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Author: Liangzhi Cao Publisher: Woodhead Publishing ISBN: 0128182229 Category : Technology & Engineering Languages : en Pages : 294
Book Description
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
Author: Liangzhi Cao Publisher: Woodhead Publishing ISBN: 0128182229 Category : Technology & Engineering Languages : en Pages : 294
Book Description
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
Author: Liangzhi Cao Publisher: Woodhead Publishing ISBN: 0323858759 Category : Science Languages : en Pages : 412
Book Description
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
Author: Derek Ray Gaston Publisher: ISBN: Category : Languages : en Pages : 224
Book Description
One important goal in nuclear reactor core simulations is the computation of detailed 3D power distributions that will enable higher confidence in licensing of next-generation reactors and lifetime extensions/power up-rates for current-generation reactors. To date, there have been only a few demonstrations of such high-fidelity deterministic neutron transport calculations. However, as computational power continues to grow, such capabilities continue to move closer to being practically realized. Predictive reactor physics needs both neutronics calculations and full-core, 3D coupled multiphysics simulations (e.g., neutronics, fuel performance, fluid mechanics, structural mechanics). Therefore, new reactor physics tools should harness supercomputers to enable full-core reactor simulations and be capable of coupling for multiphysics feedback. One candidate for full-core nuclear reactor neutronics is the method of characteristics (MOC). Recent advancements have seen a pellet-resolved 3D MOC solution for the BEAVRS benchmark. However, MOC is traditionally implemented using constructive solid geometry (CSG) that makes it difficult (if not impossible) to accurately deform material to capture physical feedback effects such as fuel pin thermal expansions, assembly bowings, or core flowering. An alternative to CSG is to use unstructured, finite-element mesh for spatial discretization of MOC. Such mesh-based geometries permit directly linking to unstructured mesh-based multiphysics tools, such as fuels performance. Utilizing unstructured mesh has been attempted in the past, but those attempts have fallen short of producing usable 3D reactor simulators. Several key issues have hindered these attempts: lack of fuel volume preservation, approximations of boundary conditions, inefficient spatial domain decompositions, excessive memory requirements, ineffective parallel load balancing, and lack of scalability on massively parallel modern computer clusters. This thesis resolves these issues by developing a massively parallel, 3D, full-core MOC code, called MOCkingbird, using unstructured meshes. Underpinning MOCkingbird is a new algorithm for parallel ray tracing: the Scalable Massively Asynchronous Ray Tracing (SMART) algorithm. This algorithm enables efficient parallel ray-tracing across the full reactor domain, alleviating issues of reduced convergence associated with standard parallel MOC algorithms. In addition, to enable full-core simulation using unstructured mesh MOC, several new algorithms are developed, including reactor mesh generation, sparse parallel communication, parallel cyclic track generation, and weighted partitioning. Within this work MOCkingbird and SMART are tested for scalability from 10 to 20,000 cores on the Lemhi supercomputer at Idaho National Laboratory. Accuracy is tested using a suite of benchmarks that ultimately culminate in a first-of-a-kind, 3D, full-core, simulation of the BEAVRS benchmark using unstructured mesh MOC.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
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.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
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.
Author: James Ernest Banfield Publisher: ISBN: Category : Neutrons Languages : en Pages : 170
Book Description
Using a semi-implicit direct kinetics (SIDK) method that is developed in this dissertation, a finer neutron energy discretization and improved fidelity for transient radiation transport calculations are facilitated to reduce uncertainties and conservatisms in transient power and temperature predictions. These capabilities are implemented within a parallel computational solver framework, which is able to represent an arbitrary number of neutron energy groups, angles, and spatial discretizations, while internally coupled to an unstructured finite element multi-physics code for temperature and displacement calculations. This capability is demonstrated on a three-dimensional control rod ejection simulation run in parallel utilizing forty-four neutron energy groups. An improved transient nuclear reactor simulation capability is developed by adapting the steady-state radiation transport code Denovo to solve the time-dependent Boltzmann transport equation for transient power distributions. The developed SIDK method is compared to fully-implicit direct kinetics, higher order time integration methods, as well as various computational benchmarks. Errors resulting from time integration, spatial discretization, angular treatment, multi-group treatment, homogenization of temperature, and power over the time step representation are explored. For verification, the SIDK method is developed and tested externally and independently employing a few-group time-dependent neutron diffusion code which is compared to one and two-dimensional benchmarks with and without temperature feedbacks. The results of the semi-implicit direct kinetics method (SIDK) are shown to be accurate to within ~0.2% of direct kinetics and to execute roughly an order of magnitude faster, using a consistent space and time discretization. For sufficiently severe transients, the direct method is shown to produce lower errors with medium time steps than the SIDK method with fine steps, but proves to be subject to more severe oscillations at very coarse time steps than the SIDK method, in addition to producing similar errors (within 0.2 %) at medium spatial discretization with consistent time steps. The objective of this dissertation is to provide developers of next generation high-performance computing neutron kinetics methods a guide to the benefits and costs of the dominant discretization strategies of time, space, neutron energy, and angle for the solution of the time-dependent Boltzmann transport equation.
Author: Liangzhi Cao
Author: Liangzhi Cao
Author: Liangzhi Cao
Author: Guriĭ Ivanovich Marchuk
Author: Bengt G. Carlson
Author: Derek Ray Gaston
Author: Bengt G. Carlson
Author: 
Author: Elmer Eugene Lewis
Author: 
Author: James Ernest Banfield