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Coupling Thermal Hydraulics With Neutronics For Pebble Bed High Temperature Reactor Calculations
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Book Synopsis Coupling Thermal Hydraulics with Neutronics for Pebble-bed High Temperature Reactor Calculations by : E. C. Verkerk
Download or read book Coupling Thermal Hydraulics with Neutronics for Pebble-bed High Temperature Reactor Calculations written by E. C. Verkerk and published by . This book was released on 1998 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Coupled Neutronics and Thermal-hydraulics Modeling for Pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR) by : Xin Wang
Download or read book Coupled Neutronics and Thermal-hydraulics Modeling for Pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR) written by Xin Wang and published by . This book was released on 2018 with total page 136 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Tightly Coupled Neutronics and Thermal-hydraulics Using Open-source Software by :
Download or read book Tightly Coupled Neutronics and Thermal-hydraulics Using Open-source Software written by and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book THR-TH written by and published by . This book was released on 1984 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.
Book Synopsis Neutronic and Thermal-hydraulic Coupling Study on High Temperature Gas-cooled Reactor by : Cheng-Kai Tai
Download or read book Neutronic and Thermal-hydraulic Coupling Study on High Temperature Gas-cooled Reactor written by Cheng-Kai Tai and published by . This book was released on 2017 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model by :
Download or read book Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle the larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.
Book Synopsis Coupled High Fidelity Thermal Hydraulics and Neutronics for Reactor Safety Simulations by :
Download or read book Coupled High Fidelity Thermal Hydraulics and Neutronics for Reactor Safety Simulations written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un'SCRAM'ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.
Book Synopsis Coupled Neutronics, Thermal Hydraulics Analysis of a High Performance Light Water Reactor Fuel Assembly by : Christine Lylin Waata
Download or read book Coupled Neutronics, Thermal Hydraulics Analysis of a High Performance Light Water Reactor Fuel Assembly written by Christine Lylin Waata and published by . This book was released on 2006 with total page 89 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nuclear Reactor Thermal Hydraulics and Other Applications by : Donna Guillen
Download or read book Nuclear Reactor Thermal Hydraulics and Other Applications written by Donna Guillen and published by BoD – Books on Demand. This book was released on 2013-02-13 with total page 204 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book includes contributions from researchers around the world on numerical developments and applications to predict fluid flow and heat transfer, with an emphasis on thermal hydraulics computational fluid dynamics. Our ability to simulate larger problems with greater fidelity has vastly expanded over the past decade. The collection of material presented in this book augments the ever-increasing body of knowledge concerning the important topic of thermal hydraulics. Featured topics include coolant channel analysis, thermal hydraulic transport and mixing, as well as hydrodynamics and heat transfer processes. The contents of this book will interest researchers, scientists, engineers and graduate students.
Book Synopsis Optimized Core Design and Fuel Management of a Pebble-bed Type Nuclear Reactor by : Brian Boer
Download or read book Optimized Core Design and Fuel Management of a Pebble-bed Type Nuclear Reactor written by Brian Boer and published by . This book was released on 2008 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. As one of the fourth generation nuclear reactors, the VHTR is characterized by high plant efficiency and a high fuel discharge burn-up level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core.
Book Synopsis The Development of a Thermal Hydraulic Feedback Mechanism with a Quasi-fixed Point Iteration Scheme for Control Rod Position Modeling for the TRIGSIMS-TH Application by : Veronica Karriem
Download or read book The Development of a Thermal Hydraulic Feedback Mechanism with a Quasi-fixed Point Iteration Scheme for Control Rod Position Modeling for the TRIGSIMS-TH Application written by Veronica Karriem and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations.The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings.The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data.The PSBR is unique in many ways and there are no off-the-shelf codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings. The CTF code can be used as a thermal hydraulic stand-alone modeling code. The TRIGSIMS-TH code generates and expands channel thermal hydraulic input model that is capable of analyzing the flow in and around the core construct. The tool can be used to analyze future changes such as the safety analysis of the D2O tank changes.The TRIGSIMS-TH code system is an automated tool. Using a generalized input, -it will generate all the needed code-specific input files for the various applications.
Book Synopsis Multiphase Flow and Heat Transfer in Pebble Bed Reactor Core by : Shengyao Jiang
Download or read book Multiphase Flow and Heat Transfer in Pebble Bed Reactor Core written by Shengyao Jiang and published by Springer Nature. This book was released on 2020-11-19 with total page 510 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book introduces readers to gas flows and heat transfer in pebble bed reactor cores. It addresses fundamental issues regarding experimental and modeling methods for complex multiphase systems, as well as relevant applications and recent research advances. The numerical methods and experimental measurements/techniques used to solve pebble flows, as well as the content on radiation modeling for high-temperature pebble beds, will be of particular interest. This book is intended for a broad readership, including researchers and practitioners, and is sure to become a key reference resource for students and professionals alike.
Book Synopsis Nuclear Thermal Hydraulics by : Hajime Akimoto
Download or read book Nuclear Thermal Hydraulics written by Hajime Akimoto and published by Springer. This book was released on 2016-11-10 with total page 464 pages. Available in PDF, EPUB and Kindle. Book excerpt: This volume provides fundamentals of nuclear thermal–hydraulics for reactor design and safety assessment. It also describes the basis for assessing cooling performance of nuclear reactors under accidental conditions. The descriptions in this book are virtually self-contained, beyond the assumption that readers are familiar with the introductory levels of nuclear engineering. This book helps readers understand the processes for nuclear reactor plant design and the most important factors in nuclear thermal-hydraulics.
Book Synopsis Coupled Neutronic Thermal Fluid Dynamic Modelling of a Very High Temperature Reactor by :
Download or read book Coupled Neutronic Thermal Fluid Dynamic Modelling of a Very High Temperature Reactor written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Neutronic and Thermal-hydraulics Coupling for Simulations of the TRIGA Mark II Reactor by : Romain Henry
Download or read book Neutronic and Thermal-hydraulics Coupling for Simulations of the TRIGA Mark II Reactor written by Romain Henry and published by . This book was released on 2017 with total page 126 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Handbook of Generation IV Nuclear Reactors by : Igor Pioro
Download or read book Handbook of Generation IV Nuclear Reactors written by Igor Pioro and published by Woodhead Publishing. This book was released on 2022-12-07 with total page 1112 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. The book teaches the reader about available technologies, future prospects and the feasibility of each concept presented, equipping them users with a strong skillset which they can apply to their own work and research. Provides a fully updated, revised and comprehensive handbook dedicated entirely to generation IV nuclear reactors Includes new trends and developments since the first publication, as well as brand new case studies and appendices Covers the latest research, developments and design information surrounding generation IV nuclear reactors
Book Synopsis Proceedings of the International Conference on the Physics of Nuclear Science and Technology by :
Download or read book Proceedings of the International Conference on the Physics of Nuclear Science and Technology written by and published by . This book was released on 1998 with total page 852 pages. Available in PDF, EPUB and Kindle. Book excerpt:
