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Author: Samuel Christopher Shaner Publisher: ISBN: Category : Languages : en Pages : 140
Book Description
Monte Carlo is increasingly being used to perform high-fidelity, steady-state neutronics analysis of power reactor geometries on today’s leadership class supercomputers. Extending Monte Carlo to time dependent problems has proven to be a formidable challenge due to the significant computational resource and data processing requirements. In this thesis, a transient methodology is proposed and implemented to enable accurate and computationally tractable time dependent Monte Carlo analysis. The frequency transform method has been described and implemented in Monte Carlo for the first time. The attractiveness of this method lies in its ability to accurately capture the space and time dependent distribution of the delayed neutron source throughout a transient. Nuances to the algorithmic implementation are described and validated through a series of simple analytical test problems. Comparison with the adiabatic method currently employed for Monte Carlo transient analysis shows significant improvement in the spatial distribution and magnitude of the power for a negative reactivity insertion transient in the 2D and 3D C5G7 geometry. To aid in understanding the effect of statistical uncertainty in the tallied quantities on the time dependent flux solution, a simplified point kinetics model was developed and used for insightful analysis on simple transient test problems. This revealed how the time dependent flux profiles for a series of independent trials can be approximated by a normal distribution at low uncertainties in the tallied reactivity, but deviates from a normal distribution at relatively modest uncertainties in reactivity. Given the compuational constraints of solving large problems, having a simple model that can provide insight on the expected behavior and flux distribution can be very valuable. The frequency transform methodology belongs to a class of indirect space-time factorization methods that perform high-order calculations (e.g. Monte Carlo) over long time steps and low-order, computationally-efficient calculations (e.g. Point Kinetics) over short time steps as an approach to balance performance and accuracy. The coarse mesh finite difference (CMFD) diffusion operator is employed as the low-order solver in Monte Carlo transient analysis for the first time. The CMFD diffusion operator is attractive due to its potential to increase the time step size between the computationally expensive high-order solves. Implementing this methodology is important as continuous energy Monte Carlo is reactor-agnostic and able to treat complex geometries without difficulty, opening up the possibility of solving transients on new experimental geometries for which there is little data.
Author: Samuel Christopher Shaner Publisher: ISBN: Category : Languages : en Pages : 140
Book Description
Monte Carlo is increasingly being used to perform high-fidelity, steady-state neutronics analysis of power reactor geometries on today’s leadership class supercomputers. Extending Monte Carlo to time dependent problems has proven to be a formidable challenge due to the significant computational resource and data processing requirements. In this thesis, a transient methodology is proposed and implemented to enable accurate and computationally tractable time dependent Monte Carlo analysis. The frequency transform method has been described and implemented in Monte Carlo for the first time. The attractiveness of this method lies in its ability to accurately capture the space and time dependent distribution of the delayed neutron source throughout a transient. Nuances to the algorithmic implementation are described and validated through a series of simple analytical test problems. Comparison with the adiabatic method currently employed for Monte Carlo transient analysis shows significant improvement in the spatial distribution and magnitude of the power for a negative reactivity insertion transient in the 2D and 3D C5G7 geometry. To aid in understanding the effect of statistical uncertainty in the tallied quantities on the time dependent flux solution, a simplified point kinetics model was developed and used for insightful analysis on simple transient test problems. This revealed how the time dependent flux profiles for a series of independent trials can be approximated by a normal distribution at low uncertainties in the tallied reactivity, but deviates from a normal distribution at relatively modest uncertainties in reactivity. Given the compuational constraints of solving large problems, having a simple model that can provide insight on the expected behavior and flux distribution can be very valuable. The frequency transform methodology belongs to a class of indirect space-time factorization methods that perform high-order calculations (e.g. Monte Carlo) over long time steps and low-order, computationally-efficient calculations (e.g. Point Kinetics) over short time steps as an approach to balance performance and accuracy. The coarse mesh finite difference (CMFD) diffusion operator is employed as the low-order solver in Monte Carlo transient analysis for the first time. The CMFD diffusion operator is attractive due to its potential to increase the time step size between the computationally expensive high-order solves. Implementing this methodology is important as continuous energy Monte Carlo is reactor-agnostic and able to treat complex geometries without difficulty, opening up the possibility of solving transients on new experimental geometries for which there is little data.
Author: Miriam A. Kreher Publisher: ISBN: Category : Languages : en Pages : 0
Book Description
Monte Carlo neutron transport is the gold standard for accurate neutronics simulation of nuclear reactors in steady-state because each term of the neutron transport equation can be directly tallied using continuous-energy cross sections rather than needing to make approximations in energy, angle, or geometry. However, the time dependent equation includes time derivatives of flux and delayed neutron precursors which are difficult to tally. While it is straightforward to explicitly model delayed neutron precursors, and thus solve the time dependent problem in Direct Monte Carlo, this is such a costly approach that the practical length of transient calculations is limited to about 1 second. In order to solve longer problems, a high-order/low-order approach was adopted that uses the omega method to approximate the time derivatives as frequencies. These frequencies are spatially distributed and provided by a low-order Time Dependent Coarse Mesh Finite Difference diffusion solver. While this scheme has been previously applied to prescribed transients, thermal feedback is now incorporated to provide a fully self-propagating Monte Carlo transient multiphysics solver which can be applied to transients of several seconds long. Several recently developed techniques are used in the implementation of the proposed coupling approaches. Firstly, underrelaxed Monte Carlo, which is a steady-state technique that stabilizes the search for temperature distributions, is applied to find initial conditions. Secondly, tally derivatives are a Monte Carlo perturbation technique that can identify how a tally will change with respect to a small change in the system. Test problems of varying complexity are carried out in flow-initiated transients to show the versatility of these methods. Overall, this multi-level, multiphysics, transient solver provides a bridge between high fidelity Monte Carlo neutronics and the fast multi-group diffusion methods that are currently used in safety analysis.
Author: Nicholas Edward Horelik Publisher: ISBN: Category : Languages : en Pages : 158
Book Description
Monte Carlo (MC) neutral particle transport methods have long been considered the gold-standard for nuclear simulations, but high computational cost has limited their use significantly. However, as we move towards higher-fidelity nuclear reactor analyses the method has become competitive with traditional deterministic transport algorithms for the same level of accuracy, especially considering the inherent parallelism of the method and the ever-increasing concurrency of modern high performance computers. Yet before such analysis can be practical, several algorithmic challenges must be addressed, particularly in regards to the memory requirements of the method. In this thesis, a robust domain decomposition algorithm is proposed to alleviate this, along with models and analysis to support its use for full-scale reactor analysis. Algorithms were implemented in the full-physics Monte Carlo code OpenMC, and tested for a highly-detailed PWR benchmark: BEAVRS. The proposed domain decomposition implementation incorporates efficient algorithms for scalable inter-domain particle communication in a manner that is reproducible with any pseudo-random number seed. Algorithms are also proposed to scalably manage material and tally data with on-the-fly allocation during simulation, along with numerous optimizations required for scalability as the domain mesh is refined and divided among thousands of compute processes. The algorithms were tested on two supercomputers, namely the Mira Blue Gene/Q and the Titan XK7, demonstrating good performance with realistic tallies and materials requiring over a terabyte of aggregate memory. Performance models were also developed to more accurately predict the network and load imbalance penalties that arise from communicating particles between distributed compute nodes tracking different spatial domains. These were evaluated using machine properties and tallied particle movement characteristics, and empirically validated with observed timing results from the new implementation. Network penalties were shown to be almost negligible with per-process particle counts as low as 1000, and load imbalance penalties higher than a factor of four were not observed or predicted for finer domain meshes relevant to reactor analysis. Load balancing strategies were also explored, and intra-domain replication was shown to be very effective at improving parallel efficiencies without adding significant complexity to the algorithm or burden to the user. Performance of the strategy was quantified with a performance model, and shown to agree well with observed timings. Imbalances were shown to be almost completely removed for the finest domain meshes. Finally, full-core studies were carried out to demonstrate the efficacy of domain-decomposed Monte Carlo in tackling the full scope of the problem. A detailed mesh required for a robust depletion treatment was used, and good performance was demonstrated for depletion tallies with 206 nuclides. The largest runs scored six reaction rates for each nuclide in 51M regions for a total aggregate memory requirement of 1.4TB, and particle tracking rates were consistent with those observed for smaller non-domain- decomposed runs with equivalent tally complexity. These types of runs were previously not achievable with traditional Monte Carlo methods, and can be accomplished with domain decomposition with between 1.4x and 1.75x overhead with simple load balancing.
Author: Christophe Demazière Publisher: Academic Press ISBN: 012815070X Category : Technology & Engineering Languages : en Pages : 370
Book Description
Modelling of Nuclear Reactor Multiphysics: From Local Balance Equations to Macroscopic Models in Neutronics and Thermal-Hydraulics is an accessible guide to the advanced methods used to model nuclear reactor systems. The book addresses the frontier discipline of neutronic/thermal-hydraulic modelling of nuclear reactor cores, presenting the main techniques in a generic manner and for practical reactor calculations.The modelling of nuclear reactor systems is one of the most challenging tasks in complex system modelling, due to the many different scales and intertwined physical phenomena involved. The nuclear industry as well as the research institutes and universities heavily rely on the use of complex numerical codes. All the commercial codes are based on using different numerical tools for resolving the various physical fields, and to some extent the different scales, whereas the latest research platforms attempt to adopt a more integrated approach in resolving multiple scales and fields of physics. The book presents the main algorithms used in such codes for neutronic and thermal-hydraulic modelling, providing the details of the underlying methods, together with their assumptions and limitations. Because of the rapidly expanding use of coupled calculations for performing safety analyses, the analysists should be equally knowledgeable in all fields (i.e. neutron transport, fluid dynamics, heat transfer).The first chapter introduces the book's subject matter and explains how to use its digital resources and interactive features. The following chapter derives the governing equations for neutron transport, fluid transport, and heat transfer, so that readers not familiar with any of these fields can comprehend the book without difficulty. The book thereafter examines the peculiarities of nuclear reactor systems and provides an overview of the relevant modelling strategies. Computational methods for neutron transport, first at the cell and assembly levels, then at the core level, and for one-/two-phase flow transport and heat transfer are treated in depth in respective chapters. The coupling between neutron transport solvers and thermal-hydraulic solvers for coarse mesh macroscopic models is given particular attention in a dedicated chapter. The final chapter summarizes the main techniques presented in the book and their interrelation, then explores beyond state-of-the-art modelling techniques relying on more integrated approaches. - Covers neutron transport, fluid dynamics, and heat transfer, and their interdependence, in one reference - Analyses the emerging area of multi-physics and multi-scale reactor modelling - Contains 71 short videos explaining the key concepts and 77 interactive quizzes allowing the readers to test their understanding
Author: Jun Wang Publisher: Woodhead Publishing ISBN: 0128181915 Category : Technology & Engineering Languages : en Pages : 612
Book Description
Nuclear Power Plant Design and Analysis Codes: Development, Validation, and Application presents the latest research on the most widely used nuclear codes and the wealth of successful accomplishments which have been achieved over the past decades by experts in the field. Editors Wang, Li,Allison, and Hohorst and their team of authors provide readers with a comprehensive understanding of nuclear code development and how to apply it to their work and research to make their energy production more flexible, economical, reliable and safe.Written in an accessible and practical way, each chapter considers strengths and limitations, data availability needs, verification and validation methodologies and quality assurance guidelines to develop thorough and robust models and simulation tools both inside and outside a nuclear setting. This book benefits those working in nuclear reactor physics and thermal-hydraulics, as well as those involved in nuclear reactor licensing. It also provides early career researchers with a solid understanding of fundamental knowledge of mainstream nuclear modelling codes, as well as the more experienced engineers seeking advanced information on the best solutions to suit their needs. - Captures important research conducted over last few decades by experts and allows new researchers and professionals to learn from the work of their predecessors - Presents the most recent updates and developments, including the capabilities, limitations, and future development needs of all codes - Incudes applications for each code to ensure readers have complete knowledge to apply to their own setting
Author: Matthew Shawn Ellis Publisher: ISBN: Category : Languages : en Pages : 321
Book Description
The ability to model and simulate nuclear reactors during steady state and transient conditions is important for designing efficient and safe nuclear power systems. The accurate simulation of a nuclear reactor is particularly challenging because the multiple physical processes within the reactor are tightly coupled, which requires that the numerical methods used to resolve each physical process can accurately and efficiently transfer and utilize data from other applications. Monte Carlo methods are desirable for solving the neutron transport equation required in reactor analysis because of the inherent accuracy of the method, but the Computational Solid Geometry (CSG) representation of the physical geometry makes it difficult to accurately and efficiently perform multiphysics reactor analyses with other applications that utilize finite element or finite volume representations. To address this limitation, a multiphysics coupling framework that minimizes the need for spatial discretization in the Monte Carlo geometry is presented in this thesis. The coupling framework uses Functional Expansion Tallies to transfer multiphysics information from the Monte Carlo application to other multiphysics tools. Additionally, the coupling framework uses a modified method for transporting neutrons through spatially continuous total macroscopic cross section distributions in order to incorporate continuous multiphysics feedback fields such as fuel temperature and coolant density into the Monte Carlo simulation. It has been shown that separable Zernike and Legendre Function Expansion Tallies can effectively reconstruct a continuous distribution of fission power density. Additionally, using a prototypical three-dimensional Light Water Reactor pin cell, the method used to transport neutrons through a continuously varying fuel temperature and coolant density distribution was shown to be 1.7 times faster than a comparable discretized simulation with volume-averaged properties, while still providing a high level of accuracy. Finally, in order to make the overall multiphysics coupling scheme useful for reactor analyses, a novel spatially continuous depletion methodology was developed and investigated. With the spatially continuous depletion methodology, number densities can be represented as a linear combination of polynomials, and those polynomial representations can be integrated through time to predict reactor operation. The spatially continuous depletion methodology was able to accurately predict the eigenvalue and number density distributions in a two-dimensional LWR pin cell depletion containing Gd-157 from a 2 weight percent GdO2 and seven other nuclides in the depletion matrix. Analyses of the spatially continuous depletion methodology showed that significant reductions in the number of tallied values could be achieved if polynomial representations were optimized for each nuclide reaction rate. From the depletion simulations in this thesis, a 23% reduction in the required number of reaction rate tallies compared to a lower-fidelity, 10 radial ring pin discretization was shown to be achievable with nuclide polynomial optimization. In addition to showing potential for reductions in tally memory and computational requirements, the spatially continuous depletion simulation was shown to be equal in computational performance to a discrete simulation with 10 radial rings and 8 azimuthal cuts, while providing a much higher level of spatial fidelity in number density concentrations.
Author: Michael Ljungberg Publisher: CRC Press ISBN: 1439841098 Category : Medical Languages : en Pages : 361
Book Description
From first principles to current computer applications, Monte Carlo Calculations in Nuclear Medicine, Second Edition: Applications in Diagnostic Imaging covers the applications of Monte Carlo calculations in nuclear medicine and critically reviews them from a diagnostic perspective. Like the first edition, this book explains the Monte Carlo method and the principles behind SPECT and PET imaging, introduces the reader to some Monte Carlo software currently in use, and gives the reader a detailed idea of some possible applications of Monte Carlo in current research in SPECT and PET. New chapters in this edition cover codes and applications in pre-clinical PET and SPECT. The book explains how Monte Carlo methods and software packages can be applied to evaluate scatter in SPECT and PET imaging, collimation, and image deterioration. A guide for researchers and students developing methods to improve image resolution, it also demonstrates how Monte Carlo techniques can be used to simulate complex imaging systems.
Author: Samuel Christopher Shaner
Author: Samuel Christopher Shaner
Author: Jingang Liang
Author: Miriam A. Kreher
Author: Nicholas Edward Horelik
Author: Christophe Demazière
Author: Jun Wang
Author: Matthew Shawn Ellis
Author: Bart Laurens Sjenitzer
Author: Michael Ljungberg
Author: