Current advancements in nuclear reactor core design are pushing reactor cores towards greater heterogeneity in an attempt to make nuclear power more sustainable in terms of fuel utilization and long-term disposal needs. These new designs are now being limited by the accuracy of the core simulators/methods. Increasing attention has been given to full core transport as the flux module in future core simulators. However, the current transport methods, due to their significant memory and computational time requirements, are not practical for whole core calculations. While most researchers are working on developing new acceleration and phase space parallelization techniques for the current fine mesh transport methods, this dissertation focuses on the development of a practical heterogeneous coarse mesh transport method. In this thesis, a heterogeneous coarse mesh transport method is extended from two to three dimensions in Cartesian geometry and new techniques are developed to reduce the strain on computational resources. The high efficiency of the method is achieved by decoupling the problem into a series of fixed source calculations in smaller sub-volume elements (e.g. coarse meshes). This decoupling lead to shifting the computation time to a priori calculations of response functions in unique sub-volumes in the system. Therefore, the method is well suited for large problems with repeated geometry such as those found in nuclear reactor cores. Even though the response functions can be generated with any available existing fine-mesh (deterministic or stochastic) code, a stochastic method was selected in this dissertation. Previous work in two dimensions used discrete polynomial expansions that are better suited for treating discrete variables found in pure deterministic transport methods. The amount of data needed to represent very heterogeneous problems accurately became quite large making the three dimensional extension impractical. The deterministic method was thus replaced by a stochastic response function generator making the transition to continuous variables fairly simple. This choice also improves the geometry handling capability of the coarse mesh method.

This work developed a stylized three dimensional benchmark problem based on Argonne National Laboratory's conceptual Advanced Burner Test Reactor design. This reactor is a sodium cooled fast reactor designed to burn recycled fuel to generate power while transmuting long term waste. The specification includes heterogeneity at both the assembly and core levels while the geometry and material compositions are both fully described. After developing the benchmark, 15 group cross sections were developed so that it could be used for transport code method verification. Using the aforementioned benchmark and 15 group cross sections, the Coarse-Mesh Transport Method (COMET) code was compared to Monte Carlo code MCNP5 (MCNP). Results were generated for three separate core cases: control rods out, near critical, and control rods in. The cross section groups developed do not compare favorably to the continuous energy model; however, the primary goal of these cross sections is to provide a common set of approachable cross sections that are widely usable for numerical methods development benchmarking. Eigenvalue comparison results for MCNP vs. COMET are strong, with two of the models within one standard deviation and the third model within one and a third standard deviation. The fission density results are highly accurate with a pin fission density average of less than 0.5% for each model.

This thesis presents a whole-core benchmark of the European Pressurized Reactor (EPR) using multiple transport methods. The core specifications were taken directly from the Final Safety Analysis Report (FSAR) submitted to the Nuclear Regulatory Commission (NRC) and the reactor was modeled in a stylized manner while maintaining full heterogeneity at the pin and assembly level. The geometry and material specifications are given as well as problem-specific cross sections for 2, 4, and 8 energy group calculations. Cross sections were generated using HELIOS, a lattice depletion code based on the Collision Probability Method (CPM). The multi-group cross sections were utilized in the reference calculation, COMET calculation, and response function generation. The reference solution was obtained via an MCNP model identical to the one implemented in COMET. Specific steps towards constructing and running a COMET calculation are outlined. Detailed results including assembly eigenvalues, core eigenvalues, and pin fission densities are presented.

Les schémas de calcul déterministes permettent une modélisation à moindre coût du comportement de la population de neutrons en réacteur, mais sont traditionnellement construits sur des approximations (décomposition réseau/cœur, homogénéisation spatiale et énergétique…). La thèse revient sur une partie de ces sources d’erreur, de façon à rapprocher la méthode déterministe d’un schéma de référence. L’objectif est de profiter des architectures informatiques modernes (HPC) pour résoudre le problème neutronique à l’échelle du cœur 3D, tout en préservant l’opérateur de transport et une partie des hétérogénéités de la géométrie. Ce travail est réalisé au sein du solveur cœur Sn Minaret de la plateforme de calcul Apollo3® pour des réacteurs à neutrons rapides.Une méthode de décomposition de domaine en espace, est retenue. L'idée consiste à décomposer un problème de grande dimension en sous-problèmes "indépendants" de taille réduite. La convergence vers la solution globale est assurée par échange de flux angulaires entre sous-domaines au cours d'un processus itératif. En favorisant un recours massif au parallélisme, les méthodes de décomposition de domaine contribuent à lever les contraintes en mémoire et temps de calcul. La mise en place d'un parallélisme hybride, couplant les technologies MPI et OpenMP, est en particulier propice au passage sur supercalculateur. Une méthode d'accélération de type Coarse Mesh Rebalance  est ajoutée pour pallier à la pénalité de convergence constatée sur la méthode de décomposition de domaine. Le potentiel du nouveau schéma est finalement mis en évidence sur un coeur CFV 3D, construit en préservant l'hétérogénéité des assemblages absorbants. / Deterministic calculation schemes are devised to numerically solve the neutron transport equation in nuclear reactors. Dealing with core-sized problems is very challenging for computers, so much that the dedicated core codes have no choice but to allow simplifying assumptions (assembly- then core-scale steps…). The PhD work aims to correct some of these ‘standard’ approximations, in order to get closer of reference calculations: thanks to important increases in calculation capacities (HPC), nowadays one can solve 3D core-sized problems, using both high mesh refinement and the transport operator. Developments were performed inside the Sn core solver Minaret, from the new CEA neutronics platform Apollo3® for fast neutrons reactors of the CFV-kind.This work focuses on a Domain Decomposition Method in space. The fundamental idea involves splitting a core-sized problem into smaller and 'independent' subproblems. Angular flux is exchanged between adjacent subdomains. In doing so, all combined subproblems converge to the global solution at the outcome of an iterative process. Domain decomposition is well-suited to massive parallelism, allowing much more ambitious computations in terms of both memory requirements and calculation time. An hybrid MPI/OpenMP parallelism is chosen to match the supercomputers architecture. A Coarse Mesh Rebalance accelration technique is added to balance the convergence penalty observed using Domain Decomposition. The potential of the new calculation scheme is demonstrated on a 3D core of the CFV-kind, using an heterogeneous description of the absorbent rods.

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.

Ce travail de thèse est consacré à la mise en œuvre d’une méthode de décomposition de domaine appliquée à l’équation du transport. L’objectif de ce travail est l’accès à des solutions déterministes haute-fidélité permettant de correctement traiter les hétérogénéités des réacteurs nucléaires, pour des problèmes dont la taille varie d’un motif d’assemblage en 3 dimensions jusqu’à celle d’un grand cœur complet en 3D. L’algorithme novateur développé au cours de la thèse vise à optimiser l’utilisation du parallélisme et celle de la mémoire. La démarche adoptée a aussi pour but la diminution de l’influence de l’implémentation parallèle sur les performances. Ces objectifs répondent aux besoins du projet APOLLO3, développé au CEA et soutenu par EDF et AREVA, qui se doit d’être un code portable (pas d’optimisation sur une architecture particulière) permettant de réaliser des modélisations haute-fidélité (best estimate) avec des ressources allant des machines de bureau aux calculateurs disponibles dans les laboratoires d’études. L’algorithme que nous proposons est un algorithme de Jacobi Parallèle par Bloc Multigroupe. Chaque sous domaine est un problème multigroupe à sources fixes ayant des sources volumiques (fission) et surfaciques (données par les flux d’interface entre les sous domaines). Le problème multigroupe est résolu dans chaque sous domaine et une seule communication des flux d’interface est requise par itération de puissance. Le rayon spectral de l’algorithme de résolution est rendu comparable à celui de l’algorithme de résolution classique grâce à une méthode d’accélération non linéaire par la diffusion bien connue nommée Coarse Mesh Finite Difference. De cette manière une scalabilité idéale est atteignable lors de la parallélisation. L’organisation de la mémoire, tirant parti du parallélisme à mémoire partagée, permet d’optimiser les ressources en évitant les copies de données redondantes entre les sous domaines. Les architectures de calcul à mémoire distribuée sont rendues accessibles par un parallélisme hybride qui combine le parallélisme à mémoire partagée et à mémoire distribuée. Pour des problèmes de grande taille, ces architectures permettent d’accéder à un plus grand nombre de processeurs et à la quantité de mémoire nécessaire aux modélisations haute-fidélité. Ainsi, nous avons réalisé plusieurs exercices de modélisation afin de démontrer le potentiel de la réalisation : calcul de cœur et de motifs d’assemblages en 2D et 3D prenant en compte les contraintes de discrétisation spatiales et énergétiques attendues. / This thesis is devoted to the implementation of a domain decomposition method applied to the neutron transport equation. The objective of this work is to access high-fidelity deterministic solutions to properly handle heterogeneities located in nuclear reactor cores, for problems’ size ranging from colorsets of assemblies to large reactor cores configurations in 2D and 3D. The innovative algorithm developed during the thesis intends to optimize the use of parallelism and memory. The approach also aims to minimize the influence of the parallel implementation on the performances. These goals match the needs of APOLLO3 project, developed at CEA and supported by EDF and AREVA, which must be a portable code (no optimization on a specific architecture) in order to achieve best estimate modeling with resources ranging from personal computer to compute cluster available for engineers analyses. The proposed algorithm is a Parallel Multigroup-Block Jacobi one. Each subdomain is considered as a multi-group fixed-source problem with volume-sources (fission) and surface-sources (interface flux between the subdomains). The multi-group problem is solved in each subdomain and a single communication of the interface flux is required at each power iteration. The spectral radius of the resolution algorithm is made similar to the one of a classical resolution algorithm with a nonlinear diffusion acceleration method: the well-known Coarse Mesh Finite Difference. In this way an ideal scalability is achievable when the calculation is parallelized. The memory organization, taking advantage of shared memory parallelism, optimizes the resources by avoiding redundant copies of the data shared between the subdomains. Distributed memory architectures are made available by a hybrid parallel method that combines both paradigms of shared memory parallelism and distributed memory parallelism. For large problems, these architectures provide a greater number of processors and the amount of memory required for high-fidelity modeling. Thus, we have completed several modeling exercises to demonstrate the potential of the method: 2D full core calculation of a large pressurized water reactor and 3D colorsets of assemblies taking into account the constraints of space and energy discretization expected for high-fidelity modeling.