Bibliografias temáticas / Few-Group Homogenized Cross Sections

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Literatura científica selecionada sobre o tema "Few-Group Homogenized Cross Sections"

Autor: Grafiati
Publicado: 14 de dezembro de 2024
Última modificação: 31 de julho de 2025

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Artigos de revistas sobre o assunto "Few-Group Homogenized Cross Sections"

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Szames, E., K. Ammar, D. Tomatis, and J. M. Martinez. "FEW-GROUP CROSS SECTIONS MODELING BY ARTIFICIAL NEURAL NETWORKS." EPJ Web of Conferences 247 (2021): 06029. http://dx.doi.org/10.1051/epjconf/202124706029.

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This work deals with the modeling of homogenized few-group cross sections by Artificial Neural Networks (ANN). A comprehensive sensitivity study on data normalization, network architectures and training hyper-parameters specifically for Deep and Shallow Feed Forward ANN is presented. The optimal models in terms of reduction in the library size and training time are compared to multi-linear interpolation on a Cartesian grid. The use case is provided by the OECD-NEA Burn-up Credit Criticality Benchmark [1]. The Pytorch [2] machine learning framework is used.
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Tomatis, Daniele. "A multivariate representation of compressed pin-by-pin cross sections." EPJ Nuclear Sciences & Technologies 7 (2021): 8. http://dx.doi.org/10.1051/epjn/2021006.

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Since the 80’s, industrial core calculations employ the two-step scheme based on prior cross sections preparation with few energy groups and in homogenized reference geometries. Spatial homogenization in the fuel assembly quarters is the most frequent calculation option nowadays, relying on efficient nodal solvers using a coarse mesh. Pin-wise reaction rates are then reconstructed by dehomogenization techniques. The future trend of core calculations is moving however toward pin-by-pin explicit representations, where few-group cross sections are homogenized in the single pins at many physical c
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Nguyen, Dinh Quoc Dang, and Emiliano Masiello. "Representation of few-group homogenized cross section by multi-variate polynomial regression." EPJ Web of Conferences 302 (2024): 02002. http://dx.doi.org/10.1051/epjconf/202430202002.

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In this paper, a representation of few-group homogenized cross section by multi-variate polynomial regression is presented. The method is applied on the few-group assembly homogenized cross sections of the assembly 22UA from the benchmark X2VVER[1], generated by the lattice transport code APOLLO3®[2], and conducted over a Cartesian grid of parametric state-points. The regression model [3, 4] allow to input a significantly larger number of points for training compared to the number of monomials, thus yielding higher accuracy than polynomial interpolation without being affected by the choice of
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Szames, E., K. Ammar, D. Tomatis, and J. M. Martinez. "FEW-GROUP CROSS SECTIONS LIBRARY BY ACTIVE LEARNING WITH SPLINE KERNELS." EPJ Web of Conferences 247 (2021): 06012. http://dx.doi.org/10.1051/epjconf/202124706012.

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This work deals with the representation of homogenized few-groups cross sections libraries by machine learning. A Reproducing Kernel Hilbert Space (RKHS) is used for different Pool Active Learning strategies to obtain an optimal support. Specifically a spline kernel is used and results are compared to multi-linear interpolation as used in industry, discussing the reduction of the library size and of the overall performance. A standard PWR fuel assembly provides the use case (OECD-NEA Burn-up Credit Criticality Benchmark [1]).
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Nguyen, Dinh Q. D., Emiliano Masiello, and Daniele Tomatis. "MPOGen: A Python package to prepare few-group homogenized cross sections for core calculations by APOLLO3®." Nuclear Engineering and Design 417 (February 2024): 112802. http://dx.doi.org/10.1016/j.nucengdes.2023.112802.

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Henry, Romain, Yann Périn, Kiril Velkov, and Sergei Pavlovich Nikonov. "3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET." EPJ Web of Conferences 247 (2021): 06015. http://dx.doi.org/10.1051/epjconf/202124706015.

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A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems. Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes,
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Galchenko, V. V., А. М. Abdulaev, and І. І. Shlapak. "USING SOFTWARE BASED ON THE MONTE CARLO METHOD FOR RECEIVING THE FEW-GROUP HOMOGENIZED MACROSCOPIC INTERACTION CROSS-SECTIONS." Odes’kyi Politechnichnyi Universytet Pratsi, no. 3(53) (2017): 37–42. http://dx.doi.org/10.15276/opu.3.53.2017.05.

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Cao, Liangzhi, Yong Liu, Wei Shen, and Qingming He. "Development of a hybrid method to improve the sensitivity and uncertainty analysis for homogenized few-group cross sections." Journal of Nuclear Science and Technology 54, no. 7 (2017): 769–83. http://dx.doi.org/10.1080/00223131.2017.1315973.

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Truffinet, Olivier, Karim Ammar, Jean-Philippe Argaud, Nicolas Gérard Castaing, and Bertrand Bouriquet. "Multi-output gaussian processes for the reconstruction of homogenized cross-sections." EPJ Web of Conferences 302 (2024): 02006. http://dx.doi.org/10.1051/epjconf/202430202006.

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Deterministic nuclear reactor simulators employing the prevalent two-step scheme often generate a substantial amount of intermediate data at the interface of their two subcodes, which can impede the overall performance of the software. The bulk of this data comprises “few-groups homogenized cross-sections” or HXS, which are stored as tabulated multivariate functions and interpolated inside the core simulator. A number of mathematical tools have been studied for this interpolation purpose over the years, but few meet all the challenging requirements of neutronics computation chains: extreme acc
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Truffinet, Olivier, Karim Ammar, Jean-Philippe Argaud, Nicolas Gérard Castaing, and Bertrand Bouriquet. "Adaptive sampling of homogenized cross-sections with multi-output gaussian processes." EPJ Web of Conferences 302 (2024): 02010. http://dx.doi.org/10.1051/epjconf/202430202010.

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In another talk submitted to this conference, we presented an efficient new framework based on multi-outputs gaussian processes (MOGP) for the interpolation of few-groups homogenized cross-sections (HXS) inside deterministic core simulators. We indicated that this methodology authorized a principled selection of interpolation points through adaptive sampling. We here develop this idea by trying simple sampling schemes on our problem. In particular, we compare sample scoring functions with and without integration of leave-one-out errors, and obtained with single-output and multi-output gaussian
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Mais fontes

Teses / dissertações sobre o assunto "Few-Group Homogenized Cross Sections"

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Nguyen, Dinh Quoc Dang. "Representation of few-group homogenized cross sections by polynomials and tensor decomposition." Electronic Thesis or Diss., université Paris-Saclay, 2024. http://www.theses.fr/2024UPASP142.

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Cette thèse se concentre sur l'étude de la modélisation mathématique des sections efficaces homogénéisées à peu de groupes, un élément essentiel du schéma à deux étapes, qui est largement utilisé dans les simulations de réacteurs nucléaires. À mesure que les demandes industrielles nécessitent de plus en plus des maillages spatiaux et énergétiques fins pour améliorer la précision des calculs cœur, la taille de la bibliothèque des sections efficaces peut devenir excessive, entravant ainsi les performances des calculs cœur. Il est donc essentiel de développer une représentation qui minimise l'uti
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Szames, Esteban Alejandro. "Few group cross section modeling by machine learning for nuclear reactor." Thesis, université Paris-Saclay, 2020. http://www.theses.fr/2020UPASS134.

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Pour estimer la répartition de la puissance au sein d’un réacteur nucléaire, il est nécessaire de coupler des modélisations neutroniques et thermohydrauliques. De telles simulations doivent disposer des valeurs sections efficaces homogénéisées à peu de groupes d’énergies qui décrivent les interactions entre les neutrons et la matière. Cette thèse est consacrée à la modélisation des sections efficaces par des techniques académiques innovantes basées sur l’apprentissage machine. Les premières méthodes utilisent les modèles à noyaux du type RKHS (Reproducing Kernel Hilbert Space) et les secondes
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Cai, Li. "Condensation et homogénéisation des sections efficaces pour les codes de transport déterministes par la méthode de Monte Carlo : Application aux réacteurs à neutrons rapides de GEN IV." Thesis, Paris 11, 2014. http://www.theses.fr/2014PA112280/document.

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Dans le cadre des études de neutronique menées pour réacteurs de GEN-IV, les nouveaux outils de calcul des cœurs de réacteur sont implémentés dans l’ensemble du code APOLLO3® pour la partie déterministe. Ces méthodes de calculs s’appuient sur des données nucléaires discrétisée en énergie (appelées multi-groupes et généralement produites par des codes déterministes eux aussi) et doivent être validées et qualifiées par rapport à des calculs basés sur la méthode de référence Monte-Carlo. L’objectif de cette thèse est de mettre au point une technique alternative de production des propriétés nucléa
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Kim, Myung Hyun. "The use of bilinearly weighted cross sections for few-group transient analysis." Thesis, Massachusetts Institute of Technology, 1988. http://hdl.handle.net/1721.1/14375.

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Botes, Danniëll. "Few group cross section representation based on sparse grid methods / Danniëll Botes." Thesis, North-West University, 2012. http://hdl.handle.net/10394/8845.

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This thesis addresses the problem of representing few group, homogenised neutron cross sections as a function of state parameters (e.g. burn-up, fuel and moderator temperature, etc.) that describe the conditions in the reactor. The problem is multi-dimensional and the cross section samples, required for building the representation, are the result of expensive transport calculations. At the same time, practical applications require high accuracy. The representation method must therefore be efficient in terms of the number of samples needed for constructing the representation, storage requireme
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Capítulos de livros sobre o assunto "Few-Group Homogenized Cross Sections"

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Wang, Weixiang, WenPei Feng, KeFan Zhang, Guangliang Yang, Tao Ding, and Hongli Chen. "A Moose-Based Neutron Diffusion Code with Application to a LMFR Benchmark." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_43.

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AbstractMOOSE (Multiphysics Object-Oriented Simulation Environment) is a powerful finite element multi-physics coupling framework, whose object-oriented, extensive system is conducive to the development of various simulation tools. In this work, a full-core MOOSE-based Neutron Diffusion application is developed, and a 3D PWR benchmark 3D-IAEA with given group constants is applied for code verification. Then the MOOSE-based Neutron Diffusion application is applied to the calculation of a Sodium-cooled Fast Reactor (SFR) benchmark, together with the research on homogenized few-group constants ge
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Qin, Shuai, Qingming He, Jiahe Bai, Wenchang Dong, Liangzhi Cao, and Hongchun Wu. "Group Constants Generation Based on NECP-MCX Monte Carlo Code." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_9.

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AbstractThe reliability of few-group constants generated by lattice physics calculation is significant for the accuracy of the conventional two-step method in neutronics calculation. The deterministic method is preferred in the lattice calculation due to its efficiency. However, it is difficult for the deterministic method to treat the resonance self-shielding effect accurately and handle complex geometries. Compared to the deterministic method, the Monte Carlo method has the characteristics of using continuous-energy cross section and the powerful capability of geometric modeling. Therefore,
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Przybyła, Marcin, and Piotr Włodarczak. "Osada z okresu późnego eneolitu na stanowisku 12 w Kazimierzy Wielkiej (A settlement from the Late Eneolithic period at the site in Kazimierza Wielka)." In Kazimierza Wielka, stanowisko 12. Od neolitycznej osady do cmentarzyska z okresu wpływów rzymskich. Wydawnictwo Profil-Archeo, 2024. https://doi.org/10.33547/oda-sah.12.kaz.04.

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At an archaeological site in Kazimierza Wielka, the remains of the late phase of Funnel Beaker culture with Baden influences have been discovered. This period is sometimes referred to as the Funnel Beaker-Baden phase. Archaeological material associated with this phase has been recorded in the fills of 15 pits with trapezoidal or basin-shaped cross-sections. Almost all artefacts found in their fills are pottery fragments. The excavated features yielded a total of 746 pottery sherds and only single spindle whorls, as well as several bone and flint tools. The reconstructed vessel forms include mu
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Trabalhos de conferências sobre o assunto "Few-Group Homogenized Cross Sections"

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Bokov, P. M., D. Botes, and Kostadin Ivanov. "Hierarchical Interpolation of Homogenized Few-Group Neutron Cross-Sections on Samples with Uncorrelated Uncertainty." In International Conference on Physics of Reactors 2022. American Nuclear Society, 2022. http://dx.doi.org/10.13182/physor22-37615.

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Hu, Tianliang, Liangzhi Cao, Hongchun Wu, and Kun Zhuang. "Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67316.

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A code system has been developed in this paper for the dynamics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY and TANSY_K which is based on OpenFOAM to perform the steady-state full-core coupled simulations and dynamics simulation. For verification and application of the codes sequence, the simulation of a representative mol
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Zhang, Hongbo, Chuntao Tang, Weiyan Yang, Guangwen Bi, and Bo Yang. "Development and Verification of the PWR Lattice Code PANDA." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66573.

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Lattice code generates homogenized few-group cross sections for core neutronics code. It is an important component of the nuclear design code system. The development and improvement of lattice codes are always significant topics in reactor physics. The PANDA code is a PWR lattice code developed by Shanghai Nuclear Engineering Research and Design Institute (SNERDI). It starts from the 70-group library, and performs the resonance calculation based on the Spatially Dependent Dancoff Method (SDDM). The 2D heterogeneous transport calculation is performed without any group collapse and cell homogeni
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Ratti, Luca, Guido Mazzini, Marek Ruščák, and Valerio Giusti. "Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82607.

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The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios. This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE mode
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Nie, Jingyu, Binqian Li, Yingwei Wu, et al. "Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on OpenMC/COMSOL." In 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-135246.

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Abstract As an advanced small nuclear reactor, the heat pipe reactor possesses several advantages, including high energy density, long operational lifetime, compact size, and strong adaptability to various environments, making it an optimal choice for specialized energy needs in future applications, such as deep-sea and deep-space domains. In this study, we developed a code system using OpenMC/COMSOL for neutron and thermodynamic simulations. The continuous-energy Monte Carlo code, OpenMC, was employed to generate homogenized cross-section databases, offering significant modeling flexibility c
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Ahmed, Rizwan, Gyunyoung Heo, Dong-Keun Cho, and Jongwon Choi. "Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea." In ASME 2010 13th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2010. http://dx.doi.org/10.1115/icem2010-40201.

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Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning
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Rohde, U., S. Mittag, U. Grundmann, P. Petkov, and J. Ha´dek. "Application of a Step-Wise Verification and Validation Procedure to the 3D Neutron Kinetics Code DYN3D Within the European NURESIM Project." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75446.

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A generic strategy of core physics codes benchmarking was elaborated within the European NURESIM code platform development. In this paper, the application of this step-wise procedure to benchmarking the 3D neutron kinetics code DYN3D for applications to VVER-type reactors is described. Numerical and experimental benchmark problems were considered for code verification and validation. Examples of these benchmarks including benchmark set-up and results obtained by use of DYN3D in comparison with other codes are given. First, mathematical problems with given cross sections are used for the verifi
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Yuan, Yuan, Guoming Liu, and Peng Zhang. "Verification of the RMC-SaraGR Nuclear Design Code System Based on the HTTR Benchmark." In 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-135368.

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Abstract In order to perform a detailed nuclear analysis for the small modular prismatic HTGRs, the RMC-SaraGR nuclear design code system has been developed. The verification of the code system using the HTTR benchmark is reported in this paper. The detailed HTTR models have been established with the continuous-energy Monte Carlo code RMC to provide the reference solutions and the 25-group cross sections, which are further corrected by the super homogenization (SPH) factors before being used for the homogeneous core transport calculations by SaraGR or the multi-group RMC (RMC-MG). Various core
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Yang, Wankui, Baoxin Yuan, Songbao Zhang, Haibing Guo, Yaoguang Liu, and Li Deng. "A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81709.

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Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor t
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Jevremovic, Tatjana, Mathieu Hursin, Nader Satvat, John Hopkins, Shanjie Xiao, and Godfree Gert. "Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89561.

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The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteris
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