modelling of nuclear reactor multi physics

Modelling Of Nuclear Reactor Multi Physics
Author: Christophe Demazière
Publisher: Academic Press
Release Date: 2019-11-19
Pages: 368
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

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

Multi Physics Approach To The Modelling And Analysis Of Molten Salt Reactors
Author: Lelio Luzzi
Publisher: Nova Novinka
Release Date: 2012
Pages: 140
ISBN:
Available Language: English, Spanish, And French
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Multi-Physics Modelling (MPM) is an innovative simulation technique that looks very promising for the employment in the field of nuclear engineering as an integrative analysis support in the design development of current and innovative nuclear reactors. This book presents a Multi-Physics Modelling (MPM) approach to the analysis of nuclear reactor core behaviour, developed to study the coupling between neutronics and thermo-hydrodynamics. Reference is made to the Molten Salt Reactor, one of the innovative nuclear systems under development in the framework of the Generation IV International Forum, but the same methodology can be applied to other reactor systems.

Nuclear Reactor Multiphysics Via Bond Graph Formalism
Author: Eugeny Sosnovsky
Publisher:
Release Date: 2014
Pages: 216
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the individual physics separately and exchanges the information at every time step. This approach has limited accuracy, and so recently, there has been an increased interest in fully coupled physics simulation. The bond graph formalism has recently been suggested as a potential paradigm for reactor multiphysics simulation; this work develops the tools necessary to utilize bond graphs for practical transient reactor analysis. The bond graph formalism was first introduced to solve the multiphysics problem in electromechanical systems. Over the years, it has been used in many fields including nuclear engineering, but with limited scope due to its perceived impracticality in large systems. Bond graph formalism works by first representing a discretized multiphysics system using a group of graph elements, connected with bonds; the bonds transport conserved quantities, and the elements impose the relations between them. The representation can be automatically converted into a state derivative vector, which can be integrated in time. In an earlier work, the bond graph formalism was first applied to neutron diffusion, and coupled to diffusive heat transfer in a 1D slab reactor. In this work, methods are developed to represent, using bond graphs, 2D and 3D multigroup neutron diffusion with precursors, nonlinear point kinetics, and basic nearly-incompressible 1D flow for fully coupled reactor simulation. High-performance, matrix-based bond graph processing methods were developed to support the simulation of medium- and large-scale problems. A pressurized water reactor point kinetics, single-channel rod ejection benchmark problem was used to verify the nonlinear point kinetics representation. 2D and 3D boiling water reactor control blade drop problems were also successfully simulated with the matrix-based bond graph processing code. The code demonstrated 3rd-order convergence in time, a very desirable property of fully coupled time integrators.

Low Order Multiphysics Coupling Techniques For Nuclear Reactor Applications
Author: Erik Daniel Walker
Publisher:
Release Date: 2017
Pages: 118
ISBN:
Available Language: English, Spanish, And French
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The accurate modeling and simulation of nuclear reactor designs depends greatly on the ability to couple differing sets of physics together. Current coupling techniques most often use a fixed-point, or Picard, iteration scheme in which each set of physics is solved separately, and the resulting solutions are passed between each solver. In the work presented here, two different coupling techniques are investigated: a Jacobian-Free Newton-Krylov (JFNK) approach and a new methodology called Coarse Mesh Finite Difference Coupling (CMFD-Coupling). What both of these techniques have in common is that they are applied to the low-order CMFD system of equations. This allows for the multiphysics feedback effects to be captured on the low-order system without having to perform a neutron transport solve.The JFNK and CMFD-Coupling approaches were implemented in the MPACT (Michigan Parallel Analysis based on Characteristic Tracing) neutron transport code, which is being developed for the Consortium for Advanced Simulation of Light Water Reactors (CASL). These methods were tested on a wide range of practical reactor physics problems, from a 2D pin cell to a massively parallel 3D full core problem. Initially, JFNK was implemented only as an eigenvalue solver without any feedback enabled. However this led to greatly increased runtimes without any obvious benefit. When multiphysics problems were investigated with both JFNK and CMFD-Coupling, it was concluded that CMFD-Coupling outperformed JFNK in terms of both accuracy and runtime for every problem. When applied to large full core problems with multiple sources of strong feedback enabled, CMFD-Coupling reduced the overall number of transport sweeps required for convergence.

Nuclear Power Plant Design And Analysis Codes
Author: Jun Wang
Publisher: Woodhead Publishing
Release Date: 2020-11-10
Pages: 608
ISBN:
Available Language: English, Spanish, And French
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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.

Verification And Validation Of High Fidelity Multi Physics Simulation Codes For Advanced Nuclear Reactors  Year 2
Author:
Publisher:
Release Date: 2014
Pages:
ISBN:
Available Language: English, Spanish, And French
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Advances Of Computational Fluid Dynamics In Nuclear Reactor Design And Safety Assessment
Author: Jyeshtharaj Joshi
Publisher: Woodhead Publishing
Release Date: 2019-06-15
Pages: 700
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment presents the latest computational fluid dynamic technologies. It includes an evaluation of safety systems for reactors using CFD and their design, the modeling of Severe Accident Phenomena Using CFD, Model Development for Two-phase Flows, and Applications for Sodium and Molten Salt Reactor Designs. Editors Joshi and Nayak have an invaluable wealth of experience that enables them to comment on the development of CFD models, the technologies currently in practice, and the future of CFD in nuclear reactors. Readers will find a thematic discussion on each aspect of CFD applications for the design and safety assessment of Gen II to Gen IV reactor concepts that will help them develop cost reduction strategies for nuclear power plants. Presents a thematic and comprehensive discussion on each aspect of CFD applications for the design and safety assessment of nuclear reactors Provides an historical review of the development of CFD models, discusses state-of-the-art concepts, and takes an applied and analytic look toward the future Includes CFD tools and simulations to advise and guide the reader through enhancing cost effectiveness, safety and performance optimization

High Performance Computing Applications In Numerical Simulation And Edge Computing
Author: Changjun Hu
Publisher: Springer Nature
Release Date: 2019-08-28
Pages: 247
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

This book constitutes the referred proceedings of two workshops held at the 32nd ACM International Conference on Supercomputing, ACM ICS 2018, in Beijing, China, in June 2018. This volume presents the papers that have been accepted for the following workshops: Second International Workshop on High Performance Computing for Advanced Modeling and Simulation in Nuclear Energy and Environmental Science, HPCMS 2018, and First International Workshop on HPC Supported Data Analytics for Edge Computing, HiDEC 2018. The 20 full papers presented during HPCMS 2018 and HiDEC 2018 were carefully reviewed and selected from numerous submissions. The papers reflect such topics as computing methodologies; parallel algorithms; simulation types and techniques; machine learning.

OpenFOAM
Author: J. Miguel Nóbrega
Publisher: Springer
Release Date: 2019-01-24
Pages: 536
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

This book contains selected papers of the 11th OpenFOAM® Workshop that was held in Guimarães, Portugal, June 26 - 30, 2016. The 11th OpenFOAM® Workshop had more than 140 technical/scientific presentations and 30 courses, and was attended by circa 300 individuals, representing 180 institutions and 30 countries, from all continents. The OpenFOAM® Workshop provided a forum for researchers, industrial users, software developers, consultants and academics working with OpenFOAM® technology. The central part of the Workshop was the two-day conference, where presentations and posters on industrial applications and academic research were shown. OpenFOAM® (Open Source Field Operation and Manipulation) is a free, open source computational toolbox that has a larger user base across most areas of engineering and science, from both commercial and academic organizations. As a technology, OpenFOAM® provides an extensive range of features to solve anything from complex fluid flows involving chemical reactions, turbulence and heat transfer, to solid dynamics and electromagnetics, among several others. Additionally, the OpenFOAM technology offers complete freedom to customize and extend its functionalities.

Deterministic Numerical Methods For Unstructured Mesh Neutron Transport Calculation
Author: Liangzhi Cao
Publisher: Woodhead Publishing
Release Date: 2020-09-11
Pages: 288
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

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

BERRU Predictive Modeling
Author: Dan Gabriel Cacuci
Publisher: Springer
Release Date: 2018-12-29
Pages: 451
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

This book addresses the experimental calibration of best-estimate numerical simulation models. The results of measurements and computations are never exact. Therefore, knowing only the nominal values of experimentally measured or computed quantities is insufficient for applications, particularly since the respective experimental and computed nominal values seldom coincide. In the author’s view, the objective of predictive modeling is to extract “best estimate” values for model parameters and predicted results, together with “best estimate” uncertainties for these parameters and results. To achieve this goal, predictive modeling combines imprecisely known experimental and computational data, which calls for reasoning on the basis of incomplete, error-rich, and occasionally discrepant information. The customary methods used for data assimilation combine experimental and computational information by minimizing an a priori, user-chosen, “cost functional” (usually a quadratic functional that represents the weighted errors between measured and computed responses). In contrast to these user-influenced methods, the BERRU (Best Estimate Results with Reduced Uncertainties) Predictive Modeling methodology developed by the author relies on the thermodynamics-based maximum entropy principle to eliminate the need for relying on minimizing user-chosen functionals, thus generalizing the “data adjustment” and/or the “4D-VAR” data assimilation procedures used in the geophysical sciences. The BERRU predictive modeling methodology also provides a “model validation metric” which quantifies the consistency (agreement/disagreement) between measurements and computations. This “model validation metric” (or “consistency indicator”) is constructed from parameter covariance matrices, response covariance matrices (measured and computed), and response sensitivities to model parameters. Traditional methods for computing response sensitivities are hampered by the “curse of dimensionality,” which makes them impractical for applications to large-scale systems that involve many imprecisely known parameters. Reducing the computational effort required for precisely calculating the response sensitivities is paramount, and the comprehensive adjoint sensitivity analysis methodology developed by the author shows great promise in this regard, as shown in this book. After discarding inconsistent data (if any) using the consistency indicator, the BERRU predictive modeling methodology provides best-estimate values for predicted parameters and responses along with best-estimate reduced uncertainties (i.e., smaller predicted standard deviations) for the predicted quantities. Applying the BERRU methodology yields optimal, experimentally validated, “best estimate” predictive modeling tools for designing new technologies and facilities, while also improving on existing ones.

Particle Based Methods
Author: Eugenio Oñate
Publisher: Springer Science & Business Media
Release Date: 2011-02-17
Pages: 268
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

The book contains 11 chapters written by relevant scientists in the field of particle-based methods and their applications in engineering and applied sciences. The chapters cover most particle-based techniques used in practice including the discrete element method, the smooth particle hydrodynamic method and the particle finite element method. The book will be of interest to researchers and engineers interested in the fundamentals of particle-based methods and their applications.

Model Validation And Uncertainty Quantification  Volume 3
Author: H. Sezer Atamturktur
Publisher: Springer Science & Business Media
Release Date: 2014-04-11
Pages: 427
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

This third volume of eight from the IMAC - XXXII Conference, brings together contributions to this important area of research and engineering. The collection presents early findings and case studies on fundamental and applied aspects of Structural Dynamics, including papers on: Linear Systems Substructure Modelling Adaptive Structures Experimental Techniques Analytical Methods Damage Detection Damping of Materials & Members Modal Parameter Identification Modal Testing Methods System Identification Active Control Modal Parameter Estimation Processing Modal Data

Thermo Magnetic Systems For Space Nuclear Reactors
Author: Carlos O. Maidana
Publisher: Springer
Release Date: 2014-09-16
Pages: 53
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

Introduces the reader to engineering magnetohydrodynamics applications and presents a comprehensive guide of how to approach different problems found in this multidisciplinary field. An introduction to engineering magnetohydrodynamics, this brief focuses heavily on the design of thermo-magnetic systems for liquid metals, with emphasis on the design of electromagnetic annular linear induction pumps for space nuclear reactors. Alloy systems that are liquid at room temperature have a high degree of thermal conductivity far superior to ordinary non-metallic liquids. This results in their use for specific heat conducting and dissipation applications. For example, liquid metal-cooled reactors are typically very compact and can be used in space propulsion systems and in fission reactors for planetary exploration. Computer aided engineering (CAE), computational physics and mathematical methods are introduced, as well as manufacturing and testing procedures. An overview on space nuclear systems is also included. This brief is an invaluable tool for design engineers and applied physicists as well as to graduate students in nuclear and mechanical engineering or in applied physics.

NURESIM
Author:
Publisher:
Release Date: 2009
Pages: 47
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

The objective of the NURESIM Integrated Project (FP6) which lasted from 2005 to 2008, with contributions of 18 organizations from 13 European countries, was to start the development of a European Reference Simulation Platform for Nuclear Reactors (so-called NURESIM) and to deliver its first versions. This development has followed a roadmap which is consistent with the SRA (Strategic Research Agenda) of the European SNE-TP (Sustainable Nuclear Energy Technology Platform) and resulted in the delivery of two successive versions during the course of the project. Consistently with the NURESIM roadmap, the development of the platform goes on now in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries. NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results. The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between nonconforming meshes). More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more userfriendly. The platform also provides the informatics environment for testing and comparing different codes. For this purpose, it is essential to permit connection of the codes in a standardized way. The standards are being progressively built, concurrently with the process of developing the platform. The NURESIM platform and the individual models, solvers and codes are being validated through challenging applications corresponding to nuclear reactor situations, and including reference calculations, experiments and plant data. Quantitative deterministic and statistical sensitivity and uncertainty analyses tools are also developed and provided through the platform. A Users' Group of European and non-European countries, including vendors, utilities, TSO, and additional research organizations (beyond the current partners) has also been established in order to enhance the role of the platform in meeting the needs of the nuclear industry, as applied to current and future nuclear reactors. This Final Activity Report summarizes the achievements of the platform in core physics, thermalhydraulics, multi-physics, uncertainties and code integration at the end of the NURESIM project.

Issues In Nuclear Energy Technologies  2013 Edition
Author:
Publisher: ScholarlyEditions
Release Date: 2013-05-01
Pages: 340
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

Issues in Nuclear Energy Technologies / 2013 Edition is a ScholarlyEditions™ book that delivers timely, authoritative, and comprehensive information about Fusion Energy. The editors have built Issues in Nuclear Energy Technologies: 2013 Edition on the vast information databases of ScholarlyNews.™ You can expect the information about Fusion Energy in this book to be deeper than what you can access anywhere else, as well as consistently reliable, authoritative, informed, and relevant. The content of Issues in Nuclear Energy Technologies: 2013 Edition has been produced by the world’s leading scientists, engineers, analysts, research institutions, and companies. All of the content is from peer-reviewed sources, and all of it is written, assembled, and edited by the editors at ScholarlyEditions™ and available exclusively from us. You now have a source you can cite with authority, confidence, and credibility. More information is available at http://www.ScholarlyEditions.com/.

Nuclear Energy Materials And Reactors   Volume I
Author: Yassin A. Hassan
Publisher: EOLSS Publications
Release Date: 2010-09-22
Pages: 428
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

Nuclear Energy Materials and Reactors is a component of Encyclopedia of Energy Sciences, Engineering and Technology Resources in the global Encyclopedia of Life Support Systems (EOLSS), which is an integrated compendium of twenty one Encyclopedias. Nuclear energy is a type of technology involving the controlled use of nuclear fission to release energy for work including propulsion, heat, and the generation of electricity. The theme on Nuclear Energy Materials and Reactors discusses: Fundamentals of Nuclear Energy; Nuclear Physics; Nuclear Interactions; Nuclear Reactor Theory; Nuclear Reactor Design; Nuclear Reactor Kinetics; Reactivity Changes; Nuclear Power Plants; Pressurized Water Reactors; Boiling Water Reactors; Pressurized Heavy Water Reactors; Heavy Water Light Water Reactors; Advanced Gas Cooled Reactors; Light Water Graphite Reactors; High Temperature Gas Cooled Reactors; Pebble Bed Modular Reactor; Radioactive Wastes, Origins, Classification and Management; Nuclear Reactor Overview and Reactor Cycles; The Nuclear Reactor Closed Cycle; Safety of Boiling Water Reactors; Supercritical Water-Cooled Nuclear Reactors: Review and Status; The Gas-Turbine Modular Helium Reactor; Application of Risk Assessment to Nuclear Power Plants; Production and Recycling Resources for Nuclear Fission. These two volumes are aimed at the following five major target audiences: University and College students Educators, Professional practitioners, Research personnel and Policy analysts, managers, and decision makers.

Assessing The Reliability Of Complex Models
Author: National Research Council
Publisher: National Academies Press
Release Date: 2012-07-26
Pages: 144
ISBN:
Available Language: English, Spanish, And French
EBOOK SYNOPSIS:

Advances in computing hardware and algorithms have dramatically improved the ability to simulate complex processes computationally. Today's simulation capabilities offer the prospect of addressing questions that in the past could be addressed only by resource-intensive experimentation, if at all. Assessing the Reliability of Complex Models recognizes the ubiquity of uncertainty in computational estimates of reality and the necessity for its quantification. As computational science and engineering have matured, the process of quantifying or bounding uncertainties in a computational estimate of a physical quality of interest has evolved into a small set of interdependent tasks: verification, validation, and uncertainty of quantification (VVUQ). In recognition of the increasing importance of computational simulation and the increasing need to assess uncertainties in computational results, the National Research Council was asked to study the mathematical foundations of VVUQ and to recommend steps that will ultimately lead to improved processes. Assessing the Reliability of Complex Models discusses changes in education of professionals and dissemination of information that should enhance the ability of future VVUQ practitioners to improve and properly apply VVUQ methodologies to difficult problems, enhance the ability of VVUQ customers to understand VVUQ results and use them to make informed decisions, and enhance the ability of all VVUQ stakeholders to communicate with each other. This report is an essential resource for all decision and policy makers in the field, students, stakeholders, UQ experts, and VVUQ educators and practitioners.

REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS
Author:
Publisher:
Release Date: 2010
Pages:
ISBN:
Available Language: English, Spanish, And French
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Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The core level results may then be compared to the radiochemistry results from the dissolved fuel samples and a decision whether further more in-depth modeling should be performed. The SRNL is in the process of analyzing multiple research reactor fuels to determine the best means to provide forensic data for attribution and assess codes and modeling methods for attribution. As several fuel samples are analyzed, this work will allow improved SNM forensics of spent research reactor fuel. This will enable the establishment of a research reactor fuel database of SNM materials, and allow an attempt of an inverse analysis if research reactor material is diverted and seized.

Solid State And Nuclear Physics
Author:
Publisher: Krishna Prakashan Media
Release Date:
Pages:
ISBN:
Available Language: English, Spanish, And French
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