Download or read book Advanced Methodology to Simulate Boiling Water Reactor Transient Using Coupled Thermal hydraulic neutron kinetic Codes written by and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Boiling Water Reactor Transient Instability Studies of Ringhals 1 Reactor Using TRACE Coupled with PARCS written by Robert Allen Walls and published by . This book was released on 2009 with total page 95 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book Light water reactor Coupled Neutronic and Thermal hydraulic Codes written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented.

Download or read book Nuclear Power Plant Design and Analysis Codes written by Jun Wang and published by Woodhead Publishing. This book was released on 2020-11-10 with total page 612 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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

Download or read book Stability Analysis of the Boiling Water Reactor written by Rui Hu (Ph. D.) and published by . This book was released on 2010 with total page 348 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book RAMONA 4B a Computer Code with Three dimensional Neutron Kinetics for BWR and SBWR System Transient Models and Correlations written by and published by . This book was released on 1998 with total page 440 pages. Available in PDF, EPUB and Kindle. Book excerpt: This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the codes̀ capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

Download or read book Effective Spatial Mapping for Coupled Code Analysis of Thermal hydraulics neutron kinetics of Boiling Water Reactors written by and published by . This book was released on 2013 with total page 148 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Multidimensional Reactor Kinetics Modeling written by and published by . This book was released on 1996 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: There is general agreement that for many light water reactor transient calculations, it is-necessary to use a multidimensional neutron kinetics model coupled to a thermal-hydraulics model for satisfactory results. These calculations are needed for a variety of applications for licensing safety analysis, probabilistic risk assessment (PRA), operational support, and training. The latter three applications have always required best-estimate models, but in the past applications for licensing could be satisfied with relatively simple models. By using more sophisticated best-estimate models, the consequences of these calculations are better understood, and the potential for gaining relief from restrictive operating limits increases. Hence, for all of the aforementioned applications, it is important to have the ability to do best-estimate calculations with multidimensional neutron kinetics models. coupled to sophisticated thermal-hydraulic models. Specifically, this paper reviews the status of multidimensional neutron kinetics modeling which would be used in conjunction with thermal-hydraulic models to do core dynamics calculations, either coupled to a complete NSSS representation or in isolation. In addition, the paper makes recommendations as to what should be the state-of-the-art for the next ten years. The review is an update to a previous review of the status as of ten years ago. The general requirements for a core dynamics code and the modeling available for such a code, discussed in that review, are still applicable. The emphasis in the current review is on the neutron kinetics assuming that the necessary thermal-hydraulic capability exists. In addition to discussing the basic neutron kinetics, discussion is given of related modeling (other than thermal- hydraulics). The capabilities and limitations of current computer codes are presented to understand the state-of-the-art and to help clarify the future direction of model development in this area.

Download or read book Coupled 3D Reactor Kinetics and Thermal hydraulic Code Development Activities at the U S Nuclear Regulatory Commission written by and published by . This book was released on 1999 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: The USNRC version of the 3D neutron kinetics code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the USNRC thermal-hydraulic (T/H) codes RELAP5 and the consolidated TRAC (merger of TRAC-BF1 and TRAC-PF1). These coupled codes may be used to audit license safety analysis submittals where 3D spatial kinetics and thermal-hydraulic effects are important. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the T/H and neutronic codes function independently and utilize the Parallel Virtual Machine software to communicate with each other through code specific Data Mapping Routines, and a General Interface. RELAP5/PARCS validation results are presented for two NEACRP rod ejection benchmark problems. The validation of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

Download or read book Linear and Non linear Stability Analysis in Boiling Water Reactors written by Alfonso Prieto Guerrero and published by Woodhead Publishing. This book was released on 2018-10-15 with total page 474 pages. Available in PDF, EPUB and Kindle. Book excerpt: Linear and Non-Linear Stability Analysis in Boiling Water Reactors: The Design of Real-Time Stability Monitors presents a thorough analysis of the most innovative BWR reactors and stability phenomena in one accessible resource. The book presents a summary of existing literature on BWRs to give early career engineers and researchers a solid background in the field, as well as the latest research on stability phenomena (propagation phenomena in BWRs), nuclear power monitors, and advanced computer systems used to for the prediction of stability. It also emphasizes the importance of BWR technology and embedded neutron monitoring systems (APRMs and LPRMs), and introduces non-linear stability parameters that can be used for the onset detection of instabilities in BWRs. Additionally, the book details the scope, advantages, and disadvantages of multiple advanced linear and non linear signal processing methods, and includes analytical case studies of existing plants. This combination makes Linear and Non-Linear Stability Analysis in Boiling Water Reactors a valuable resource for nuclear engineering students focusing on linear and non-linear analysis, as well as for those working and researching in a nuclear power capacity looking to implement stability methods and estimate decay ratios using non-linear techniques. - Explores the nuclear stability of Boiling Water Reactors based on linear and non-linear models - Evaluates linear signal processing methods such as autoregressive models, Fourier-based methods, and wavelets to calculate decay ratios - Proposes novel non-linear signal analysis techniques linked to non-linear stability indicators - Includes case studies of various existing nuclear power plants as well as mathematical models and simulations

Download or read book Semi implicit Thermal hydraulic Coupling of Advanced Subchannel and System Codes for Pressurized Water Reactor Transient Applications written by Maria Desamparados Soler-Martinez and published by . This book was released on 2011 with total page 114 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Neutronics Methods for Transient and Safety Analysis of Fast Reactors written by Marchetti, Marco and published by KIT Scientific Publishing. This book was released on 2017-03-02 with total page 164 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Handbook on Thermal Hydraulics in Water Cooled Nuclear Reactors written by Francesco D'Auria and published by Elsevier. This book was released on 2024-07-29 with total page 932 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 1, Foundations and Principles 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

Download or read book A Three dimensional Transient Neutronics Routine for the TRAC PF1 Reactor Thermal Hydraulic Computer Code written by and published by . This book was released on 1990 with total page 290 pages. Available in PDF, EPUB and Kindle. Book excerpt: No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.

Download or read book Steady State Coupled Neutronic and Thermal Hydraulic Analysis of Boiling Water Reactor Systems with the DYNOBOSS Code written by Bryce S. Knight and published by . This book was released on 1997 with total page 91 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book STREAC written by E. P. Hunger and published by . This book was released on 1961 with total page 98 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Coupling Between Monte Carlo Neutron Transport and Thermal hydraulics for the Simulation of Transients Due to Reactivity Insertions written by Margaux Faucher and published by . This book was released on 2019 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: One of the main issues for the study of a reactor behaviour is to model the propagation of the neutrons, described by the Boltzmann transport equation, in the presence of multi-physics phenomena, such as the coupling between neutron transport, thermal-hydraulics and thermomecanics. Thanks to the growing computer power, it is now feasible to apply Monte Carlo methods to the solution of non-stationary transport problems in reactor physics, which play an instrumental role in producing reference numerical solutions for the analysis of transients occurring during normal and accidental behaviour.The goal of this Ph. D. thesis is to develop, verify and test a coupling scheme between the Monte Carlo code TRIPOLI-4 and thermal-hydraulics, so as to provide a reference tool for the simulation of reactivity-induced transients in PWRs.We have first tested the kinetic capabilities of TRIPOLI-4 (i.e., time dependent without thermal-hydraulics feedback), evaluating the different existing methods and implementing new techniques. Then, we have developed a multi-physics interface for TRIPOLI-4, and more specifically a coupling scheme between TRIPOLI-4 and the thermal-hydraulics sub-channel code SUBCHANFLOW. Finally, we have performed a preliminary analysis of the stability of the coupling scheme. Indeed, due to the stochastic nature of the outputs produced by TRIPOLI-4, uncertainties are inherent to our coupling scheme and propagate along the coupling iterations. Moreover, thermal-hydraulics equations are non linear, so the prediction of the propagation of the uncertainties is not straightforward.