Boiling Water Reactor Simulations, Models, and Benchmarking Using the Thermal Hydraulics Sub-channel Code CTF. PDF Download
Are you looking for read ebook online? Search for your book and save it on your Kindle device, PC, phones or tablets. Download Boiling Water Reactor Simulations, Models, and Benchmarking Using the Thermal Hydraulics Sub-channel Code CTF. PDF full book. Access full book title Boiling Water Reactor Simulations, Models, and Benchmarking Using the Thermal Hydraulics Sub-channel Code CTF. by Christopher Gosdin. Download full books in PDF and EPUB format.
Author: Christopher Gosdin
Publisher:
ISBN:
Category :
Languages : en
Pages :
Get Book Here
Book Description
CTF, the version of the thermal-hydraulic sub-channel code COBRA-TF being jointly developed and maintained by Pennsylvania State University (PSU) and Oak Ridge National Laboratory (ORNL) for applications in the U.S. Department of Energy (DOE) supported Consortium for Advanced Simulation of Light Water Reactors (CASL) project, uses a two-fluid, three-field representation of two-phase flow, which makes the code capable of modeling two-phase flow in Boiling Water Reactors (BWR) during nominal operating conditions. The sub-channel code CTF is used for Pressurized Water Reactors (PWR) for best-estimate evaluations of the nuclear reactor safety margins; however, due to its capabilities, CTF is powerful and valuable computational tool for modeling BWRs. CTF has been subjected to a strict verification procedure, by addressing the mathematical accuracy of the numerical solutions on multiple stages. The code was then validated using numerous of experimental databases, including the U.S. Nuclear Regulatory Commission (NRC) / Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA-OECD) Boiling Water Reactor Full Bundle Tests (BFBT) Benchmark. The BFBT benchmark contains a large amount of test cases representative of BWRs steady-state and off-nominal operating conditions, which makes it one of the most widely used benchmark for validating BWR modeling tools. Two of the main experimental tests involve critical power tests and void distribution tests. Specific experimental cases were chosen and simulated using CTF. Statistical studies were carried out on the void distribution cases to evaluate the code modeling uncertainties. This thesis also focuses on application of CTF to mini- and whole-core BWR calculations on a pin-cell resolved level; as well as on demonstrating that CTF can properly model bypass flow in BWR cores. To increase the confidence in the CTF's BWR modeling capabilities, extensive simulations have been performed using the international NEA-OECD / US NNRC Oskarshamn-2 benchmark, including modeling of a single and 2x2 assemblies on a pin-by-pin level, and a full core model on an assembly level. Each model is varied, with an increasing amount of detail. The results demonstrate that CTF is capable of modeling basic and complex BWR core configurations and operating conditions. Using the three Oskarshamn-2 simulations, CTF's capabilities of modeling BWRs was further verified.
Author: Christopher Gosdin
Publisher:
ISBN:
Category :
Languages : en
Pages :
Get Book Here
Book Description
CTF, the version of the thermal-hydraulic sub-channel code COBRA-TF being jointly developed and maintained by Pennsylvania State University (PSU) and Oak Ridge National Laboratory (ORNL) for applications in the U.S. Department of Energy (DOE) supported Consortium for Advanced Simulation of Light Water Reactors (CASL) project, uses a two-fluid, three-field representation of two-phase flow, which makes the code capable of modeling two-phase flow in Boiling Water Reactors (BWR) during nominal operating conditions. The sub-channel code CTF is used for Pressurized Water Reactors (PWR) for best-estimate evaluations of the nuclear reactor safety margins; however, due to its capabilities, CTF is powerful and valuable computational tool for modeling BWRs. CTF has been subjected to a strict verification procedure, by addressing the mathematical accuracy of the numerical solutions on multiple stages. The code was then validated using numerous of experimental databases, including the U.S. Nuclear Regulatory Commission (NRC) / Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA-OECD) Boiling Water Reactor Full Bundle Tests (BFBT) Benchmark. The BFBT benchmark contains a large amount of test cases representative of BWRs steady-state and off-nominal operating conditions, which makes it one of the most widely used benchmark for validating BWR modeling tools. Two of the main experimental tests involve critical power tests and void distribution tests. Specific experimental cases were chosen and simulated using CTF. Statistical studies were carried out on the void distribution cases to evaluate the code modeling uncertainties. This thesis also focuses on application of CTF to mini- and whole-core BWR calculations on a pin-cell resolved level; as well as on demonstrating that CTF can properly model bypass flow in BWR cores. To increase the confidence in the CTF's BWR modeling capabilities, extensive simulations have been performed using the international NEA-OECD / US NNRC Oskarshamn-2 benchmark, including modeling of a single and 2x2 assemblies on a pin-by-pin level, and a full core model on an assembly level. Each model is varied, with an increasing amount of detail. The results demonstrate that CTF is capable of modeling basic and complex BWR core configurations and operating conditions. Using the three Oskarshamn-2 simulations, CTF's capabilities of modeling BWRs was further verified.
Author: Rui Hu (Ph. D.)
Publisher:
ISBN:
Category :
Languages : en
Pages : 348
Get Book Here
Book Description
Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been considerable concern about the effects of such oscillations when coupled with neutronic feedback. The current trend of increasing reactor power density and relying more extensively on natural circulation for core cooling may have consequences for the stability characteristics of new BWR designs. This work addresses a wide range of issues associated with the BWR stability: 1) flashing-induced instability and natural circulation BWR startup; 2) stability of the BWRs with advanced designs involving high power :densities; 3) modeling assumptions in stability analysis methods; and 4) the fuel clad performance during power and flow oscillations. To capture the effect of flashing on density wave oscillations during low pressure startup conditions, a code named FISTAB has been developed in the frequency domain. The code is based on a single channel thermal-hydraulic model of the balance of the water/steam circulation loop, and incorporates the pressure dependent water/steam thermodynamic properties, from which the evaporation due to flashing is captured. The functionality of the FISTAB code is confirmed by testing the experimental results at SIRIUS-N facility. Both stationary and perturbation results agree well with the experimental results. The proposed ESBWR start-up procedure under natural convection conditions has been examined by the FISTAB code. It is confirmed that the examined operating points along the ESBWR start-up trajectory from TRACG simulation will be stable. To avoid the instability resulting from the transition from single-phase natural circulation to two-phase circulation, a simple criterion is proposed for the natural convection BWR start-up when the steam dome pressure is still low. Using the frequency domain code STAB developed at MIT, stability analyses of some proposed advanced BWRs have been conducted, including the high power density BWR core designs using the Large Assembly with Small Pins (LASP) or Cross Shape Twisted (CST) fuel designs developed at MIT, and the Hitachi's RBWR cores utilizing a hard neutron spectrum and even higher power density cores. The STAB code is the predecessor of the FISTAB code, and thermodynamic properties of the coolant are only dependent on system pressure in STAB. It is concluded that good stability performance of the LASP core and the CST core can be maintained at nominal conditions, even though they have 20% higher reactor thermal power than the reference core. Power uprate does not seem to have significant effects on thermal-hydraulic stability performance when the power-to-flow ratio is maintained. Also, both the RBWR-AC and RBWR-TB2 designs are found viable from a stability performance point of view, even though the core exit qualities are almost 3 times those of a traditional BWR. The stability of the RBWRs is enhanced through the fast transient response of the shorter core, more flat power and power-to-flow ratio distributions, less negative void feedback coefficient, and the core inlet orifice design. To examine the capability of coupled 3D thermal-hydraulics and neutronics codes for stability analysis, USNRC's latest system analysis code, TRACE, is chosen in this work. Its validation for stability analysis and comparison with the frequency domain approach, have been performed against the Ringhals 1 stability tests. Comprehensive assessment of modeling choices on TRACE stability analysis has been made, including effects of timespatial discretization, numerical schemes, thermal-hydraulic channel grouping, neutronics modeling, and control system modeling. The predictions from both the TRACE and STAB codes are found in reasonably good agreement with the Ringhals 1 test results. The biases for the predicted global decay ratio are about 0.07 in TRACE results, and -0.04 in STAB results. However, the standard deviations of decay ratios are both large, around 0.1, indicating large uncertainties in both analyses. Although the TRACE code uses more sophisticated neutronic and thermal hydraulic models, the modeling uncertainty is not less than that of the STAB code. The benchmark results of both codes for the Ringhals stability test are at the same level of accuracy. The fuel cladding integrity during power oscillations without reactor scram is examined by using the FRAPTRAN code, with consideration of both the stress-strain criterion and thermal fatigue. Under the assumed power oscillation conditions for high burn-up fuel, the cladding can satisfy the stress-strain criteria in the ASME Code. Also, the equivalent alternating stress is below the fatigue threshold stress, thus the fatigue limit is not violated. It can be concluded that under a large amount of the undamped power oscillation cycles, the cladding would not fail, and the fuel integrity is not compromised.
Author: R. M. Hiatt
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 196
Get Book Here
Book Description
REPP, a digital computer method for designing pressure water and boiling water reactor cores within specified heat transfer and fuel centerline temperature limits is presented. The method incorporates the Westinghouse W-2 and W-3 empirical correlations and a theoretical hot channel model to predict burnout conditions in a rod bundle. Two geometries are considered; rods in a triangular array and rods in a square lattice. The heat transfer problem solved is a one-dimensional analysis. Pressure drop is considered for four types of fuel-pin spacers. Variable heat generation rate through the fuel-pin and sintering in low density fuels are also included.
Author: Francesco D'Auria
Publisher: Elsevier
ISBN: 032385611X
Category : Technology & Engineering
Languages : en
Pages : 1012
Get Book Here
Book Description
Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 2, Modelling includes all new chapters which delve deeper into the topic, adding context and practical examples to help readers apply learnings to their own setting. Topics covered include experimental thermal-hydraulics and instrumentation, numerics, scaling and containment in thermal-hydraulics, as well as a title dedicated to good practices in verification and validation. This book will be a valuable reference for graduate and undergraduate students of nuclear or thermal engineering, as well as researchers in nuclear thermal-hydraulics and reactor technology, engineers working in simulation and modeling of nuclear reactors, and more. In addition, nuclear operators, code developers and safety engineers will also benefit from the practical guidance provided. Presents a comprehensive analysis on the connection between nuclear power and thermal hydraulics Includes end-of-chapter questions, quizzes and exercises to confirm understanding and provides solutions in an appendix Covers applicable nuclear reactor safety considerations and design technology throughout
Author: Jun Wang
Publisher: Woodhead Publishing
ISBN: 0128181915
Category : Technology & Engineering
Languages : en
Pages : 612
Get Book Here
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: Robert Allen Walls
Publisher:
ISBN:
Category :
Languages : en
Pages : 95
Get Book Here
Book Description
Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered from simulations aids operators and designers in planning for overall plant operation and most importantly, safety. The United States (US) Nuclear Regulatory Commission (NRC) in researching reactor plant safety utilizes several computer codes, or models. The two codes used for work on this thesis are TRACE and PARCS. TRACE (TRAC RELAP5 Advanced Computational Engine) is a thermal-hydraulic code that models the coolant system under numerous variables in operating conditions. Coolant flow is especially important and the ability to model two-phase flow is essential in modeling boiling water reactors. Two-phase flow modeling is integral as it models the vast differences in flow from the bottom of the core to the top at the steam separators. TRACE has the ability to reproduce these essential parameters. PARCS (Purdue Advanced Reactor Core Simulator) is a multi-dimensional reactor kinetics code. TRACE coupled with PARCS has the computing power to provide accurate coupled power and flow distributions under various reactor transients or casualties. TRACE/PARCS was previously validated for use with Pressurized Water Reactor (PWR) transient analysis using the OECD/NEA Main Steam Line Break Benchmark. This thesis focuses on the evaluation of Boiling Water Reactor (BWR) transient analysis, mainly the NEA Ringhals 1 Stability Benchmark from 1996. This benchmark performed a series of tests on the Ringhals 1 reactor during the beginning of cycles 14, 15, 16, and 17. Three techniques for initiating instabilities (pressure perturbation, control rod perturbation, and simulated noise) were performed on each test point during each cycle. The steady state data as well as the transient results predicted by TRACE/PARCS reasonably agree with the measured data from the NEA Ringhals 1 Stability Benchmark.
Author: Richard T. Lahey
Publisher:
ISBN:
Category : Science
Languages : en
Pages : 658
Get Book Here
Book Description
This edition of the classic monograph gives a comprehensive overview of the thermal-hydraulic technology underlying the design, operation, and safety assessment of boiling water reactors. In addition, new material on pressure suppression containment technology is presented.
Author: E. P. Hunger
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 98
Get Book Here
Book Description
Author: Veronica Karriem
Publisher:
ISBN:
Category :
Languages : en
Pages :
Get Book Here
Book Description
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.
Author: J. M. Healzer
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 76
Get Book Here
Book Description
