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Languages : en
Pages : 12
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SCALE is a broadly used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 12
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Book Description
SCALE is a broadly used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
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Languages : en
Pages :
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Monte Carlo tools in SCALE are commonly used in criticality safety calculations as well as sensitivity and uncertainty analysis, depletion, and criticality alarm system analyses. Recent improvements in the continuous-energy data generated by the AMPX code system and significant advancements in the continuous-energy treatment in the KENO Monte Carlo eigenvalue codes facilitate the use of SCALE Monte Carlo codes to model geometrically complex systems with enhanced solution fidelity. The addition of continuous-energy treatment to the SCALE Monaco code, which can be used with automatic variance reduction in the hybrid MAVRIC sequence, provides significant enhancements, especially for criticality alarm system modeling. This paper describes some of the advancements in continuous-energy Monte Carlo codes within the SCALE code system.
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Languages : en
Pages : 39
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This report describes the MCV (Monte Carlo - Vectorized)Monte Carlo neutron transport code [Brown, 1982, 1983; Brown and Mendelson, 1984a]. MCV is a module in the RACER system of codes that is used for Monte Carlo reactor physics analysis. The MCV module contains all of the neutron transport and statistical analysis functions of the system, while other modules perform various input-related functions such as geometry description, material assignment, output edit specification, etc. MCV is very closely related to the 05R neutron Monte Carlo code [Irving et al., 1965] developed at Oak Ridge National Laboratory. 05R evolved into the 05RR module of the STEMB system, which was the forerunner of the RACER system. Much of the overall logic and physics treatment of 05RR has been retained and, indeed, the original verification of MCV was achieved through comparison with STEMB results. MCV has been designed to be very computationally efficient [Brown, 1981, Brown and Martin, 1984b; Brown, 1986]. It was originally programmed to make use of vector-computing architectures such as those of the CDC Cyber- 205 and Cray X-MP. MCV was the first full-scale production Monte Carlo code to effectively utilize vector-processing capabilities. Subsequently, MCV was modified to utilize both distributed-memory [Sutton and Brown, 1994] and shared memory parallelism. The code has been compiled and run on platforms ranging from 32-bit UNIX workstations to clusters of 64-bit vector-parallel supercomputers. The computational efficiency of the code allows the analyst to perform calculations using many more neutron histories than is practical with most other Monte Carlo codes, thereby yielding results with smaller statistical uncertainties. MCV also utilizes variance reduction techniques such as survival biasing, splitting, and rouletting to permit additional reduction in uncertainties. While a general-purpose neutron Monte Carlo code, MCV is optimized for reactor physics calculations. It has the capability of performing iterated-source (criticality), multiplied-fixed-source, and fixed-source calculations. MCV uses a highly detailed continuous-energy (as opposed to multigroup) representation of neutron histories and cross section data. The spatial modeling is fully three-dimensional (3-D), and any geometrical region that can be described by quadric surfaces may be represented. The primary results are region-wise reaction rates, neutron production rates, slowing-down-densities, fluxes, leakages, and when appropriate the eigenvalue or multiplication factor. Region-wise nuclidic reaction rates are also computed, which may then be used by other modules in the system to determine time-dependent nuclide inventories so that RACER can perform depletion calculations. Furthermore, derived quantities such as ratios and sums of primary quantities and/or other derived quantities may also be calculated. MCV performs statistical analyses on output quantities, computing estimates of the 95% confidence intervals as well as indicators as to the reliability of these estimates. The remainder of this chapter provides an overview of the MCV algorithm. The following three chapters describe the MCV mathematical, physical, and statistical treatments in more detail. Specifically, Chapter 2 discusses topics related to tracking the histories including: geometry modeling, how histories are moved through the geometry, and variance reduction techniques related to the tracking process. Chapter 3 describes the nuclear data and physical models employed by MCV. Chapter 4 discusses the tallies, statistical analyses, and edits. Chapter 5 provides some guidance as to how to run the code, and Chapter 6 is a list of the code input options.
Author: Andreas Kling
Publisher: Springer Science & Business Media
ISBN: 3642182119
Category : Science
Languages : en
Pages : 1200
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Book Description
This book focuses on the state of the art of Monte Carlo methods in radiation physics and particle transport simulation and applications. Special attention is paid to algorithm development for modeling, and the analysis of experiments and measurements in a variety of fields.
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Category :
Languages : en
Pages :
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Book Description
The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.
Author: William L. Dunn
Publisher: Elsevier
ISBN: 0128197455
Category : Science
Languages : en
Pages : 594
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Book Description
Exploring Monte Carlo Methods, Second Edition provides a valuable introduction to the numerical methods that have come to be known as "Monte Carlo." This unique and trusted resource for course use, as well as researcher reference, offers accessible coverage, clear explanations and helpful examples throughout. Building from the basics, the text also includes applications in a variety of fields, such as physics, nuclear engineering, finance and investment, medical modeling and prediction, archaeology, geology and transportation planning. Provides a comprehensive yet concise treatment of Monte Carlo methods Uses the famous "Buffon’s needle problem" as a unifying theme to illustrate the many aspects of Monte Carlo methods Includes numerous exercises and useful appendices on: Certain mathematical functions, Bose Einstein functions, Fermi Dirac functions and Watson functions
Author: Jun Wang
Publisher: Woodhead Publishing
ISBN: 0128181915
Category : Technology & Engineering
Languages : en
Pages : 612
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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: Nicholas Cameron Sly
Publisher:
ISBN:
Category :
Languages : en
Pages : 117
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Book Description
While Monte Carlo simulation has been recognized as a powerful numerical method for use in radiation transport, it has required a mixture of methods development and hardware advancement to meet these expectations in practical applications. In an effort to continue this advancement for uses of Monte Carlo simulation in ever larger capacities, Oak Ridge National Laboratory is developing the Shift hybrid deterministic/Monte Carlo code to be massively-parallel for use on parallel computing systems of all sizes. As part of this development, verification of the Monte Carlo parts of the code is needed to confirm that the current version of the code is operating properly, by matching the results of similar, currently available codes, as well as allowing for testing of the code in the future, to ensure that subsequent code changes and the implementation of new capabilities don’t adversely affect the results. This research starts that verification using some basic reactor criticality benchmarks. The Shift code has been shown to agree within three standard deviations with MCNP and KENO, two of the most widely used Monte Carlo criticality codes. Also investigated was the efficiency of the Shift code as it currently stands, scaling with the number of processors the code is run on as well as the number of particles being simulated. The code was found to scale well, as long as there are enough particles to make the transport take significantly more time than the inter-cycle communication between compute nodes.
Author: Christopher Perfetti
Publisher: Woodhead Publishing
ISBN: 9780128154007
Category :
Languages : en
Pages : 390
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Book Description
Monte Carlo Methods and Codes for Nuclear Engineering Analysis provides a comprehensive survey of the state-of-the-art in radiation transport methods used by Monte Carlo (MC) codes. It then goes on to explore the real-world implementation of these methods in codes used by nuclear and scientists engineers, considering the advantages and disadvantages of the various techniques, design philosophies, and algorithm implementations. After a foreword and introduction giving a brief history of Monte Carlo methods, code systems, and their applications in nuclear science and engineering, subsequent chapters describe the fundamentals of Monte Carlo radiation transport methods by dividing the field into a number of topics or focus areas. The subjects selected include potential geometry and particle tracking, nuclear data, variance reduction, time-dependent analysis and parallel computing. Each chapter presents a comprehensive survey of the state-of-the-art implementations, algorithms, and methodologies used by production-level Monte Carlo codes for the area. A concluding chapter provides a handy summary by briefly listing the methods used by key Monte Carlo codes for each focus area in several tables. This book is an essential guide to Monte Carlo methods and codes for nuclear scientists, engineers and code developers in academia and industry and students studying this topic. discusses and compares the radiation transport methods in real-life Monte Carlo (MC) codes used by nuclear scientists and engineers presents in one convenient volume information previously scattered between conference papers, journal articles, and code manuals, thus allowing MC code users to compare the features and make and educated selections of the codes best meeting their needs chapters begin at a level that is appropriate for readers who are unfamiliar with the field, then go on to address the state-of-the-art
Author: Travis Owings Mock
Publisher:
ISBN:
Category :
Languages : en
Pages :
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ABSTRACT: Monte Carlo stochastic methods of radiation system analysis are among the most popular of computation techniques. Alternatively, deterministic analysis often remains a minority calculation method among nuclear engineers, since large memory requirements inhibit abilities to accomplish 3-D modeling. As a result, deterministic codes are often limited to diffusion type solvers with transport corrections, or limited geometry capabilities such as 1-D or 2-D geometry approximations. However, there are some 3-D deterministic codes with parallel capabilities, used in this work, to abate such issues. The future of radiation systems analysis is undoubtedly evaluation through parallel computation. Large scale heterogeneous systems are especially difficult to model on one machine due to large memory demands, and it becomes advantageous not only to split computational requirements through individual particle interactions, (as is the method for stochastic parallelization), but also to split the geometry of the problem across machines in parallel. In this effort, first presented is a method for multigroup cross section generation for deterministic code use, followed by radiation system analysis performed using parallel 3-D MCNP5 (Monte Carlo) and parallel 3-D PENTRAN (Sn deterministic) such that these two independent calculation methods were used in order to boost confidence of the final results. Two different radiation systems were modeled: an eigenvalue/criticality problem, and a fixed source shielding problem. The work shows that in some systems, stochastic methods are not easily converged, and that tandem use of deterministic calculations provides, at the very least, another means by which the evaluator can increase problem solving efficiency and accuracy. Following this, lessons learned are presented, followed by conclusions and future work.
