Nonlinear Dynamic Analysis of Nuclear Reactor Primary Coolant Systems
Nonlinear Dynamic Analysis of Nuclear Reactor Primary Coolant Systems

Author:

Publisher:

Published: 1979

Total Pages:

ISBN-13:

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The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development.

Nuclear Reactor Cooling Systems
Nuclear Reactor Cooling Systems

Author: Charles Nehme

Publisher: Charles Nehme

Published:

Total Pages: 73

ISBN-13:

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Nuclear reactor cooling systems are the backbone of nuclear power plants, playing a critical role in managing the immense heat generated by nuclear fission reactions and ensuring the safe operation of these facilities. These intricate systems are designed to effectively remove heat from the reactor core, transfer it to a secondary loop, and ultimately convert it into electricity. This book delves into the intricacies of nuclear reactor cooling systems, providing a comprehensive overview of their design, operation, and safety considerations. It explores the fundamental principles of heat transfer, the selection of suitable coolants, and the intricate flow patterns within the reactor core. Readers will gain a deep understanding of the different types of cooling systems, including pressurized water reactors (PWRs), boiling water reactors (BWRs), and gas-cooled reactors (GCRs). They will also gain insights into the advanced cooling concepts being explored for future nuclear power plants. Alongside these technical aspects, the book highlights the critical importance of safety in nuclear reactor cooling systems. Emergency cooling systems are meticulously designed to prevent and mitigate potential accidents, ensuring the protection of public health and the environment. This book is intended for a broad audience, including engineering students, professionals working in the nuclear industry, and those seeking a deeper understanding of nuclear power generation. It provides a valuable resource for anyone interested in learning about the intricate workings of nuclear reactor cooling systems and their role in harnessing nuclear energy responsibly.

Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor
Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor

Author: Samet Egemen Arda

Publisher:

Published: 2016

Total Pages: 142

ISBN-13:

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A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.

Nuclear Reactor Kinetics and Control
Nuclear Reactor Kinetics and Control

Author: Jeffery Lewins

Publisher: Elsevier

Published: 2013-10-22

Total Pages: 282

ISBN-13: 1483278751

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Nuclear Reactor Kinetics and Control highlights the application of classical control methods in the frequency space to the dynamic processes of a nuclear reactor. This book contains nine chapters and begins with an introduction to some important mathematical theories related to nuclear engineering, such as the Laplace and Fourier transforms, linear system stability, and the probability theory. The succeeding chapters deal with the frequency space of classical linear design. A chapter describes a stochastic model for the “lumped reactor and presents equations that measure the departure from the mean, as well as representative experiments or applications of the theory to neutron detection. The discussion then shifts to the aspects of reliability and its consequences for safety of nuclear reactors and some techniques for nonlinear studies centered on the use of the state space and its equations in the time domain. The final chapter introduces the modern electric analogue computer and derives the patching or programming rules that can be use to find solutions to problems of interest using the analogous behavior of electric circuits. This chapter also provide examples of intrinsic interest in nuclear engineering showing the programming involved and typical results, including the slower transients of xenon poisoning and fuel burn-up. This book is intended for nuclear engineers, physicists, applied mathematicians, and nuclear engineering undergraduate and postgraduate students.

Mathematical methods in Nuclear reactor Dynamics
Mathematical methods in Nuclear reactor Dynamics

Author: Ziya Akcasuh

Publisher: Elsevier

Published: 2012-12-02

Total Pages: 473

ISBN-13: 0323149081

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Mathematical Methods in Nuclear Reactor Dynamics covers the practical and theoretical aspects of point-reactor kinetics and linear and nonlinear reactor dynamics. The book, which is a result of the lectures given at the University of Michigan, is composed of seven chapters. The opening chapter of the book describes various physical phenomena influencing the temporal behavior of neutrons to provide insights into the physics of reactor dynamics and the interrelationships between various diverse phenomena. The text then presents a set of equations, called point kinetic equation, which describes the time behavior of the total power generated in the medium. The book also provides a short discussion on Gyftopoulos modification and Becker’s formulation. The next chapters explore the exact methods for solving the feedback-free point kinetic equations for a number of reactivity insertions and the validity of the various approximate methods of solution. The book also examines the derivation of models for a certain reactor type and briefly discusses the validity of these models in certain cases against experimental data. A chapter focuses on a concise presentation of the stability theory of linear systems with feedback. Lastly, the concepts of stability in nonlinear reactor systems and the criteria for asymptotic stability in the large as well as in a finite domain of initial disturbances are covered in the concluding chapter. The text is an ideal source for nuclear engineers and for those who have adequate background in reactor physics and operational and applied mathematics.

Correlation of Analysis with High Level Vibration Test Results for Primary Coolant Piping
Correlation of Analysis with High Level Vibration Test Results for Primary Coolant Piping

Author:

Publisher:

Published: 1992

Total Pages: 7

ISBN-13:

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Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results.

Dynamic Analysis of Coolant Circulation in Boiling Water Nuclear Reactors
Dynamic Analysis of Coolant Circulation in Boiling Water Nuclear Reactors

Author: Chathilingath K. Sanathanan

Publisher:

Published: 1964

Total Pages: 58

ISBN-13:

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The dynamics of two-phase flow through the coolant channels of a natural-circulation boiling water nuclear reactor is studied analytically. One-dimensional conservation equations describing the flow through each channel are written in a linearized perturbed form, and Laplace transformation in time is performed. A systematic procedure is developed to approximate the solution. The solution may be oscillatory both in time and space, and the stability depends largely upon the steady-state profile of velocity and void fraction along the channel, as well as the channel length. The simplifying assumption made by earlier investigators that the slip ratio is constant along the channel length is shown to yield results close to the true solution.