Download or read book Quantifying Irradiation Defects in Zirconium Alloys written by Levente Balogh and published by . This book was released on 2018 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-induced dislocations significantly affect the mechanical properties of zirconium alloys, altering slip and influencing creep and growth. Thus, the quantitative characterization of irradiation defects as a function of fluence, cold work, and/or thermal treatments is important for models that attempt to predict their impact on properties. Whole-pattern diffraction line-profile analysis (DLPA) is a well-established modern tool for microstructure characterization based on first-principle physical models for dislocation density measurements in plastically deformed materials. However, applying these DLPA methods directly to irradiated materials yields higher than expected dislocation density values compared with historical transmission electron microscopy (TEM) measurements and past line-broadening analysis studies calibrated to TEM observations. In an effort to understand these differences, a new microstructural model was developed for DLPA to specifically address dislocation structures consisting of elliptical a- and c-component loops. To compare the refined DLPA method with TEM measurements, high-resolution neutron diffraction patterns on nonirradiated and irradiated Zr-2.5Nb samples were collected with the Neutron Powder Diffractometer instrument at the Los Alamos Neutron Science Center and were evaluated. High-resolution TEM measurements were performed at the Reactor Materials Testing Laboratory, Queen's University, for comparison with the DLPA results. The capabilities and inherent uncertainties of both the refined DLPA and TEM methods are compared and discussed in detail. We show that the differences between the density values provided by DLPA and TEM are inherent to the methods and can be reconciled with the interpretation of the data.
Download or read book Experimental and Modeling Approach of Irradiation Defects Recovery in Zirconium Alloys written by Joël Ribis and published by . This book was released on 2008 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: During neutron irradiation, both interstitial and vacancy loops are formed in high concentration in zirconium alloys. Due to this high density of loops, the material is considerably hardened, but the recovery of the radiation damage during a heat treatment leads to a progressive softening of the irradiated material. The recovery of the radiation induced hardening has been investigated using microhardness tests. Transmission electron microscopy (TEM) observations performed on irradiated foils have also shown that the loop density falls while the loop size increases during the thermal annealing. Furthermore, the TEM analysis has revealed that only vacancy loops are present in the material after long term annealing, the interstitial loops having entirely disappeared. A numerical cluster dynamic modeling has also been used in order to reproduce the material recovery for various annealing conditions. The microstructural evolution during mechanical testing with various loading conditions has also been studied. It has been shown that during a creep test with low applied stress (130 MPa) and high temperature (450°C), the microstructure evolution can essentially be explained by the thermal recovery of the loops leading to glide of dislocations as found for an non-irradiated material. At intermediate temperature (400°C), it is shown that for low stress level (130 MPa) the microstructure evolution can also be explained by the thermal recovery of loops, whereas for higher stress (250 MPa), sweeping of loops by gliding dislocations can also occur. In addition, for an applied stress of 130 MPa and a temperature of 400°C, dislocation density is higher in the irradiated material than in the non-irradiated material deformed in the same conditions. It is also shown that secondary slip systems are more activated in the irradiated material than in the non-irradiated material. From this detailed analysis, the mechanical behavior during creep is interpreted in terms of microscopic deformation mechanisms.
Download or read book Diffusion and Defect Studies in Zirconium and some of its Alloys written by R.P. Agarwala and published by Trans Tech Publications Ltd. This book was released on 2004-04-27 with total page 90 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book is divided into two parts: the first part describes diffusion processes, and the second part describes radiation damage to - and cold-working of - zirconium and some of its important alloys.
Download or read book Irradiation Damage Recovery in Some Zirconium Alloys written by GJC Carpenter and published by . This book was released on 1974 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: The recovery of irradiation damage in a number of zirconium alloys has been studied by means of hardness measurements. The experiments were designed to examine the effect of different solutes (copper, aluminum, titanium, niobium, molybdenum) and metallurgical condition on the stability of irradiation-induced defect clusters.
Download or read book Creep of Zirconium Alloys in Nuclear Reactors written by and published by ASTM International. This book was released on with total page 308 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Study of Point Defect Mobilities in Zirconium During Electron Irradiation in a HVEM written by Atomic Energy of Canada Limited and published by Chalk River, Ont. : Chalk River Laboratories. This book was released on 1993 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: A high voltage electron microscope (HVEM) was used to investigate the nature of intrinsic point effects in a-Zirconium by direct observation of dislocation climb and cavity growth or shrinkage. The material used was Marz-grade zirconium that had been pre-irradiated with neutrons at 740 K in the Dounreay Fast Reactor. Dislocation loops of vacancy character that were produced during the neutron irradiation were studied by further irradiation with electrons in the HVEM. C-component network dislocations and cavities were also studied.
Download or read book Transmission electron microscopy of irradiation induced defects on zirconium alloys written by D. O. Northwood and published by . This book was released on 1976 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The formation of c component defects in zirconium alloys during neutron irradiation written by M. Griffiths and published by . This book was released on 1987 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Impact of Irradiation Damage Recovery During Transportation on the Subsequent Room Temperature Tensile Behavior of Irradiated Zirconium Alloys written by B. Bourdiliau and published by . This book was released on 2010 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on a nuclear power plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep, a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1 % Nb and stress-relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400 and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The transmission electron microscopy examinations, conducted on recrystallized Zr-1 % Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation or at the beginning of dry storage.
Download or read book Mechanical Properties of Irradiated Zirconium Zircaloy and Aluminum written by Richard E. Schreiber and published by . This book was released on 1961 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Neutron Irradiation Induced Defect Structures in Zirconium written by RG. Blake and published by . This book was released on 1979 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: A transmission electron microscope (TEM) was used to study the neutron irradiation-induced defect structures in zirconium from various sources irradiated to fluences up to 1.3 x 1025 neutrons (n)/m-2 > 0.1 MeV and temperatures in the range 478 to 823 K. Below 673 K the predominant form of damage consists of 1/3 1120 dislocation loops. At higher irradiation temperatures, depending on specimen purity, the defects may include faulted 1/6 2023 loops and voids in addition to the perfect 1/3 1120 loops. The temperature regime of the stability of the various defects was defined. Quantitative measurements showing the effect of irradiation temperature and specimen purity are presented and discussed.
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
Download or read book Characterisation of Neutron Irradiation Damage in Zirconium Alloys written by P. M. Kelly and published by . This book was released on 1977 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation damage recovery in some zirconium alloys written by G. J. C. Carpenter and published by . This book was released on 1974 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effect of Stress on Radiation Damage in Neutron Irradiated Zirconium Alloys written by CE. Coleman and published by . This book was released on 1977 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: Structures developed in zirconium alloys during irradiation creep have been characterized by transmission electron microscopy (TEM). Alloys investigated were annealed Zircaloy-2, cold-worked Zircaloy-2 and cold-worked Zr-2.5Nb pressure tube material. Thin films were taken from material deformed in the NRU, NRX and Pickering-3 reactors at temperatures of 530 to 600 K under stresses of 117 to 552 MPa giving strains in the range 0.14 to 8.8 percent. Stress-induced orientation of dislocation loops makes a negligible contribution to irradiation creep at all stresses. At the lower stresses (and hence strains), the size and distribution of the damage is unaffected by stress, being the same in the head and gage sections of creep specimens. At higher stresses (strains), there is much clearing of the damage by plastic deformation. The deformation however is very uneven, producing structures in different grains of the same specimen that can show no deformation, swaths cleared of irradiation damage, or dislocation tangles or cell formation. The relevance of these TEM observations to irradiation creep mechanisms is discussed.
Download or read book Zirconium in the Nuclear Industry written by Gerry D. Moan and published by ASTM International. This book was released on 2002 with total page 891 pages. Available in PDF, EPUB and Kindle. Book excerpt: Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
Download or read book Peculiarities of Structural and Behavioral Changes of Some Zirconium Alloys at Damage Doses of Up to 50 Dpa written by VN. Shishov and published by . This book was released on 2004 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: The irradiation-induced damage of zirconium alloys subjected to neutron irradiation up to dose levels of ~50 dpa was investigated. Specimens of unalloyed zirconium, Zr-1%Nb, Zr-2.5%Nb and Zr-1%Nb-1.3%Sn-0.4%Fe were irradiated in the BOR-60 reactor over the temperature range 320-420°C. The dose dependence of the irradiation growth strain increased sharply in zirconium and Zr-Nb irradiated at ~350°C at doses above ~10 dpa. In the case of Zr-1%Nb-1.3%Sn-0.4%Fe, it increased at doses of ~37 dpa. Upon increasing the irradiation temperature to 420°C, a sharp accelerated irradiation growth of the Zr-1%Nb alloy began shifting up to about 30 dpa. For the Zr- 1%Nb-1.3%Sn-0.4%Fe, no change of the irradiation growth rate was observed up to a dose of 55 dpa. The onset of increased irradiation growth in alloys correlates with the occurrence of c-component dislocation loops which coincides with a loss of coherence of finely-dispersed precipitates. Post-irradiation annealing experiments demonstrated that a delay in loop formation leads to displacement of the "break-away" beginning in the dose dependence of the irradiation growth in the direction of high doses. The a+c-type dislocation loops were also formed in Zr-1%Nb alloy at high doses, but their influence on the change of macroscopic properties was not observed.
