Semi-implicit Direct Kinetics Methodology for Deterministic, Time-dependent, Three-dimensional, and Fine-energy Neutron Transport Solutions PDF Download
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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: 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: Publisher: ISBN: Category : Languages : en Pages : 8
Book Description
A deterministic method for solving the time-dependent, three-dimensional Boltzmann transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multi-dimensional neutronic systems.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems.
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: Publisher: ISBN: Category : Languages : en Pages :
Book Description
A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments.
Author: Mathieu Hursin Publisher: ISBN: Category : Languages : en Pages : 264
Book Description
The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, burnup, control rod position, etc ...) anticipated during the transient. The core calculation is then performed using the few group cross sections in a core simulator which uses some type of coarse mesh nodal method. The multi-step approach was identified as inadequate for several applications such as the design of MOX cores and other highly hetereogeneous, high leakage core designs. Because of the considerable advances in computing power over the last several years, there has been interest in high-fidelity solutions of the Boltzmann Transport equation. A practical approach developed for high-fidelity solutions of the 3D transport equation is the 2D-1D methodology in which the method of characteristics (MOC) is applied to the heterogeneous 2D planar problem and a lower order solution is applied to the axial problem which is, generally, more uniform. This approach was implemented in the DeCART code. Recently, there has been interest in extending such approach to the simulations of design basis accidents, such as control rod ejection accidents also known as reactivity initiated accidents (RIA). The current 2D-1D algorithm available in DeCART only provide 1D axial solution based on the diffusion theory whose accuracy deteriorates in case of strong flux gradient that can potentially be observed during RIA simulations. The primary ojective of the dissertation is to improve the accuracy and range of applicability of the DeCART code and to investigate its ability to perform a full core transient analysis of a realistic RIA. The specific research accomplishments of this work include: * The addition of more accurate 2D-1D coupling and transverse leakage splitting options to avoid the occurrence of negative source terms in the 2D MOC equations and the subsequent failure of the DeCART calculation and the improvement of the convergence of the 2D-1D method. * The implementation of a higher order transport axial solver based on NEM-Sn derivation of the Boltzmann equation. * Improved handling of thermal hydraulic feedbacks by DeCART during transient calculations. * A consistent comparison of the DeCART transient methodology with the current multistep approach (PARCS) for a realistic full core RIA. An efficient direct whole core transport calculation method involving the NEM-Sn formulation for the axial solution and the MOC for the 2-D radial solution was developed. In this solution method, the Sn neutron transport equations were developed within the framework of the Nodal Expansion Method. A RIA analysis was performed and the DeCART results were compared to the current generation of LWR core analysis methods represented by the PARCS code. In general there is good overall agreement in terms of global information from DeCART and PARCS for the RIA considered. However, the higher fidelity solution in DeCART provides a better spatial resolution that is expected to improve the accuracy of fuel performance calculations and to enable reducing the margin in several important reactor safety analysis events such as the RIA.
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: James Ernest Banfield
Author: James Ernest Banfield
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Author: Bengt G. Carlson
Author: Mathieu Hursin
Author: Liangzhi Cao
Author: Sedat Goluoglu
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