Download or read book Deterministic Numerical Methods for Unstructured Mesh Neutron Transport Calculation written by Liangzhi Cao and published by Woodhead Publishing. This book was released on 2020-08-30 with total page 294 pages. Available in PDF, EPUB and Kindle. Book excerpt: Deterministic Numerical Methods for Unstructured-Mesh Neutron Transport Calculation presents the latest deterministic numerical methods for neutron transport equations (NTEs) with complex geometry, which are of great demand in recent years due to the rapid development of advanced nuclear reactor concepts and high-performance computational technologies. This book covers the wellknown methods proposed and used in recent years, not only theoretical modeling but also numerical results. This book provides readers with a very thorough understanding of unstructured neutron transport calculations and enables them to develop their own computational codes. The fundamentals, numerical discretization methods, algorithms, and numerical results are discussed. Researchers and engineers from utilities and research institutes are provided with examples on how to model an advanced nuclear reactor, which they can then apply to their own research projects and lab settings. - Combines the theoretical models with numerical methods and results in one complete resource - Presents the latest progress on the topic in an easy-to-navigate format

Download or read book Resonance Self Shielding Calculation Methods in Nuclear Reactors written by Liangzhi Cao and published by Woodhead Publishing. This book was released on 2022-10-01 with total page 412 pages. Available in PDF, EPUB and Kindle. Book excerpt: Resonance Self-Shielding Calculation Methods in Nuclear Reactors presents the latest progress in resonance self-shielding methods for both deterministic and Mote Carlo methods, including key advances over the last decade such as high-fidelity resonance treatment, resonance interference effect and multi-group equivalence. As the demand for high-fidelity resonance self-shielding treatment is increasing due to the rapid development of advanced nuclear reactor concepts and progression in high performance computational technologies, this practical book guides students and professionals in nuclear engineering and technology through various methods with proven high precision and efficiency. - Presents a collection of resonance self-shielding methods, as well as numerical methods and numerical results - Includes new topics in resonance self-shielding treatment - Provides source codes of key calculations presented

Download or read book Numerical Methods in the Theory of Neutron Transport written by Guriĭ Ivanovich Marchuk and published by Harwood Academic Publishers. This book was released on 1986 with total page 632 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Numerical Formulation and Solution of Neutron Transport Problems written by Bengt G. Carlson and published by . This book was released on 1964 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Numerical Solution of Transient and Steady state Neutron Transport Problems written by Bengt G. Carlson and published by . This book was released on 1959 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Verification Validation of High Order Short Characteristics Based Deterministic Transport Methodology on Unstructured Grids written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code's numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory's Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above- listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

Download or read book Computational Methods of Neutron Transport written by Elmer Eugene Lewis and published by Wiley-Interscience. This book was released on 1984 with total page 440 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book The Discrete Sn Approximation to Transport Theory written by Clarence E. Lee and published by . This book was released on 1962 with total page 434 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book On the Numerical Integration of the Neutron Transport Equation written by Herbert Bishop Keller and published by . This book was released on 1955 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt: A procedure for the direct numerical integration of the steady-state, elastic scattering neutron transport equation is presented.

Download or read book Deterministic and Monte Carlo Neutron Transport Calculation for Greifswald 1 and Comparison with Ex vessel Measured Data written by N. Khrennikov and published by . This book was released on 2006 with total page 7 pages. Available in PDF, EPUB and Kindle. Book excerpt: The results of a study of neutron and gamma field functionals derived by deterministic Sn and Monte Carlo calculation methods and by neutron activation measurements in application to the ex-vessel cavity of the VVER-440 reactor Greifswald-1 are presented. A good agreement of deterministic and stochastic calculation results with each other as well as with measurement results was found for neutron threshold detector reaction rates at ex-vessel positions. The influence of different numbers of cross-sectional groups on the calculation results is demonstrated.

Download or read book Deterministic Methods for Time dependent Stochastic Neutron Transport written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A numerical method is presented for solving the time-dependent survival probability equation in general (lD/2D/3D) geometries using the multi group SNmethod. Although this equation was first formulated by Bell in the early 1960's, it has only been applied to stationary systems (for other than idealized point models) until recently, and detailed descriptions of numerical solution techniques are lacking in the literature. This paper presents such a description and applies it to a dynamic system representative of a figurative criticality accident scenario.

Download or read book Numerical Methods for Neutron Transport Calculations of Nuclear Reactors written by Andrea Barbarino and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book A Solution of the Neutron Transport Equation written by J. Certaine and published by . This book was released on 1955 with total page 140 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Some numerical methods for solving the neutron transport equation written by C. Budd and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Performance Modeling of Unstructured Mesh Particle Transport Computations written by and published by . This book was released on 2004 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: The performance of unstructured mesh applications presents a number of complexities and subtleties that do not arise for dense structured meshes. From a prograniming point of view, handling an unstructured mesh has an increased complexity to manage the necessary data structures and interactions between mesh-cells. From a performance point of view, there are added dificulties in understanding both the processing time on a single processor and the scaling characteristics. In this work we present a performance model for the calculation of deterministic SN transport on unstructured meshes. It builds upon earlier work that successfully modeled the same calculation on structured meshes. The model captures the key processing characteristics and is parametric using both the system performance data (latency, bandwidth, processing rate etc.) and application data (mesh size etc.) as input. The model is validated on two clusters [an HP Alphaserver and an Itanium-2 system] showing high accuracy. Importantly it is also shown that a single formulation of the model can be used to predict the performance of two quite different implementations of the same calculation.

Download or read book A Method of Moments for Solving the Neutron Transport Equation written by Bengt G. Carlson and published by . This book was released on 1965 with total page 50 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book A Novel Equivalence Method for High Fidelity Hybrid Stochastic deterministic Neutron Transport Simulations written by Guillaume Louis Giudicelli and published by . This book was released on 2020 with total page 542 pages. Available in PDF, EPUB and Kindle. Book excerpt: With ever increasing available computing resources, the traditional nuclear reactor physics computation schemes, that trade off between spatial, angular and energy resolution to achieve low cost highly-tuned simulations, are being challenged. While existing schemes can reach few-percent accuracy for the current fleet of light water reactors, thanks to a plethora of astute engineering approximations, they cannot provide sufficient accuracy for evolutionary reactor designs with highly heterogeneous geometries. The decades-long process to develop and qualify these simulation tools is also not in phase with the fast-paced development of innovative new reactor designs seeking to address the climate crisis. Enabled by those computing resources, high fidelity Monte Carlo methods can easily tackle challenging geometries, but they lack the computational and algorithmic efficiency of deterministic methods. However, they are increasingly being used for group cross section generation. Downstream highly parallelized 3D deterministic transport can then use those cross sections to compute accurate solutions at the full core scale. This hybrid computation scheme makes the most of both worlds to achieve fast and accurate reactor physics simulations. Among the few remaining approximations are neglecting the angular dependence of group cross sections, which lead to an over-estimation of resonant absorption rates, especially for the lower resonances of 238U. This thesis presents a novel equivalence method based on introducing discontinuities in the track angular fluxes, with a polar dependence of discontinuity factors to preserve the polar dependence of the neutron currents as well as removing the self-shielding error. This new method is systematically benchmarked against the state-of-the-art method, SuPerHomogenization in three different approaches to obtaining equivalence factors: a same-scale iterative approach, a multiscale approach, and a single-step non-iterative approach. Both methods show remarkable agreement with a reference Monte Carlo solution on a wide array of test cases, from 2D pin cells to 3D full core calculations, for the iterative and the multi-scale approaches. The self-shielding error is eliminated, improving significantly the predictive capabilities of the scheme for the distribution of 238U absorption in the core. A single-step non-iterative approach to obtaining equivalence factors is also pursued, and was shown to only be adequate with the novel discontinuity factor-based method. This study is largely enabled by a significant optimization effort of the 3D deterministic neutron transport solver. By leveraging low level parallelism through vectorization of the multi-group neutron transport equation, by increasing the memory locality of the method of characteristics implementation and with a novel inter-domain communication algorithm enabling a near halving of memory requirements, the 3D full core case can now be tackled with only 50 nodes on an industrial sized computing cluster rather than the many thousands of nodes on a TOP20 supercomputer used previously. This thesis presents fully resolved solutions to the steady-state multi-group neutron transport equation for full-core 3D light water reactors, and these solutions are comparable to gold-standard continuous-energy Monte Carlo solutions.