Author: Ryanne Ariel Kennedy Publisher: ISBN: Category : Nuclear energy Languages : en Pages : 414
Book Description
Abstract: In support of the nation's nuclear energy industry, the Innovations in Nuclear Infrastructure and Education (INIE) program was established in 2002 by the Department of Energy. Its function is to strengthen university nuclear engineering education programs through improved and original use of university research and training reactors. The Ohio State University (OSU) is part of the INIE consortium consisting of Penn State University, OSU, Purdue University, University of Illinois (Urbana-Champaign), University of Michigan and University of Wisconsin -- Madison. For improving research reactor utilization and to meet objectives consistent with the goals of the INIE program, a full facility model of the OSU Research Reactor (OSURR) was assembled using the PENTRAN (Parallel Environment Neutral particle Transport) 3-D discrete ordinates code (version 9.36b). The focus of this thesis is the creation and benchmark of a full facility model of the OSU Research Reactor using the discrete ordinates transport code PENTRAN and the Monte Carlo code MCNP5. Storing the full phase-space information for an exact geometry model of the OSU Research Reactor using the discrete ordinates code PENTRAN would require a few thousand gigabytes of computer memory. This large memory requirement is a result of the fine spatial meshing essential for modeling the very thin layers of cladding and fuel over the whole core. Such a large model is unrealistically cumbersome even considering the parallel memory and phase-space decomposition capability of PENTRAN. Hence, it was essential to consider some level of homogenization of different material regions including fuel, clad, and/or moderator/coolant in the discrete ordinates model. Several parametric analyses were performed in an attempt to understand the impact of systematic uncertainties in the models that occur as a result of modeling approximations and the homogenization of core regions. These parametric studies were performed using PENTRAN and MCNP and included the analysis of several categories of uncertainty in the research reactor such as fuel impurities and uncertainty in the core's geometry. In addition to analyses of the PENTRAN model, several tests were performed during the construction of the MCNP model. Detailed studies of the control rods and irradiation facilities were performed and compared to experimental data. An irradiation experiment was performed to collect data in three irradiation facilities in the OSURR facility. The thermal flux was calculated in each of these locations using the experimental data and compared directly to the results of the full core models built with MCNP and PENTRAN. In addition to benchmarking the model flux results with this experiment, an eigenvalue comparison was made for three different rod configurations for both codes. Overall, agreement was seen between experimental data, MCNP results, and PENTRAN results. The eigenvalue results from different rod configurations were within the uncertainty that was calculated from parametric analyses of the OSURR core. The flux distributions over the core matched well between MCNP and PENTRAN, and all discrepancies were accounted for by analysis of the homogenization effects and other differences between the models.
Author: Nicholas Cameron Sly Publisher: ISBN: Category : Languages : en Pages : 117
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
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: Alireza Haghighat Publisher: CRC Press ISBN: 042958220X Category : Mathematics Languages : en Pages : 214
Book Description
Fully updated with the latest developments in the eigenvalue Monte Carlo calculations and automatic variance reduction techniques and containing an entirely new chapter on fission matrix and alternative hybrid techniques. This second edition explores the uses of the Monte Carlo method for real-world applications, explaining its concepts and limitations. Featuring illustrative examples, mathematical derivations, computer algorithms, and homework problems, it is an ideal textbook and practical guide for nuclear engineers and scientists looking into the applications of the Monte Carlo method, in addition to students in physics and engineering, and those engaged in the advancement of the Monte Carlo methods. Describes general and particle-transport-specific automated variance reduction techniques Presents Monte Carlo particle transport eigenvalue issues and methodologies to address these issues Presents detailed derivation of existing and advanced formulations and algorithms with real-world examples from the author’s research activities
Author: John S. Hendricks Publisher: Springer Nature ISBN: 3031041291 Category : Science Languages : en Pages : 316
Book Description
This open access book is a pedagogical, examples-based guide to using the Monte Carlo N-Particle (MCNP®) code for nuclear safeguards and non-proliferation applications. The MCNP code, general-purpose software for particle transport simulations, is widely used in the field of nuclear safeguards and non-proliferation for numerous applications including detector design and calibration, and the study of scenarios such as measurement of fresh and spent fuel. This book fills a gap in the existing MCNP software literature by teaching MCNP software usage through detailed examples that were selected based on both student feedback and the real-world experience of the nuclear safeguards group at Los Alamos National Laboratory. MCNP input and output files are explained, and the technical details used in MCNP input file preparation are linked to the MCNP code manual. Benefiting from the authors’ decades of experience in MCNP simulation, this book is essential reading for students, academic researchers, and practitioners whose work in nuclear physics or nuclear engineering is related to non-proliferation or nuclear safeguards. Each chapter comes with downloadable input files for the user to easily reproduce the examples in the text.
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Author: Ryanne Ariel Kennedy
Author: Nicholas Cameron Sly
Author: Magdi M. H. Ragheb
Author: Christopher Perfetti
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Author: OECD Nuclear Energy Agency
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Author: Alireza Haghighat
Author: John S. Hendricks