Development of a Coupled Neutronics/thermal-hydraulics/fuel Thermo-mechanics Multiphysics Tool for Best-estimate PWR Core Simulations PDF Download

Author: Joaquín Rubén Basualdo Perelló
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 13

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This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the "VERAIn" common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal-hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal-hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.

Author: Hj Matzke
Publisher: North Holland
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 352

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This volume brings together 47 papers from scientists involved in the fabrication of new nuclear fuels, in basic research of nuclear materials, their application and technology as well as in computer codes and modelling of fuel behaviour. The main emphasis is on progress in the development of non-oxide fuels besides reporting advances in the more conventional oxide fuels. The two currently performed large reactor safety programmes CORA and PHEBUS-FP are described in invited lectures. The contributions review basic property measurements, as well as the present state of fuel performance modelling. The performance of today's nuclear fuel, hence UO2, at high burnup is also reviewed with particular emphasis on the recently observed phenomenon of grain subdivision in the cold part of the oxide fuel at high burnup, the so-called rim effect. Similar phenomena can be simulated by ion implantation in order to better elucidate the underlying mechanism and reviews on high resolution electron microscopy provide further information.

Author: Thomas J. Santner
Publisher: Springer
ISBN: 1493988476
Category : Mathematics
Languages : en
Pages : 436

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This book describes methods for designing and analyzing experiments that are conducted using a computer code, a computer experiment, and, when possible, a physical experiment. Computer experiments continue to increase in popularity as surrogates for and adjuncts to physical experiments. Since the publication of the first edition, there have been many methodological advances and software developments to implement these new methodologies. The computer experiments literature has emphasized the construction of algorithms for various data analysis tasks (design construction, prediction, sensitivity analysis, calibration among others), and the development of web-based repositories of designs for immediate application. While it is written at a level that is accessible to readers with Masters-level training in Statistics, the book is written in sufficient detail to be useful for practitioners and researchers. New to this revised and expanded edition: • An expanded presentation of basic material on computer experiments and Gaussian processes with additional simulations and examples • A new comparison of plug-in prediction methodologies for real-valued simulator output • An enlarged discussion of space-filling designs including Latin Hypercube designs (LHDs), near-orthogonal designs, and nonrectangular regions • A chapter length description of process-based designs for optimization, to improve good overall fit, quantile estimation, and Pareto optimization • A new chapter describing graphical and numerical sensitivity analysis tools • Substantial new material on calibration-based prediction and inference for calibration parameters • Lists of software that can be used to fit models discussed in the book to aid practitioners

Author: Thomas James Dolan
Publisher: Elsevier
ISBN: 0323993567
Category : Technology & Engineering
Languages : en
Pages : 1068

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Molten Salt Reactors and Thorium Energy, Second Edition is a fully updated comprehensive reference on the latest advances in MSR research and technology. Building on the successful first edition, Tom Dolan and the team of experts have fully updated the content to reflect the impressive advances from the last 5 years, ensuring this book continues to be the go-to reference on the topic. This new edition covers progress made in MSR design, details innovative experiments, and includes molten salt data, corrosion studies and deployment plans. The successful case studies section of the first edition have been removed, expanded, and fully updated, and are now published in a companion title called Global Case Studies on Molten Salt Reactors. Readers will gain a deep understanding of the advantages and challenges of MSR development and thorium fuel use, as well as step-by-step guidance on the latest in MSR reactor design. Each chapter provides a clear introduction, covers technical issues and includes examples and conclusions, while promoting the sustainability benefits throughout. A fully updated comprehensive handbook on Molten Salt Reactors and Thorium Energy, written by a team of global experts Covers MSR applications, technical issues, reactor types and reactor designs Includes 3 brand new chapters which reflect the latest advances in research and technology since the first edition published Presents case studies on molten salt reactors which aid in the transition to net zero by providing abundant clean, safe energy to complement wind and solar powe

Author:
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ISBN:
Category :
Languages : en
Pages :

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The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.

Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201165213
Category :
Languages : en
Pages : 312

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Flexible operation and related power changes can have a direct impact on fuel integrity through pellet-cladding interaction/stress corrosion cracking (PCI/SCC) phenomena, which could lead to fuel failures in certain conditions.

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:
Publisher:
ISBN:
Category :
Languages : en
Pages : 14

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A novel object-oriented modular mapping methodology for externally coupled neutronics-thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. In addition, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers.

Author: International Atomic Energy Agency
Publisher: International Atomic Energy Agency
ISBN: 9789201396105
Category : Political Science
Languages : en
Pages : 169

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"This publication is based on the experience of an IAEA coordinated research project on control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments. Presented in this publication are the benchmark analyses of the natural circulation test performed before the definite shutdown of the reactor. The experimental data gathered during these tests represent a unique resource to carry out validation analyses and code-to-code comparisons. The benchmark analyses allowed participants to investigate and verify several system and safety codes currently used in the analyses of liquid metal thermal hydraulics phenomena in sodium fast reactors."--Publisher's description.