condições de continuidade nas interfaces dos nodos do domínio espacial. Esta técnica caracteriza um problema inverso, pois a partir da solução de malha grossa do método SGF, podemos reconstruir o fluxo angular de nêutrons em qualquer ponto do domínio.
We offer in this work an application of the SGF method, cf., spectral Greens function, to generate the angular fluxes at the region interfaces of multilayer slabs, using high-order angular quadrature sets in the one-speed discrete ordinates (SN) formulation of the neutral particle transport equation. Using the SGF coarse-mesh numerical solution, we first determine the arbitrary constants of the analytical general solution of the SN equations within each spatial node. Then, we use the angular quadrature formula to estimate the expression of the scalar flux distribution, that we substitute into the isotropic scattering source term of the transport equation. We solve analytically the slab-geometry transport equation, with the scattering source so approximated, in order to generate the angular flux profile within each spatial node. As the SGF method generates coarse-mesh numerical solution, which is completely free from spatial truncation errors, we expect that the offered approximate analytical reconstruction scheme be accurate enough for the localized angular flux distribution, considering the node interface continuity conditions within the domain. This technique is thought of as an inverse problem since from the SGF coarse-mesh nodal solution we are able to reconstruct the angular flux profile at any point inside the domain.

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Um método espectronodal é desenvolvido para problemas de transporte de partículas neutras de fonte fixa, multigrupo de energia em geometria cartesiana na formulação de ordenadas discretas (SN). Para geometria unidimensional o método espectronodal multigrupo denomina-se método spectral Greens function (SGF) com o esquema de inversão nodal (NBI) que converge solução numérica para problemas SN multigrupo em geometria unidimensional, que são completamente livre de erros de truncamento espacial para ordem L de anisotropia de espalhamento desde que L < N. Para geometria X; Y o método espectronodal multigrupo baseia-se em integrações transversais das equações SN no interior dos nodos de discretização espacial, separadamente nas direções coordenadas x e y. Já que os termos de fuga transversal são aproximados por constantes, o método nodal resultante denomina-se SGF-constant nodal (SGF-CN), que é aplicado a problemas SN multigrupo de fonte fixa em geometria X; Y com espalhamento isotrópico. Resultados numéricos são apresentados para ilustrar a eficiência dos códigos SGF e SGF-CN e a precisão das soluções numéricas convergidas em cálculos de malha grossa.
A spectral nodal method is described for neutral particle energy multigroup fixed-source transport problems in cartesian geometry in the discrete ordinates (SN) formulation. For slab geometry the offered multigroup spectral nodal method is referred to as the spectral Greens function (SGF) method with the one-node block inversion (NBI) iterative scheme, which converges numerical solutions to multigroup slab-geometry SN problems, that are completely free from spatial truncation errors for scattering anisotropy of order L, provided L < N. For X; Y-geometry, the offered multigroup spectral nodal method is based on transverse integrations of the SN equations inside the discretization nodes, separately in x- and y- coordinate directions. Since the transverse-leakage terms are approximated by constants, the resulting nodal method is referred to as the multigroup SGF-contant nodal (SGF-CN) method, which is applied for multigroup X; Y-geometry fixed-source SN problems with isotropic scattering. Numerical results are presented to illustrate the efficiency of the SGF and SGF-CN codes and the accuracy of the converged numerical solutions in coarse-mesh calculations.

Nete trabalho é apresentada uma solução analílica para o problema de ordenada discreta unidimensional e multigrupo de transporle de neutrons em simetria planar. A idéia básica da formulação proposta consiste na aplicação da transformada de Laplace na equação de ordenada discreta. Para a solução do sistema linear resultante, uma solução explícila para a matriz lnversa é estabelecida. Dessa forma, o fluxo angular é obtido, por inversão analítica, em termos do fluxo angular em x=O. Essa formulação é aplicada a problemas de domínio finito e semi-infinito. No primeiro caso, os valores de fluxo angular desconhecidos na fronteira em x=O, são determinados a partir dos valores conhecidos do fluxo angular em x=a; no segundo caso é usada a condição de que o fluxo angular é limilado no infinito. Foram tratados problemas homogêneos e heterogêneos para a placa plana com um grupo de neutrons e multigrupo.O problema inverso, que consiste na determinação do fluxo incidente na fronteira a partir de valores do fluxo escalar no interior do domínio, também foi resolvido. Os resullados obtidos para os problemas acima descritos, apresentaram uma boa comparação com os resultados disponíveis na literatura.

O método LTSN tem sido utilizado na resolução de uma classe abrangente de problemas de transporte de partículas neutras que são reduzidos a um sistema linear algébrico depois da aplicação da transformada de Laplace. Na maioria dos casos estudados os autovalores associados são reais e simétricos. Para o problema de criticalidade os autovalores associados são reais ou imaginários puros e simétricos, e para o o problema de multigrupo podem aparecer autovalores complexos. O objetivo deste trabalho consiste na generalização da formulação LTSN para problemas de transporte com autovalores complexos. Por esse motivo é focada a solução de um problema radiativo de transporte com polarização em uma placa plana. A solução apresentada fundamenta-se na aplicação da transformada de Laplace ao conjunto de equações SN dos problemas resultantes da decomposição da equação de transferência radiativa com polarização em série de Fourier, seguindo o procedimento de Chandrasekhar. Esse procedimento gera 2L + 2 sistemas lineares de ordem 4N dependentes do parâmetro complexo "s". Aqui, L é o grau de anisotropia e N a ordem de quadratura. A solução desse sistema simbólico é obtida através da aplicação da transformada inversa de Laplace depois da inversão da matriz simbólica pelo método da diagonalização. Para a obtenção das constantes de integração é assumido que os componentes do vetor de Stokes são reais e as matrizes dos autovalores e autovetores são separadas em suas partes real e imaginária. A solução LTSN para autovalores complexos é validada através da comparação da solução para uma placa com espessura unitária, grau de anisotropia L = 13, albedo de espalhamento simples $ = 0:99, coe ciente de re exão de Lambert ¸0 = 0:1 e N = 150, segundo dados da literatura consultada.

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Nesta dissertação o método espectronodal SD-SGF-CN, cf. spectral diamond spectral Green's function - constant nodal, é utilizado para a determinação dos fluxos angulares médios nas faces dos nodos homogeneizados em domínio heterogêneo. Utilizando esses resultados, desenvolvemos um algoritmo para a reconstrução intranodal da solução numérica visto que, em cálculos de malha grossa, soluções numéricas mais localizadas não são geradas. Resultados numéricos são apresentados para ilustrar a precisão do algoritmo desenvolvido.
In this dissertation the spectral nodal method SD-SGF-CN, cf. spectral diamond spectral Green's function - constant nodal, is used to determine the angular fluxes averaged along the edges of the homogenized nodes in heterogeneous domains. Using these results, we developed an algorithm for the reconstruction of the node-edge average angular fluxes within the nodes of the spatial grid set up on the domain, since more localized numerical solutions are not generated by coarse-mesh numerical methods. Numerical results are presented to illustrate the accuracy of the algorithm we offer.

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Um método numérico nodal livre de erros de truncamento espacial é desenvolvido para problemas adjuntos de transporte de partículas neutras monoenergéticas em geometria unidimensional com fonte fixa na formulação de ordenadas discretas (SN). As incógnitas no método são os fluxos angulares adjuntos médios nos nodos e os fluxos angulares adjuntos nas fronteiras dos nodos, e os valores numéricos gerados para essas quantidades são os obtidos a partir da solução analítica das equações SN adjuntas. O método é fundamentado no uso da convencional equação adjunta SN discretizada de balanço espacial, que é válida para cada nodo de discretização espacial e para cada direção discreta da quadratura angular, e de uma equação auxiliar adjunta não convencional, que contém uma função de Green para os fluxos angulares adjuntos médios nos nodos em termos dos fluxos angulares adjuntos emergentes das fronteiras dos nodos e da fonte adjunta interior. Resultados numéricos são fornecidos para ilustrarem a precisão do método proposto.
A numerical nodal method that is free from all spatial truncation errors is developed for one-speed slab-geometry discrete ordinates (SN) fixed-source adjoint neutral particle transport problems. The unknown in the method are the node-edge and the node-average adjoint angular fluxes, and the numerical values obtained for these quantities are those of the analytic solution of the adjoint SN equations. The method is based on the use of the standard spatially discretized SN balance adjoint equation, which holds in each spatial node and for each discrete ordinates direction, and a nonstandard adjoint auxiliary equation that contains a Greens function for the node-average adjoint angular fluxes in terms of the exiting adjoint angular fluxes from the node edges and the adjoint interior source. Numerical results are given to illustrate the methods accuracy.

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Neste trabalho, nós desenvolvemos um método de aproximação polinomial para obtermos as funções de transferência que aparecem nas equações auxiliares do método SGF para problemas monoenergéticos com espalhamento linearmente anisotrópico em geometria Cartesiana unidimensional. Para isto, utilizamos os polinômios de Lagrange para comparar os resultados numéricos com aqueles gerados pelo método SGF analítico aplicado a problemas SN em domínios heterogêneos. Este trabalho é um estudo preliminar para um novo propósito, que é a aproximação das exponenciais que aparecem nos termos de fuga transversal do método ExpN-SGF.
In this work we evaluate polynomial approximations to obtain the transfer functions that appear in SGF auxiliary equations (Green´s Functions) for monoenergetic linearly anisotropic scattering SN equations in one-dimensional Cartesian geometry. For this task we use Lagrange Polynomials in order to compare the numerical results with the ones generated by the standard SGF method applied to SN problems in heterogeneous domains. This work is a preliminary investigation of a new proposal for handling the transverse leakage terms that appear in the transverse-integrated one-dimensional SN equations when we use the SGF – exponential nodal method (SGF-ExpN) in multidimensional rectangular geometry.

Neste trabalho estendemos o método LTSN2D-DiagExp para problemas de transporte de nêutrons bidimensionais heterogêneos e construímos um novo algoritmo para resolver as equações de ordenadas discretas SN tridimensionais em domínios homogêneos e heterogêneos, denominado LTSN3D-DiagExp. Esses algoritmos são construídos a partir da diagonalização das matrizes de transporte SN. Os termos de fuga transversal, que surgem nas equações SN integradas transversalmente, são representados por uma função exponencial com constante de decaimento heuristicamente identificada com parâmetros materiais característicos do meio. Como os autovalores podem ter multiplicidade maior que a unidade, desenvolvemos uma análise espectral a fim de garantir a diagonalização e estudar questões de estabilidade. Um estudo sobre o condicionamento é também feito. Definimos os erros no fluxo aproximado e na fórmula da quadratura, e estabelecemos uma relação entre eles. A convergência ocorre com condções de fronteira e quadratura angular adequadas. Apresentamos os resultados numéricos gerados pelos novos métodos LTSN2D-DiagExp e LTSN3D-DiagExp aplicados a problemas disponíveis na literatura.
In this work we extend the LTSN2D-DiagExp method for heterogeneous twodimensional neutral particle transport problems and we construct a new algorithm to numerically solve three-dimensional discrete ordinates equations SN in homogeneous and heterogeneous domains, that we refer to as the LTSN3D-DiagExp method. The essence of these methods are the diagonalization of the SN transport matrices. The transverse leakage terms that appear in the transverse integrated SN equations, are represented by exponential functions with decay constant depending on the characteristics of the material associated to the medium the particles leave behind. As the eigenvalues can have multiplicity greater than one, we present a spectral analysis in order to find the eigenvalues and corresponding linearly independent eigenvectors. Moreover, a study about the condition of the transport matrix is offered. We define the errors in the approach flow and the formula of the quadrature, and establish a relation between them. The convergence occurs depending on the boundary conditions and the adequate choice of the angular quadrature scheme. We present numerical results generated by present methods (LTSN2D-DiagExp and LTSN3D-DiagExp) applied to model problems available in the literature.

Um método de matriz resposta (RM) é descrito para gerar soluções numéricas livres de erros de truncamento espacial para problemas de transporte de nêutrons monoenergéticos e com fonte fixa, em geometria unidimensional na formulação de ordenadas discretas (SN). O método RM com esquema iterativo de inversão parcial por região (RBI) converge valores numéricos para os fluxos angulares nas fronteiras das regiões que coincidem com os valores da solução analítica das equações SN, afora os erros de arredondamento da aritmética finita computacional. Desenvolvemos um esquema numérico de reconstrução espacial, que fornece a saída para os fluxos escalares de nêutrons em qualquer ponto do domínio definido pelo usuário, com um passo de avanço também escolhido pelo usuário. Resultados numéricos são apresentados para ilustrar a precisão do presente método em cálculos de malha grossa.
Presented here is a response matrix (RM) method, which solves numerically fixedsource one-speed slab-geometry neutron transport problems in the discrete ordinates (SN) formulation. The numerical solutions are completely free from spatial truncation errors. Therefore, the RM method with the RBI iterative scheme converges numerical values for the region-edge angular fluxes, which coincide with the numerical values generated from the analytical solution, apart from computational finite arithmetic considerations. A spatial reconstruction scheme has also been developed to yield the detailed profile of the scalar flux using a fixed step defined by the code user. Numerical results are given to illustrate the offered methods accuracy.

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A computação em cluster tem sido amplamente utilizada como uma alternativa de relativo baixo custo para processamento paralelo em aplicações científicas. Com a utilização do padrão de interface de troca de mensagens (MPI, do inglês Message-Passing Interface), o desenvolvimento tornou-se ainda mais acessível e difundido na comunidade científica. Uma tendência mais recente é a utilização de Unidades de Processamento Gráfico (GPU, do inglês Graphic Processing Unit), que são poderosos coprocessadores capazes de realizar centenas de instruções ao mesmo tempo, podendo chegar a uma capacidade de processamento centenas de vezes a de uma CPU. Entretanto, um microcomputador convencional não abriga, em geral, mais de duas GPUs. Portanto, propõe-se neste trabalho o desenvolvimento e avaliação de uma abordagem paralela híbrida de baixo custo na solução de um problema típico da engenharia nuclear. A ideia é utilizar a tecnologia de paralelismo em clusters (MPI) em conjunto com a de programação de GPUs (CUDA, do inglês Compute Unified Device Architecture) no desenvolvimento de um sistema para simulação do transporte de nêutrons, através de uma blindagem por meio do Método Monte Carlo. Utilizando a estrutura física de cluster composto de quatro computadores com processadores quad-core e 2 GPUs cada, foram desenvolvidos programas utilizando as tecnologias MPI e CUDA. Experimentos empregando diversas configurações, desde 1 até 8 GPUs, foram executados e comparados entre si, bem como com o programa sequencial (não paralelo). Observou-se uma redução do tempo de processamento da ordem de 2.000 vezes quando se comparada a versão paralela de 8 GPUs com a versão sequencial. Os resultados aqui apresentados são discutidos e analisados com o objetivo de destacar ganhos e possíveis limitações da abordagem proposta.
Cluster computing has been widely used as a low cost alternative for parallel processing in scientific applications. With the use of Message-Passing Interface (MPI) protocol development became even more accessible and widespread in the scientific community. A more recent trend is the use of Graphic Processing Unit (GPU), which is a powerful co-processor able to perform hundreds of instructions in parallel, reaching a capacity of hundreds of times the processing of a CPU. However, a standard PC does not allow, in general, more than two GPUs. Hence, it is proposed in this work development and evaluation of a hybrid low cost parallel approach to the solution to a nuclear engineering typical problem. The idea is to use clusters parallelism technology (MPI) together with GPU programming techniques (CUDA – Compute Unified Device Architeture) to simulate neutron transport through a slab using Monte Carlo method. By using a cluster comprised by four quad-core computers with 2 GPU each, it has been developed programs using MPI and CUDA technologies. Experiments, applying different configurations, from 1 to 8 GPUs has been performed and results were compared with the sequential (non-parallel) version. A speed up of about 2.000 times has been observed when comparing the 8- GPU with the sequential version. Results here presented are discussed and analysed with the objective of outlining gains and possible limitations of the proposed approah.

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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Coordenação de Aperfeiçoamento de Pessoal de Nível Superior
Neste trabalho, três técnicas para resolver numericamente problemas inversos de transporte de partículas neutras a uma velocidade para aplicações em engenharia nuclear são desenvolvidas. É fato conhecido que problemas diretos estacionários e monoenergéticos de transporte são caracterizados por estimar o fluxo de partículas como uma função-distribuição das variáveis independentes de espaço e de direção de movimento, quando os parâmetros materiais (seções de choque macroscópicas), a geometria, e o fluxo incidente nos contornos do domínio (condições de contorno), bem como a distribuição de fonte interior são conhecidos. Por outro lado, problemas inversos, neste trabalho, buscam estimativas para o fluxo incidente no contorno, ou a fonte interior, ou frações vazio em barras homogêneas. O modelo matemático usado tanto para os problemas diretos como para os problemas inversos é a equação de transporte independente do tempo, a uma velocidade, em geometria unidimensional e com o espalhamento linearmente anisotrópico na formulação de ordenadas discretas (SN). Nos problemas inversos de valor de contorno, dado o fluxo emergente em um extremo da barra, medido por um detector de nêutrons, por exemplo, buscamos uma estimativa precisa para o fluxo incidente no extremo oposto. Por outro lado, nos problemas inversos SN de fonte interior, buscamos uma estimativa precisa para a fonte armazenada no interior do domínio para fins de blindagem, sendo dado o fluxo emergente no contorno da barra. Além disso, nos problemas inversos SN de fração de vazio, dado o fluxo emergente em uma fronteira da barra devido ao fluxo incidente prescrito no extremo oposto, procuramos por uma estimativa precisa da fração de vazio no interior da barra, no contexto de ensaios não-destrutivos para aplicações na indústria. O código computacional desenvolvido neste trabalho apresenta o método espectronodal de malha grossa spectral Greens function (SGF) para os problemas diretos SN em geometria unidimensional para gerar soluções numéricas precisas para os três problemas inversos SN descritos acima. Para os problemas inversos SN de valor de contorno e de fonte interior, usamos a propriedade da proporcionalidade da fuga de partículas; ademais, para os problemas inversos SN de fração de vazio, oferecemos a técnica a qual nos referimos como o método físico da bissecção. Apresentamos resultados numéricos para ilustrar a precisão das três técnicas, conforme descrito nesta tese.
In this work, three techniques for numerically solving one-speed neutral particle inverse transport problems for nuclear engineering applications are developed. It is well known that direct steady-state monoenergetic transport problems are characterized by estimating the flux of particles as a distribution function of space and direction-of-motion independent variables, when the material parameters (cross sections), the geometry, and the incoming flux at the boundaries of the domain (boundary conditions), as well as the interior source distribution are known. Conversely, inverse problems, in this work, seek for estimates to the incident boundary flux, or interior source, or void fractions in homogeneous slabs. The mathematical model used for direct and inverse problems is the time-independent one-speed slab-geometry transport equation with linearly anisotropic scattering in the discrete ordinates (SN) formulation. In the boundary-value inverse problems, given the existing flux at one boundary of the slab, as measured by a neutron detector, for example, we seek for accurate estimate for the incident flux at the opposite boundary. On the other hand, in the interior source inverse SN problems, we seek for accurate estimate for the interior source stored within the slab for shielding purpose, given the exiting flux at the boundary of the slab. Furthermore, as with the void fraction inverse SN problems, given the exiting flux at one boundary of the slab due to prescribed incident flux at the opposite boundary, we seek for accurate estimate of the void fraction within the slab in the context of non-destructive testing applications in industry. The computer code developed in this work presents the coarse-mesh spectral Greens function (SGF) nodal method for direct SN problems in slab geometry to generate accurate numerical solutions to the three inverse SN problems described above. For the boundary-value and interior source inverse SN problems, we use the proportionality property of the leakage of particles; moreover, for the void fraction inverse SN problems, we offer the technique that we refer to as the physical bisection method. We present numerical results to illustrate the accuracy of the three techniques, as described in this dissertation.

Na presente tese é resolvida a equação de difusão de nêutrons estacionária, bem como problemas de cinética, em geometria unidimensional cartesiana multi-região considerando o modelo de multigrupos de energia. Um dos objetivos e inovação neste trabalho é a obtenção de uma solução aproximada com estimativa de erro, controle de precisão e na forma de uma expressão analítica. Com esse tipo de solução não há a necessidade de recorrer a esquemas de interpolação, geralmente necessários em caso de discretizações do domínio. O fluxo de nêutrons é expandido em uma série de Taylor cujos coeficientes são encontrados utilizando a equação diferencial e as condições de contorno e interface. O domínio é dividido em várias células, cujo tamanho e o grau do polinômio são ajustáveis de acordo com a precisão requerida. Para resolver o problema de autovalor é utilizado o método da potência. A metodologia é aplicada em um benchmark que consiste na solução da equação de difusão como condição inicial e na solução de problemas de cinética para diferentes transientes. Os resultados são comparados com sucesso com resultados da literatura. A convergência da série é garantida pela aplicação de um raciocínio baseado no critério de Lipschitz para funções contínuas. Cabe ressaltar que a solução obtida, em conjunto com a análise da convergência, mostra a solidez e a precisão dessa metodologia.
In the present dissertation the one-dimensional neutron diffusion equation for stationary and kinetic problems in a multi-layer slab has been solved considering the multi-group energy model. One of the objectives and innovation in this work is to obtain an approximate solution with error estimation, accuracy control and in the form of an analytical expression. With this solution there is no need for interpolation schemes, which are usually needed in case of discretization of the domain. The neutron flux is expanded in a Taylor series whose coefficients are found using the differential equation and the boundary and interface conditions. The domain is divided into several layers, whose size and the polynomial order can be adjusted according to the required accuracy. To solve the eigenvalue problem the conventional power method has been used. The methodology is applied in a benchmark problem consisting of the solution of the diffusion equation as an initial condition and solving kinetic problems for different transients. The results are compared successfully with the ones in the literature. The convergence of the series is guaranteed by applying a criterion based on the Lipschitz criterion for continuous functions. Note that the solution obtained, together with the convergence analysis, shows the robustness and accuracy of this methodology.

Several comprehensive but time consuming neutronic codes are available for performing nuclear reactor and fuel cycle evaluations. In addition, simple models utilizing collision probability theory are used to perform similar tasks with reasonable accuracy. However, the current collision probability theory treats the heterogeneous reactor configurations with a two region unit cell model. This model does not address several important reactor parameters including spatial self-shielding effects and simultaneous use of different reactor fuels within a reactor core. This dissertation studies the fidelity of expanding the collision probability theory to address the stated shortcomings through analyzing two problems. Problem 1 analyzes the effects of self-shielding. The cylindrical fuel region is divided into several sub-regions and an overall equivalent escape probability from the entire fuel region is developed based on the identified neutron transmission and escape probabilities within each fuel sub-region. The multiplication factor and radioisotopic inventory results based on modified V:BUDS (Visualize: Burnup, Depletion, Spectrum) code are in good agreement with benchmark scenarios for a reactor unit cell. The accurate multiplication factor calculation allows more accurate studies on the maximum fuel burnup and radionuclide inventories of interest in nuclear non-proliferation studies. Problem 2 analyzes the effects of simultaneous use of different fuels within a fuel lattice where the zero neutron leakage assumption across the unit cell boundaries is not valid. The developed methodology expands capabilities of the collision probability theory to a supercell model that allows existence of two different fuels. The radioisotopic inventory results for different fuels obtained from the modified V:BUDS code are in excellent agreement with the benchmark problems. These accurate results may be used in general fuel cycle and transmutation studies within power reactors.

Discretization and solving of the transport equation has been an area of great research where many methods have been developed. Under the deterministic transport methods, the method of characteristics, MOC, is one such discretization and solution method that has been applied to large-scale problems. Although these MOC, specifically long characteristics, LC, have been thoroughly applied to discretize and solve transport problems in the spatial domain, there is a need for an equally adequate time-dependent discretization. A method has been developed that uses LC discretization of the time and space variables in solving the transport equation. This space-time long characteristic, STLC, method is a discrete ordinates method that applies LC discretization in space and time and employs a least-squares approximation of sources such as the scattering source in each cell. This method encounters the same problems that previous spatial LC methods have dealt with concerning achieving all of the following: particle conservation, exact solution along a ray, and smooth variation in reaction rate for specific problems. However, quantities that preserve conservation in each cell can also be produced with this method and compared to the non-conservative results from this method to determine the extent to which this STLC method addresses the previous problems. Results from several test problems show that this STLC method produces conservative and non-conservative solutions that are very similar for most cases and the difference between them vanishes as track spacing is refined. These quantities are also compared to the results produced from a traditional linear discontinuous spatial discretization with finite difference time discretization. It is found that this STLC method is more accurate for streaming-dominate and scattering-dominate test problems. Also, the solution from this STLC method approaches the steady-state diffusion limit solution from a traditional LD method. Through asymptotic analysis and test problems, this STLC method produces a time-dependent diffusion solution in the thick diffusive limit that is accurate to O(E) and is similar to a continuous linear FEM discretization method in space with time differencing. Application of this method in parallel looks promising, mostly due to the ray independence along which the solution is computed in this method.

Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort.
UOIT

Experiments and numerical models were used to investigate an inductively coupled plasma source (ICPS) operating with a magnetic filter field. The work shows that applying magnetic filters transversely to the plasma offers several new control parameters to help enhance the properties of a plasma source. The application of these new results using magnetic enhancement is discussed with respect to both industrial plasma fabrication processes and neutral beam injection for fusion power. Experimental measurements of the power transfer efficiency of the ICPS were undertaken comparing the effect of the magnetic field for both hydrogen and argon plasmas. The location and strength of the magnetic field was varied while measurements of the plasma resistance and power transfer efficiency were performed. The changes in forward power transfer were correlated with plasma density measurements and a numerical model of the electrical plasma circuit was used to guide the optimal choice for the power system components. The results demonstrate that the magnetic field increases the total efficiency of the plasma source and that the gains are strongly dependant on the choice of location for the magnetic field. Plasma properties were then investigated across the plasma source 1 cm intervals. Experimental measurements comparing the effect of the magnetic filter on the plasma properties include: electron densities using a hairpin probe, electron energy probability functions using a compensated Langmuir probe, negative ion densities by laser photo detachment and rotational gas temperatures by optical emission spectroscopy. These measurements revealed interesting new properties of the plasma when a magnetic filter is applied including: the formation of a high density cold particle trap, changes in particle transport and drift motions, increased gas temperatures, and a peak in negative ion density under the magnetic filter center. Pulsing the plasma can greatly affect the plasma dynamics, leading to electron cooling in the afterglow and increased negative ion production. A combination of a pulsed plasma with a magnetic filter was then investigated. Measurements of the negative ion and electron populations were performed in the plasma afterglow with the magnetic filter applied. The results reveal a complex and dynamic afterglow process including strong spatial dependencies measured for diffusive transport, ambipolar breakdown and ion-ion plasma formation. The applications for this work include offering new avenues for control over processing plasma chemistry as well as initial results toward the future viability of a caesium-free pulsed negative ion neutral beam source.