Generating Unstructured Nuclear Reactor Core Meshes in Parallel

AbstractRecent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed

Validation of an Isotope Evolution Model for Apollo3 Calculations in SFR Core

RÉSUMÉ Dans le cadre du développement du code déterministe multi-filières APOLLO3, il est nécessaire de mettre au point et valider les nouveaux schémas de calcul permettant d’améliorer la prédiction des grandeurs neutroniques d’intérêt. Les études neutroniques du coeur CFV du futur prototype ASTRID (RNR refroidi au sodium) demandent notamment un calcul précis des concentrations isotopiques tout au long de son évolution sous irradiation. Le schéma ECCO/ERANOS utilisé actuellement pour les calculs des coeurs de RNR est constitué des deux étapes classiques réseau et coeur. Les sections efficaces microscopiques autoprotégées et condensées à 33 groupes sont déterminées une seule fois en début de vie à partir de calculs 2D ECCO de cellules ou assemblages en réseau infini (avec une concentration infinitésimale pour les noyaux lourds et les produits de fission non présents en début de vie). Elles sont ensuite utilisées dans une modélisation 3D ERANOS pour réaliser les calculs du coeur complet en évolution microscopique. Ce schéma fait l’hypothèse que les sections efficaces microscopiques n’évoluent pas avec l’irradiation et la modification de composition des différents milieux fissiles. La mise en place de nouveaux schémas de calcul avec le code multi-filières APOLLO3 est l’occasion de revenir sur cette hypothèse puisque d’autres possibilités utilisées jusqu’à présent dans les schémas APOLLO2/CRONOS2 des REL sont dorénavant disponibles. Elles reposent sur un paramétrage des sections efficaces autoprotégées, macroscopiques comme microscopiques, en fonction du taux de combustion, obtenues en faisant évoluer les motifs 2D élémentaires cellules ou assemblages. Ces sections sont stockées dans des bibliothèques dites "évoluantes" permettant une interpolation à l’étape coeur. L’évolution coeur peut alors être menée de 2 façons : • macroscopique : le code coeur extrapole les taux de combustion locaux pour chaque pas d’évolution et récupère, via une interpolation linéaire, les sections efficaces macroscopiques de la bibliothèque évoluante pour calculer le pas suivant. • microscopique : le code coeur résout les équations d’évolution isotopiques en temps (équations de Bateman) suivant des stratégies plus ou moins précises (simple extrapolation ou méthodes de type prédicteur-correcteur) en interpolant les sections efficaces microscopiques en fonction du taux de combustion local.----------ABSTRACT In the frame of the development of the new multi-purpose deterministic code APOLLO3, it is necessary to develop and validate the new calculation schemes capable of improving the prediction of the neutronic quantities of interest. The neutronic studies on the CFV core of the new prototype ASTRID (sodium cooled FNR) notably demand an accurate calculation of the isotopic concentrations during its evolution under irradiation. The ECCO/ERANOS scheme, currently used for FNR core calculations, is constituted by the two classic steps: lattice and core. Microscopic cross sections, self-shielded and condensed into 33 energy groups, are determined only one time at the beginning of life from 2D ECCO cell or assembly calculations in infinite lattice (with an infinitesimal concentration of heavy nuclei and fission products not initially present). They are, then, used for 3D ERANOS modelization of the whole core in micro-depletion. This scheme makes the hypothesis that microscopic cross sections do not evolve during the irradiation and the composition change of the different fissile materials. The development of the new APOLLO3 calculation schemes is an opportunity to re-discuss this hypothesis, because other possibilities, currently used in PWR APOLLO2/CRONOS2 calculation schemes, are available. They lie on the parametrization of self-shielded cross sections, both macroscopic and microscopic, as a function of the burn-up. These cross sections are obtained performing 2D cell or assembly depletion calculations. They are, then, stored in “evolving” libraries which allow their interpolation at core step. Core depletion can be performed in two ways: • macroscopic: the core code extrapolates the local burn-up for each evolution step and interpolate the macroscopic cross sections from the “evolving” library in order to evaluate the following time step. • microscopic: the core code solves the Bateman equations with a more or less accurate strategy (simple extrapolation or predictor-corrector methods), interpolating the microscopic cross sections as a function of the local burnup. This model is, a priori, more accurate than the ECCO/ERANOS one, because it takes into account self-shielding and spectrum changes during the evolution

Self-adaptive isogeometric spatial discretisations of the first and second-order forms of the neutron transport equation with dual-weighted residual error measures and diffusion acceleration

As implemented in a new modern-Fortran code, NURBS-based isogeometric analysis (IGA) spatial discretisations and self-adaptive mesh refinement (AMR) algorithms are developed in the application to the first-order and second-order forms of the neutron transport equation (NTE). These AMR algorithms are shown to be computationally efficient and numerically accurate when compared to standard approaches. IGA methods are very competitive and offer certain unique advantages over standard finite element methods (FEM), not least of all because the numerical analysis is performed over an exact representation of the underlying geometry, which is generally available in some computer-aided design (CAD) software description. Furthermore, mesh refinement can be performed within the analysis program at run-time, without the need to revisit any ancillary mesh generator. Two error measures are described for the IGA-based AMR algorithms, both of which can be employed in conjunction with energy-dependent meshes. The first heuristically minimises any local contributions to the global discretisation error, as per some appropriate user-prescribed norm. The second employs duality arguments to minimise important local contributions to the error as measured in some quantity of interest; this is commonly known as a dual-weighted residual (DWR) error measure and it demands the solution to both the forward (primal) and the adjoint (dual) NTE. Finally, convergent and stable diffusion acceleration and generalised minimal residual (GMRes) algorithms, compatible with the aforementioned AMR algorithms, are introduced to accelerate the convergence of the within-group self-scattering sources for scattering-dominated problems for the first and second-order forms of the NTE. A variety of verification benchmark problems are analysed to demonstrate the computational performance and efficiency of these acceleration techniques.Open Acces

Surface Mesh Generation based on Imprinting of S-T Edge Patches

AbstractOne of the most robust and widely used algorithms for all-hexahedral meshes is the sweeping algorithm. However, for multi- sweeping, the most difficult problems are the surface matching and interval assignment for edges on the source and target surfaces. In this paper, a new method to generate surface meshes by imprinting edge patches between the source and target surfaces is proposed. The edge patch imprinting is based on a cage-based morphing of edge patches on the different sweeping layers where deformed and undeformed cages are extracted by propagating edge patches on the linking surfaces. The imprinting results in that the source or target surfaces will be partitioned with the imprinted edge patches. After partitioning, every new source surface should be matched to a new specific target surface where surface mesh projection from one-to-one sweeping based on harmonic mapping[19] can be applied. In addition, 3D edge patches are projected onto 2D computational domains where every sweeping level is planar in order to increase the robustness of imprinting. Finally, the algorithm time complexity is discussed and examples are provided to verify the robustness of our proposed algorithm

Теплогидравлический расчёт ТВС ИРТ-3М в реакторе ИРТ-Т с различными поверхностями теплообмена

The master’s thesis has attempted to model thermal-hydraulic behavior of the fuel assembly of IRT-3M type of IRT-T reactor with various heat-conductivity surfaces.В магистерской диссертации приведено теплогидравлическое моделирование ТВС типа ИРТ-3М реактора ИРТ-Т с различными поверхностями теплообмена

Теплогидравлический расчёт ТВС ИРТ-3М в реакторе ИРТ-Т с различными поверхностями теплообмена

The master’s thesis has attempted to model thermal-hydraulic behavior of the fuel assembly of IRT-3M type of IRT-T reactor with various heat-conductivity surfaces.В магистерской диссертации приведено теплогидравлическое моделирование ТВС типа ИРТ-3М реактора ИРТ-Т с различными поверхностями теплообмена