Multigroup diffusion preconditioners for multiplying fixed-source transport problems

Several preconditioners based on multigroup di usion are developed for application to multiplying fi xed-source transport problems using the discrete ordinates method. By starting from standard, one-group, diff usion synthetic acceleration (DSA), a multigroup diff usion preconditioner is constructed that shares the same fi ne mesh as the transport problem. As a cheaper but effective alternative, a two-grid, coarse-mesh, multigroup diff usion preconditioner is examined, for which a variety of homogenization schemes are studied to generate the coarse mesh operator. Finally, a transport-corrected diff usion preconditioner based on application of the Newton-Shulz algorithm is developed. The results of several numerical studies indicate the coarse-mesh, diff usion preconditioners work very well. In particular, a coarse-mesh, transport-corrected, diff usion preconditioner reduced the computational time of multigroup GMRES by up to a factor of 17 and outperformed best-case Gauss-Seidel results by over an order of magnitude for all problems studied

Meshless Methods for the Neutron Transport Equation

Mesh-based methods for the numerical solution of partial differential equations (PDEs) require the division of the problem domain into non-overlapping, contiguous subdomains that conform to the problem geometry. The mesh constrains the placement and connectivity of the solution nodes over which the PDE is solved. In meshless methods, the solution nodes are independent of the problem geometry and do not require a mesh to determine connectivity. This allows the solution of PDEs on geometries that would be difficult to represent using even unstructured meshes. The ability to represent difficult geometries and place solution nodes independent of a mesh motivates the use of meshless methods for the neutron transport equation, which often includes spatially-dependent PDE coefficients and strong localized gradients. The meshless local Petrov-Galerkin (MLPG) method is applied to the steady-state and k-eigenvalue neutron transport equations, which are discretized in energy using the multigroup approximation and in angle using the discrete ordinates approximation. The MLPG method uses weighted residuals of the transport equation to solve for basis function expansion coefficients of the neutron angular flux. Connectivity of the solution nodes is determined by the shared support domain of overlapping meshless functions, such as radial basis functions (RBFs) and moving least squares (MLS) functions. To prevent oscillations in the neutron flux, the MLPG transport equation is stabilized by the streamline upwind Petrov-Galerkin (SUPG) method, which adds numerical diffusion to the streaming term. Global neutron conservation is enforced by using MLS basis and weight functions and appropriate SUPG parameters. The cross sections in the transport equation are approximated in accordance with global particle balance and without constraint on their spatial dependence or the location of the basis and weight functions. The equations for the strong-form meshless collocation approach are derived for comparison to the MLPG equations. Two integration schemes for the basis and weight functions in the MLPG method are presented, including a background mesh integration and a fully meshless integration approach. The method of manufactured solutions (MMS) is used to verify the resulting MLPG method in one, two and three dimensions. Results for realistic problems, including two-dimensional pincells, a reflected ellipsoid and a three-dimensional problem with voids, are verified by comparison to Monte Carlo simulations. Finally, meshless heat transfer equations are derived using a similar MLPG approach and verified using the MMS. These heat equation are coupled to the MLPG neutron transport equations, and results for a pincell are compared to values from a commercial pressurized water reactor.PHDNuclear Engineering & Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttps://deepblue.lib.umich.edu/bitstream/2027.42/145796/1/brbass_1.pd

Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation

[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.[CA] 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/10251/144771TESI

Multilevel matrix-free preconditioner to solve linear systems associated with a the time-dependent SPN equations

[EN] The evolution of the neutronic power inside of a nuclear reactor core can be approximated by means of the diffusive time-dependent simplified spherical harmonics equations (SPN). For the spatial discretization of these equations, a continuous Galerkin high order finite element method is applied to obtain a semi-discrete system of equations that is usually stiff. A semi-implicit time scheme is used for the time discretization and many linear systems are needed to be solved and previously, preconditioned. The aim of this work is to speed up the convergence of the linear systems solver with a multilevel preconditioner that uses different degrees of the polynomials used in the finite element method. Furthermore, as the matrices that appear in this type of system are very large and sparse, a matrix-free implementation of the preconditioner is developed to avoid the full assembly of the matrices. A benchmark transient tests this methodology. Numerical results show, in comparison with the block Gauss-Seidel preconditioner, an improvement in terms of number of iterations and the necessity of computational resources.This work has been partially supported by Spanish Ministerio de Economía y Competitividad under projects ENE2017-89029-P and MTM2017-85669-P. Furthermore, this work has been financed by the Generalitat Valenciana under the project PROMETEO/2018/035.Carreño, A.; Vidal-Ferràndiz, A.; Ginestar, D.; Verdú, G. (2022). Multilevel matrix-free preconditioner to solve linear systems associated with a the time-dependent SPN equations. En Proceedings of the YIC 2021 - VI ECCOMAS Young Investigators Conference. Editorial Universitat Politècnica de València. 68-77. https://doi.org/10.4995/YIC2021.2021.12510OCS687

Analysis of Iterative Methods for the Linear Boltzmann Transport Equation

In this article we consider the iterative solution of the linear system of equations arising from the discretisation of the poly-energetic linear Boltzmann transport equation using a discontinuous Galerkin finite element approximation in space, angle, and energy. In particular, we develop preconditioned Richardson iterations which may be understood as generalisations of source iteration in the mono-energetic setting, and derive computable a posteriori bounds for the solver error incurred due to inexact linear algebra, measured in a relevant problem-specific norm. We prove that the convergence of the resulting schemes and a posteriori solver error estimates are independent of the discretisation parameters. We also discuss how the poly-energetic Richardson iteration may be employed as a preconditioner for the generalised minimal residual (GMRES) method. Furthermore, we show that standard implementations of GMRES based on minimising the Euclidean norm of the residual vector can be utilized to yield computable a posteriori solver error estimates at each iteration, through judicious selections of left- and right-preconditioners for the original linear system. The effectiveness of poly-energetic source iteration and preconditioned GMRES, as well as their respective a posteriori solver error estimates, is demonstrated through numerical examples arising in the modelling of photon transport.Comment: 27 pages, 8 figure

A Multilevel in Space and Energy Solver for Multigroup Diffusion and Coarse Mesh Finite Difference Eigenvalue Problems

In reactor physics, the efficient solution of the multigroup neutron diffusion eigenvalue problem is desired for various applications. The diffusion problem is a lower-order but reasonably accurate approximation to the higher-fidelity multigroup neutron transport eigenvalue problem. In cases where the full-fidelity of the transport solution is needed, the solution of the diffusion problem can be used to accelerate the convergence of transport solvers via methods such as Coarse Mesh Finite Difference (CMFD). The diffusion problem can have O(108) unknowns, and, despite being orders of magnitude smaller than a typical transport problem, obtaining its solution is still not a trivial task. In the Michigan Parallel Characteristics Transport (MPACT) code, the lack of an efficient CMFD solver has resulted in a computational bottleneck at the CMFD step. Solving the CMFD system can comprise 50% or more of the overall runtime in MPACT when the de facto default CMFD solver is used; addressing this bottleneck is the motivation for our work. The primary focus of this thesis is the theory, development, implementation, and testing of a new Multilevel-in-Space-and-Energy Diffusion (MSED) method for efficiently solving multigroup diffusion and CMFD eigenvalue problems. As its name suggests, MSED efficiently converges multigroup diffusion and CMFD problems by leveraging lower-order systems with coarsened energy and/or spatial grids. The efficiency of MSED is verified via various Fourier analyses of its components and via testing in a 1-D diffusion code. In the later chapters of this thesis, the MSED method is tested on a variety of reactor problems in MPACT. Compared to the default CMFD solver, our implementation of MSED in MPACT has resulted in an ~8-12x reduction in the CMFD runtime required by MPACT for single statepoint calculations on 3-D, full-core, 51-group reactor models. The number of transport sweeps is also typically reduced by the use of MSED, which is able to better converge the CMFD system than the default CMFD solver. This leads to a further savings in overall runtime that is not captured by the differences in CMFD runtime.PHDNuclear Engineering & Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttps://deepblue.lib.umich.edu/bitstream/2027.42/146075/1/bcyee_1.pd

Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses

El principal objetivo de esta tesis es el desarrollo de un Método Modal para resolver dos ecuaciones: la Ecuación de la Difusión de Neutrones y la de las Ordenadas Discretas del Transporte de Neutrones. Además, este método está basado en el Método de Volúmenes Finitos para discretizar las variables espaciales. La solución de estas ecuaciones proporciona el flujo de neutrones, que está relacionado con la potencia que se produce en los reactores nucleares, por lo que es un factor fundamental para los Análisis de Seguridad Nuclear. Por una parte, la utilización del Método Modal está justificada para realizar análisis de inestabilidades en reactores. Por otra parte, el uso del Método de Volúmenes Finitos está justificado por la utilización de este método para resolver las ecuaciones termohidráulicas, que están fuertemente acopladas con la generación de energía en el combustible nuclear. En primer lugar, esta tesis incluye la definición de estas ecuaciones y los principales métodos utilizados para resolverlas. Además, se introducen los principales esquemas y características del Método de Volúmenes Finitos. También se describen los principales métodos numéricos para el Método Modal, que incluye tanto la solución de problemas de autovalores como la solución de Ecuaciones Diferenciales Ordinarias dependientes del tiempo. A continuación, se desarrollan varios algoritmos del Método de Volúmenes Finitos para el Estado Estacionario de la Ecuación de la Difusión de Neutrones. Se consigue desarrollar una formulación multigrupo, que permite resolver el problema de autovalores para cualquier número de grupos de energía, incluyendo términos de upscattering y de fisión en varios grupos de energía. Además, se desarrollan los algoritmos para realizar la computación en paralelo. La solución anterior es la condición inicial para resolver la Ecuación de Difusión de Neutrones dependiente del tiempo. En esta tesis se utiliza un Método Modal, que transforma el Sistema de Ecuaciones Diferenciales Ordinarias en uno de mucho menor tamaño, que se resuelve con el Método de la Matriz Exponencial. Además, se ha desarrollado un método rápido para estimar el flujo adjunto a partir del directo, ya que se necesita en el Método Modal. Por otra parte, se ha desarrollado un algoritmo que resuelve el problema de autovalores de la Ecuación del Transporte de Neutrones. Este algoritmo es para la formulación de Ordenadas Discretas y el Método de Volúmenes Finitos. En concreto, se han aplicado dos tipos de cuadraturas para las Ordenadas Discretas y dos esquemas de interpolación para el Método de Volúmenes Finitos. Finalmente, se han aplicado estos métodos a diferentes tipos de reactores nucleares, incluyendo reactores comerciales. Se han evaluado los valores de la constante de multiplicación y de la potencia, ya que son las variables fundamentales en los Análisis de Seguridad Nuclear. Además, se ha realizado un análisis de sensibilidad de diferentes parámetros como la malla y métodos numéricos. En conclusión, se obtienen excelentes resultados, tanto en precisión como en coste computacional.The main objective of this thesis is the development of a Modal Method to solve two equations: the Neutron Diffusion Equation and the Discrete Ordinates Neutron Transport Equation. Moreover, this method uses the Finite Volume Method to discretize the spatial variables. The solution of these equations gives the neutron flux, which is related to the power produced in nuclear reactors; thus, the neutron flux is a paramount variable in Nuclear Safety Analyses. On the one hand, the use of Modal Methods is justified because one uses them to perform instability analyses in nuclear reactors. On the other hand, it is worth using the Finite Volume Method because one uses it to solve thermalhydraulic equations, which are strongly coupled with the energy generation in the nuclear fuel. First, this thesis defines the equations mentioned above and the main methods to solve these equations. Furthermore, the thesis describes the major schemes and features of the Finite Volume Method. In addition, the author also introduces the major methods used in the Modal Method, which include the methods used to solve the eigenvalue problem, as well as those used to solve the time dependent Ordinary Differential Equations. Next, the author develops several algorithms of the Finite Volume Method applied to the Steady State Neutron Diffusion Equation. In addition, the thesis includes an improvement of the multigroup formulation, which solves problems involving upscattering and fission terms in several energy groups. Moreover, the author optimizes the algorithms to do calculations with parallel computing. The previous solution is used as initial condition to solve the time dependent Neutron Diffusion Equation. The author uses a Modal Method to do so, which transforms the Ordinary Differential Equations System into a smaller system that is solved by using the Exponential Matrix Method. Furthermore, the author developed a computationally efficient method to estimate the adjoint flux from the forward one, because the Modal Method uses the adjoint flux. Additionally, the thesis also presents an algorithm to solve the eigenvalue problem of the Neutron Transport Equation. This algorithm uses the Discrete Ordinates formulation and the Finite Volume Method. In particular, the author uses two types of quadratures for the Discrete Ordinates and two interpolation schemes for the Finite Volume Method. Finally, the author tested the developed methods in different types of nuclear reactors, including commercial ones. The author checks the accuracy of the values of the crucial variables in Nuclear Safety Analyses, which are the multiplication factor and the power distribution. Furthermore, the thesis includes a sensitivity analysis of several parameters, such as the mesh and numerical methods. In conclusion, excellent results are reported in both accuracy and computational cost.El principal objectiu d'esta tesi és el desenvolupament d'un Mètode Modal per a resoldre dos equacions: l'Equació de Difusió de Neutrons i la de les Ordenades Discretes del Transport de Neutrons. A més a més, este mètode està basat en el Mètode de Volums Finits per a discretitzar les variables espacials. La solució d'estes equacions proporcionen el flux de neutrons, que està relacionat amb la potència que es produïx en els reactors nuclears; per tant, el flux de neutrons és un factor fonamental en els Anàlisis de Seguretat Nuclear. Per una banda, la utilització del Mètode Modal està justificada per a realitzar anàlisis d'inestabilitats en reactors. Per altra banda, l'ús del Mètode de Volums Finits està justificat per l'ús d'este mètode per a resoldre les equacions termohidràuliques, que estan fortament acoblades amb la generació d'energia en el combustible nuclear. En primer lloc, esta tesi inclou la definició d'estes equacions i els principals mètodes utilitzats per a resoldre-les. A més d'això, s'introduïxen els principals esquemes i característiques del Mètode de Volums Finits. Endemés, es descriuen els principals mètodes numèrics per al Mètode Modal, que inclou tant la solució del problema d'autovalors com la solució d'Equacions Diferencials Ordinàries dependents del temps. A continuació, es desenvolupa diversos algoritmes del Mètode de Volums Finits per a l'Estat Estacionari de l'Equació de Difusió de Neutrons. Es conseguix desenvolupar una formulació multigrup, que permetre resoldre el problema d'autovalors per a qualsevol nombre de grups d'energia, incloent termes d' upscattering i de fissió en diversos grups d'energia. A més a més, es desenvolupen els algoritmes per a realitzar la computació en paral·lel. La solució anterior és la condició inicial per a resoldre l'Equació de Difusió de Neutrons dependent del temps. En esta tesi s'utilitza un Mètode Modal, que transforma el Sistema d'Equacions Diferencials Ordinàries en un problema de menor tamany, que es resol amb el Mètode de la Matriu Exponencial. Endemés, s'ha desenvolupat un mètode ràpid per a estimar el flux adjunt a partir del directe, perquè es necessita en el Mètode Modal. Per altra banda, s'ha desenvolupat un algoritme que resol el problema d'autovalors de l'Equació de Transport de Neutrons. Este algoritme és per a la formulació d'Ordenades Discretes i el Mètode de Volums Finits. En concret, s'han aplicat dos tipos de quadratures per a les Ordenades Discretes i dos esquemes d'interpolació per al Mètode de Volums Finits. Finalment, s'han aplicat estos mètodes a diversos tipos de reactors nuclears, incloent reactors comercials. S'han avaluat els valor de la constat de multiplicació i de la potència, perquè són variables fonamentals en els Anàlisis de Seguretat Nuclear. Endemés, s'ha realitzat un anàlisi de sensibilitat de diversos paràmetres com la malla i mètodes numèrics. En conclusió, es conseguix obtenir excel·lents resultats, tant en precisió com en cost computacional.Bernal García, Á. (2018). Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/112422TESI

Acceleration Techniques for Discrete-Ordinates Transport Methods with Highly Forward-Peaked Scattering

In this dissertation, advanced numerical methods for highly forward peaked scattering deterministic calculations are devised, implemented, and assessed. Since electrons interact with the surrounding environment through Coulomb interactions, the scattering kernel is highly forward-peaked. This bears the consequence that, with standard preconditioning, the standard Legendre expansion of the scattering kernel requires too many terms for the discretized equation to be solved efficiently using a deterministic method. The Diffusion Synthetic Acceleration (DSA), usually used to speed up the calculation when the scattering is weakly anisotropic, is inefficient for electron transport. This led Morel and Manteuffel to develop a one-dimensional angular multigrid (ANMG) which has proved to be very effective when the scattering is highly anisotropic. Later, Pautz et al. generalized this scheme to multidimensional geometries, but this method had to be stabilized by a diffusive filter that degrades the overall convergence of the iterative scheme. In this dissertation, we recast the multidimensional angular multigrid method without the filter as a preconditioner for a Krylov solver. This new method is stable independently of the anisotropy of the scattering and is increasingly more effective and efficient as the anisotropy increases compared to DSA preconditioning wrapped inside a Krylov solver. At the coarsest level of ANMG, a DSA step is needed. In this research, we use the Modified Interior Penalty (MIP) DSA. This DSA was shown to be always stable on triangular cells with isotropic scattering. Because this DSA discretization leads to symmetric definite-positive matrices, it is usually solved using a conjugate gradient preconditioned (CG) by SSOR but here, we show that algebraic multigrid methods are vastly superior than more common CG preconditioners such as SSOR. Another important part of this dissertation is dedicated to transport equation and diffusion solves on arbitrary polygonal meshes. The advantages of polygonal cells are that the number of unknowns needed to mesh a domain can be decreased and that adaptive mesh refinement implementation is simplified: rather than handling hanging nodes, the adapted computational mesh includes different types of polygons. Numerical examples are presented for arbitrary quadrilateral and polygonal grids

Anisotropic Adaptivity and Subgrid Scale Modelling for the Solution of the Neutron Transport Equation with an Emphasis on Shielding Applications

This thesis demonstrates advanced new discretisation and adaptive meshing technologies that improve the accuracy and stability of using finite element discretisations applied to the Boltzmann transport equation (BTE). This equation describes the advective transport of neutral particles such as neutrons and photons within a domain. The BTE is difficult to solve, due to its large phase space (three dimensions of space, two of angle and one each of energy and time) and the presence of non-physical oscillations in many situations. This work explores the use of a finite element method that combines the advantages of the two schemes: the discontinuous and continuous Galerkin methods. The new discretisation uses multiscale (subgrid) finite elements that work locally within each element in the finite element mesh in addition to a global, continuous, formulation. The use of higher order functions that describe the variation of the angular flux over each element is also explored using these subgrid finite element schemes. In addition to the spatial discretisation, methods have also been developed to optimise the finite element mesh in order to reduce resulting errors in the solution over the domain, or locally in situations where there is a goal of specific interest (such as a dose in a detector region). The chapters of this thesis have been structured to be submitted individually for journal publication, and are arranged as follows. Chapter 1 introduces the reader to motivation behind the research contained within this thesis. Chapter 2 introduces the forms of the BTE that are used within this thesis. Chapter 3 provides the methods that are used, together with examples, of the validation and verification of the software that was developed as a result of this work, the transport code RADIANT. Chapter 4 introduces the inner element subgrid scale finite element discretisation of the BTE that forms the basis of the discretisations within RADIANT and explores its convergence and computational times on a set of benchmark problems. Chapter 5 develops the error metrics that are used to optimise the mesh in order to reduce the discretisation error within a finite element mesh using anisotropic adaptivity that can use elongated elements that accurately resolves computational demanding regions, such as in the presence of shocks. The work of this chapter is then extended in Chapter 6 that forms error metrics for goal based adaptivity to minimise the error in a detector response. Finally, conclusions from this thesis and suggestions for future work that may be explored are discussed in Chapter 7.Open Acces

Multilevel Schwarz methods for multigroup radiation transport problems

The development of advanced discretization methods for the radiation transport equation is of fundamental importance, since the numerical effort of modeling increasingly complex multidimensional problems with increasing accuracy is extremely challenging. Different expressions of this equation arise in several science fields, from nuclear fission and fusion to astrophysics, climatology and combustion. Mathematically, the radiation intensity is usually a rapidly changing function, causing a considerable loss in accuracy for many discretization methods. Depending on the coefficient ranges, the equation behaves like totally different equation types, making it very difficult to find a discretization method that is efficient in all regimes. Computationally, the huge amount of unknowns involved demands not only extremely powerful computers, but also efficient numerical methods and optimized implementations. Today, solvers covering all the coefficient ranges and still being robust in the diffusion dominated case are very scarce. In the last 20 years, Discontinous Galerkin (DG) methods have been studied for the monoenergetic problem, unsuccessfully, due to lack of stability for diffusion-dominated cases. Recently, new mathematical developments have fully explained the instability and provided a remedy by using a numerical flux depending on the scattering cross section and the mesh size. The new formulation has proven to be stable and allows the application of multigrid, matrix-free methods, reducing the memory needed for such an amount of unknowns. We use these numerical methods to address the solution of a energy dependent problem with a multigroup approach. We study the diffusion approximation to the transport problem, obtaining convergence proofs for the symmetric scattering case and advances in the nonsymmetric case, using field of values analysis. For the full transport case, we discretize by means of an asymptotic preserving, weakly penalized discontinuous Galerkin method that we solve with a multigrid preconditioned GMRES solver, using nonoverlapping Schwarz smoothers for the energy and direction dependent radiative transfer problem. To address the local thermodynamic equilibrium (LTE) constraint, we use a nonlinear additive Schwarz method to precondition the Newton solver. By solving full local radiative transfer problems for each grid cell, performed in parallel on a matrix-free implementation, we achieve a method capable to address large scale calculations arising from applications such as astrophysics, atmospheric radiation calculations and nuclear applications. To the best of our knowledge, this is the first time this preconditioner combination has been used in LTE radiation transport and in several tests we show the robustness of the approach for different mesh sizes, cross sections, energy distributions and anisotropic regimes, both in the linear and nonlinear cases