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Author: Robert Allen Walls
Publisher:
ISBN:
Category :
Languages : en
Pages : 95
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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: Robert Allen Walls
Publisher:
ISBN:
Category :
Languages : en
Pages : 95
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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: Rui Hu (Ph. D.)
Publisher:
ISBN:
Category :
Languages : en
Pages : 348
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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: Elias Amselem Abecasis
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
It is well known that boiling water reactors can experience inadvertent power oscillations. When such instability occurs the core can oscillate in two different modes (in phase mode and out of phase mode). In the late 90's a stability benchmark was created using the stability data obtained from the experiments at the Swedish nuclear power plant of Ringhals-1. Data was collected from the cycles 14, 15 , 16 and 17. Later on, this data was used to validate the various models and codes with the aim of predicting the instability behavior of the core and understand the triggers of such oscillations. The current trend of increasing reactor power density and relying on natural circulation for core cooling may have consequences for the stability of modern BWR's designs. The objective of this work is to find the most important parameters affecting the stability of the BWRs and propose alternative stability maps. For this purpose a TRACE/PARCS model of the Ringhals-1 NPP will be used. Afterwards a selection of possible parameters and dimensionless numbers will be made to study its effect on stability. Once those parameters are found they will be included in the stability maps to make them more accurate.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the U.S. NRC requirement of DR is not rigorously satisfied in the low-flow/high-power region; hence, this region should be avoided during normal startup and shutdown operations. The frequency of oscillation is shown to decrease as the flow rate is reduced. Moreover, the simulation frequency of 0.5Hz determined in the low-flow/high-power region is consistent with those observed during actual instability incidents. Additional numerical simulations show that in the low-flow/high-power region, for the same initial conditions, the use of point kinetics leads to damped oscillations, whereas the model that includes the modal neutron kinetics equations results in growing nonlinear oscillations.
Author: Karl Otto Ott
Publisher:
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 384
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Book Description
This text presents the theory and methods of prediction that are the heart of nuclear reactor safety. Time-dependent reactor behavior is explained in both mathematical and physical terms. This book also explains the logic behind the working formulas and calculational methods for reactor transients and illustrates typical dynamic responses. The classical concept of point kinetics is developed in three steps, with discussion of various solutions to kinetics problems. Each chapter includes homework problems and review questions.
Author: International Atomic Energy Agency
Publisher: IAEA
ISBN:
Category : Business & Economics
Languages : en
Pages : 656
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Book Description
Describes the state of knowledge of natural circulation in water cooled nuclear power plants and passive system reliability. The publication presents information on phenomena, models, predictive tools and experiments that currently support design and analysis of natural circulation systems, and highlights areas where additional research is needed.
Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201043191
Category : Business & Economics
Languages : en
Pages : 65
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Book Description
This updated version of Nuclear Energy Series NF-T-2.1 provides information on all aspects of fuel failures in current nuclear power plant operations.
Author: Amory B. Lovins
Publisher:
ISBN:
Category : Political Science
Languages : en
Pages : 520
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Book Description
Author: Dan Gabriel Cacuci
Publisher: Springer Science & Business Media
ISBN: 0387981306
Category : Science
Languages : en
Pages : 3701
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Book Description
This is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all levels, this book provides a condensed reference on nuclear engineering since 1958.
Author:
Publisher: OECD Publishing
ISBN:
Category : Business & Economics
Languages : en
Pages : 562
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Book Description
This publication sets out the findings of an international seminar, held in Aix-en-Provence, France in March 2004, which considered recent progress in the field of pellet-clad interaction in light water reactor fuels. It also reviews current understanding of relevant phenomena and their impact on the nuclear fuel rod under the widest possible conditions, and about both uranium-oxide and mixed-oxide fuels.
