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Progress On Pellet Cladding Interaction And Stress Corrosion Cracking
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Book Synopsis Progress on Pellet-Cladding Interaction and Stress Corrosion Cracking by : International Atomic Energy Agency
Download or read book Progress on Pellet-Cladding Interaction and Stress Corrosion Cracking written by International Atomic Energy Agency and published by . This book was released on 2021-08-15 with total page 312 pages. Available in PDF, EPUB and Kindle. Book excerpt: Flexible operation and related power changes can have a direct impact on fuel integrity through pellet-cladding interaction/stress corrosion cracking (PCI/SCC) phenomena, which could lead to fuel failures in certain conditions.
Book Synopsis Progress on Pellet-Cladding Interaction and Stress Corrosion Cracking - Experimentation, Modelling and Methodologies Applied to Support the Flexible Operation of Nuclear Power Plants - Report of a Technical Meeting by : International Atomic Energy Agency
Download or read book Progress on Pellet-Cladding Interaction and Stress Corrosion Cracking - Experimentation, Modelling and Methodologies Applied to Support the Flexible Operation of Nuclear Power Plants - Report of a Technical Meeting written by International Atomic Energy Agency and published by . This book was released on 2021 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis A Stress Corrosion Cracking Model for Pellet-Cladding Interaction Failures in Light-Water Reactor Fuel Rods by : D. Cubicciotti
Download or read book A Stress Corrosion Cracking Model for Pellet-Cladding Interaction Failures in Light-Water Reactor Fuel Rods written by D. Cubicciotti and published by . This book was released on 1979 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: A model for pellet-cladding interaction (PCI) fracture of light-water reactor (LWR) fuel rods is presented, the basis of which is that Zircaloy cladding fails by iodine stress corrosion cracking (SCC). Laboratory data on iodine SCC of irradiated Zircaloy provide the primary input to the model, but unirradiated Zircaloy SCC data and theoretical analyses are utilized to broaden the regime of validity to encompass the various power reactor observations.
Book Synopsis Pellet-cladding Interaction Failures in Water Reactor Fuel Rods by : Bernardo N. Nobrega
Download or read book Pellet-cladding Interaction Failures in Water Reactor Fuel Rods written by Bernardo N. Nobrega and published by . This book was released on 1981 with total page 270 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Chemical Aspects of Pellet-cladding Interaction in Light Water Reactor Fuel Elements by :
Download or read book Chemical Aspects of Pellet-cladding Interaction in Light Water Reactor Fuel Elements written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI.
Book Synopsis Stress Corrosion Crack Initiation and Growth and Formation of Pellet-Clad Interaction Defects by : L. Lunde
Download or read book Stress Corrosion Crack Initiation and Growth and Formation of Pellet-Clad Interaction Defects written by L. Lunde and published by . This book was released on 1979 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: The critical events leading to stress corrosion failure of fuel rods are poorly understood. The aim of the present work is to acquire data on stress corrosion cracking (SCC) initiation and crack growth and to use these data for improving our understanding of fuel rod failures.
Book Synopsis Iodine-Induced Stress Corrosion of Zircaloy Fuel Cladding by : L. Brunisholz
Download or read book Iodine-Induced Stress Corrosion of Zircaloy Fuel Cladding written by L. Brunisholz and published by . This book was released on 1987 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Because of its impact on nuclear power plant reliability, fuel cladding failure induced by pellet cladding interaction is a problem of major concern. It is usually analyzed as stress corrosion cracking caused by volatile fission products like iodine.
Book Synopsis Development and Application of 3-D Fuel Performance Modeling to Assess Missing Pellet Surface Influence on Pellet Clad Interaction and Clad Failure by : Nathan Allen Capps
Download or read book Development and Application of 3-D Fuel Performance Modeling to Assess Missing Pellet Surface Influence on Pellet Clad Interaction and Clad Failure written by Nathan Allen Capps and published by . This book was released on 2016 with total page 111 pages. Available in PDF, EPUB and Kindle. Book excerpt: In the late 1970s PCI related failures caused the implementation of startup ramp restrictions. These ramp restrictions where intended to reduce the stresses caused by pellet cladding contact. These ramp restrictions had a significant impact on Westinghouse fueled PWRs, reducing PCI related failure until 2003. Through investigation into these fuel rod failures lead to the conclusion that missing pellet surfaces (MPS) were the root cause of the failures. MPS are local geometric defects in nuclear fuel pellets that result from pellet mishandling or the manufacturing process. The presence of MPS defects can cause stress concentrations in the clad of sufficient magnitude to produce through-wall cladding failure for certain combinations of fuel burnup, and reactor power level or power change. Consequently, the impact of potential MPS defects has significant ly limited the rate of power increase, or ramp rate, in both pressurized and boiling water reactors (PWRs and BWRs, respectively). Improved three-dimensional (3-D) fuel performance models of MPS defect geometry can provide better understanding of the probability for pellet clad mechanical interaction (PCMI), and correspondingly the available margin against cladding failure by stress corrosion cracking (SCC). The Bison fuel performance code has been developed within the Consortium of Advanced Simulations of Light Water Reactors (CASL) to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding stress concentrations resulting from MPS defects. Bison is built upon the Multi-physics Object-Oriented Simulation Environment (MOOSE) developed at Idaho National Laboratory (INL). MOOSE is a massively parallel finite element computational system that uses a Jacobian-free, Newton-Krylov (JFNK) method to solve coupled systems of non-linear partial differential equations. In addition, the MOOSE framework provides the ability to effectively use massively parallel computational capabilities needed to create high fidelity 3-D models of a fuel rod, as well as full-length R-Z rods, and R-Theta geometric representation. This PhD dissertation documents my contributions to the development of Bison, specifically focused on verification and validation of a 2-D, axi-symmetric version of Bison through benchmarking comparisons to Falcon model predictions and Halden Instrumented Fuel Assembly (IFA) experiments of both thermal and mechanical behavior. Initial benchmark comparisons indicate that Bison predictions agree quite well with 2-D Falcon predictions and Halden experimental data on fuel centerline temperature but that further developments are necessary for some models, including fission gas release and gaseous swelling. The mechanical behavior benchmarking study has compared predictions of clad deformation to dilatational measurements, and the results show promising agreement. Subsequently, this dissertation documents my evaluation of the cladding hoop stress distributions as a function of MPS defect geometry and the presence of discrete pellet cracks for a set of typical operating conditions in a PWR fuel rod, as a function of reactor operating history. These results provide a first step in a probabilistic approach to assess cladding failure during power maneuvers. My research provides insight into how varying pellet defect geometries affect the distribution of the cladding stress, as well as the temperature distributions within the fuel and clad; and are used to develop stress concentration factors for comparing 2-D and 3-D models. Finally, the objective of this dissertation is to develop a methodology to determine rod failure, and then to utilize the resulting failure criteria to evaluate specific historical MPS and PCI failures.
Book Synopsis Formulation for the Analysis of Pellet-cladding Mechanical Interaction by :
Download or read book Formulation for the Analysis of Pellet-cladding Mechanical Interaction written by and published by . This book was released on 1979 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A formulation has been derived for the thermoelastic analysis of pellet-cladding mechanical interaction. The formulation is based on a combination of finite element analysis and the matrix-displacement method. Boundary conditions at the pellet-cladding and pellet-pellet contacting sites can be treated realistically by considering the force and displacement relationships at the interfaces. As a result, no simulating gap elements need be used and the analysis of the two limiting cases, fuel-cladding bonding (stick) and fuel-cladding slippage (slip), follows directly from imposing the appropriate boundary conditions. From the principle of superposition, this formulation also gives a very compact scheme which can be used to study a variety of power ramp cases with ease. No additional finite element analysis is required after the solutions for a number of individual base-case problems have been obtained. Because of thermoelastic response of a fuel-element during a power ramp corresponds to that for a fast ramp rate, which subjects cladding to the most severe mechanical loading, the results from our thermoelastic analysis can be very useful in evaluating the likelihood of cladding failure when they are combined with knowledge of the cladding failure mode (plastic instability or stress-corrosion cracking). Since the failure mode for the Light Water Reactor fuel elements is predominantly stress-corrosion cracking and there is generally no time-independent cladding plastic deformation during a power ramp, our formulation may be applied directly to provide an assessment of the permissible power ramps. This could lead to a relaxation of the current restrictive and empirically based reactor-operation rules.
Book Synopsis Pellet-clad Interaction in Water Reactor Fuels by :
Download or read book Pellet-clad Interaction in Water Reactor Fuels written by and published by OECD Publishing. This book was released on 2005 with total page 562 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.
Book Synopsis Development of Zirconium-Barrier Fuel Cladding by : JS. Armijo
Download or read book Development of Zirconium-Barrier Fuel Cladding written by JS. Armijo and published by . This book was released on 1994 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper was prepared for the 1991 Kroll Award. A review is presented of the development of barrier fuel. It includes the recognition of the pellet-cladding interaction (PCI) fuel failure mode and of a coordinated program to develop understanding, mitigating strategies, and a fuel that is resistant to this failure mode. The efforts to understand PCI led to the conclusion that the dominant mechanism is stress-corrosion cracking of the Zircaloy. The invention and development of zirconium-barrier fuel was intended to provide a materials solution to this fuel failure mode. This review includes the work to understand the failure mechanism as well as the program to develop PCI-resistant fuel designs. Ultimately, the zirconium-barrier fuel was tested in power ramps to ascertain and to quantify the resistance to PCI under expected service conditions in commercial boiling water reactors (BWRs). The program that led to a large-scale demonstration in a commercial power plant (Quad Cities-2) is described briefly. Subsequent to that, program work continued with in-reactor load following and experiments in a test reactor on power cycling of barrier fuel. Finally, the performance of failed fuel is discussed briefly.
Book Synopsis Thermo-chemical-mechanical Modeling of Nuclear Fuel Behavior by : Piotr Konarski
Download or read book Thermo-chemical-mechanical Modeling of Nuclear Fuel Behavior written by Piotr Konarski and published by . This book was released on 2019 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation.
Book Synopsis Thermo-chemical-mechanical Modeling of Nuclear Fuel Behavior by : Piotr Konarski
Download or read book Thermo-chemical-mechanical Modeling of Nuclear Fuel Behavior written by Piotr Konarski and published by . This book was released on 2019 with total page 317 pages. Available in PDF, EPUB and Kindle. Book excerpt: The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation.
Book Synopsis Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon by : M. Peehs
Download or read book Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon written by M. Peehs and published by . This book was released on 1979 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: To investigate iodine stress corrosion cracking on the behavior of Zircaloy tubing, a standard test procedure with internally pressurized creep specimens (I-SCC standard test) and a split-ring test (I-SCC laboratory test) were developed. The threshold value for brittle cracking in iodine is 10-6 g iodine per square centimetre. The uniform elongation shows a clear minimum in its dependence on strain rate. Basal pole orientations in the range of ±50 to ±70 deg relative to the radial direction are the most I-SCC sensitive textures. The I-SCC process occurs in several stages; incubation, crack nucleation, and propagation. A thermodynamic evaluation indicates that I-SCC only occurs when zirconium iodides condense on the Zircaloy surface. Results show that comparisons of the I-SCC susceptibility of tubing manufactured in different manners should be made at the same point on the strain rate versus uniform elongation curves; for example, at the strain rate with the minimum of uniform elongation.
Book Synopsis Development of ZirconiumBarrier Fuel Cladding by : Herman S. Rosenbaum
Download or read book Development of ZirconiumBarrier Fuel Cladding written by Herman S. Rosenbaum and published by . This book was released on 2010 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper was prepared for the 1991 Kroll Award. A review is presented of the development of barrier fuel. It includes the recognition of the pellet-cladding interaction (PCI) fuel failure mode and of a coordinated program to develop understanding, mitigation strategies, and a fuel that is resistant to this failure mode. The efforts to understand PCI led to the conclusion that the dominant mechanism is stress-corrosion cracking of the Zircaloy. The invention and development of zirconium-barrier fuel was intended to provide a materials solution to this fuel failure made. This review includes the work to understand the failure mechanism as well as the program to develop PCI-resistant fuel designs. Ultimately, the zirconium-barrier fuel was tested in power ramps to ascertain and to quantify the resistance to PCI under expected service conditions in commercial boiling water reactors (BWRs). The program that led to a large-scale demonstration in a commercial power plant (Quad Cities-2) is described briefly. Subsequent to that, program work continued with in-reactor load following and experiments in a test reactor on power cycling of barrier fuel. Finally, the performance of failed fuel is discussed briefly. The original paper was published by ASTM International in STP 1245, Zirconium in the Nuclear Industry: Tenth International Symposium, 1994, pp. 318.
Book Synopsis Comprehensive Nuclear Materials by :
Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field
Book Synopsis Proceedings of the 2023 Water Reactor Fuel Performance Meeting by : Jianqiao Liu
Download or read book Proceedings of the 2023 Water Reactor Fuel Performance Meeting written by Jianqiao Liu and published by Springer Nature. This book was released on 2023-11-30 with total page 384 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Water Reactor Fuel Performance Meeting (WRFPM) held in Asia has merged with TopFuel in Europe and LWR Fuel Performance in the United States to form the globally most influential conference in the field of nuclear fuel research. WRFPM2023 is organized by Chinese Nuclear Society (CNS) in cooperation with the Atomic Energy Society of Japan (AESJ), Korean Nuclear Society (KNS), European Nuclear Society (ENS), American Nuclear Society (ANS), the Interna-tional Atomic Energy Agency (IAEA) with the support from China Nuclear Energy In¬dustry Corporation (CNEIC) and TVEL. Conference Topics: 1. Advances in water reactor fuel technology and testing 2. Operation and experience 3. Transient and off-normal fuel behaviour and safety related issues 4. Fuel cycle, used fuel storage and transportation 5. Innovative fuel and related issues 6. Fuel modelling, analysis and methodology
