Author: Sousan Abolhassani Publisher: ISBN: Category : Materials Languages : en Pages : 31
Book Description
This paper provides the results of investigations by transmission electron microscopy (TEM) on the selected materials from in-reactor oxidation tests in the Halden test reactor (Reference No. IFA-638) from 1998 to 2006. The objective of the IFA-638 test was to study the corrosion behavior of modern zirconium-based claddings to high burnup in pressurized water reactor water chemistry and thermal hydraulic conditions. The aim of this paper is to report on the microstructure of selected materials (ZIRLO®, E635, and Alloy A) after the irradiation to different burnup levels to determine the modifications induced by irradiation and to correlate results to their oxidation behavior. The TEM examinations revealed the nature of secondary phase particles (SPPs) and their modification under irradiation. Four types of SPPs were observed, namely ?-niobium precipitates, Zr0.5Nb0.3Fe0.2 (mainly in the ZIRLO alloy), Zr(Fe,Nb)2 (in E635), and (Cr,Fe)2Zr,Nb with varying niobium content (present in Alloy A: Zr-0.58Sn-0.31Nb-0.36Fe-0.26Cr). TEM observations showed that all three materials contained still several precipitates after irradiation and in the case of the ZIRLO alloy even after high burnups. Furthermore, the analysis of the metal side of the interface and its comparison with the oxide side led to the conclusion that all types of precipitates dissolved to some extent under irradiation and that their alloying element content decreased. The dissolution was intensified in the oxide. However, a more detailed examination showed that the ?-niobium precipitates dissolved at a slower rate, or knowing that their composition was much richer in niobium, the time needed for the precipitates to become fully depleted from niobium was longer. Regarding the amorphization under irradiation, the ?-niobium- and chromium-containing precipitates did not amorphize in the metal part of the interface. This was not the case for the other types of precipitates. Furthermore, these two types of SPP both showed delayed oxidation and due to this behavior the typical crack above the SPP in the oxide was also observed. These results are discussed to gain an improved understanding of the oxidation behavior of materials studied as a function of irradiation and residence time.
Author: Gerry D. Moan Publisher: ASTM International ISBN: 0803128959 Category : Nuclear fuel claddings Languages : en Pages : 891
Book Description
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.
Author: OECD Nuclear Energy Agency Publisher: OECD Publishing ISBN: Category : Technology & Engineering Languages : en Pages : 74
Book Description
Presents brief descriptions of 20 fuel-related safety criteria along with both the rationale for having such criteria and possible new design and operational issues which could have an effect on them.
Author: Joshua Samuel Bowman Publisher: ISBN: Category : Languages : en Pages :
Book Description
In the operation of a nuclear reactor, the performance of the fuel cladding is critical to ensuring safe and reliable operation of the reactor. The current generation of Light Water Reactors utilizes claddings made from zirconium alloys. The material used for nuclear reactors must be able to withstand temperatures above 3000C while also being exposed to water, high pressures, and radiation. During operation, the zirconium cladding corrodes and picks up hydrogen which can adversely affect its performance. The corrosion mechanisms at work have yet to be fully characterized, especially the influence of irradiation. In order to better understand the mechanisms at work and characterize the behavior of zirconium alloys under reactor conditions, the Mechanistic Understanding of Zirconium Alloy Corrosion (MUZIC) consortium focused on the autoclave corrosion (MUZIC-1) and hydrogen pickup (MUZIC-2) outside of irradiation. The MUZIC-3 effort focuses on corrosion under irradiation. While it would be optimal to test reactor-irradiated samples, the difficulties posed by irradiating, corrosion testing, and examining these samples makes ion irradiation a more appealing manner of irradiation. Using doses and temperatures adjusted for substitution of protons for neutron radiation, this experiment seeks to characterize the effects of irradiation on the base metal, oxide layer, and water, both separately and jointly, on the corrosion of zirconium alloys. In this thesis, the beginning stages of this project, part of MUZIC-3, are presented. This involves verification of the effect of proton irradiation (which is used to represent neutron irradiation) on the base metal and characterization of the irradiated samples. The corrosion testing of this irradiated material will provide a reference for the effect of irradiation induced microstructure changes to the base metal on corrosion. In order to characterize the samples, chemical analyses and observations on crystallinity of secondary phase particles are needed. Along with the analysis of second-phase precipitates, assessment of dislocation loops to observe similarities between different radiation types is also required. Accordingly, samples were irradiated with charged particles (protons and zirconium ions) at the Michigan Ion Beam Laboratory and focused ion beam samples were prepared for transmission electron microscopy examination. The microstructure of the base metal is examined for a range of doses and irradiation temperatures and compared to the microstructure created under neutron irradiation as a preliminary to corrosion testing of irradiated samples. The results are discussed in light of existing literature.
Author: Nihon Genshiryoku Kenkyūjo Publisher: Paris, France : Nuclear Energy Agency, Organisation for Economic Co-operation and Development ISBN: Category : Science Languages : en Pages : 370
Book Description
In response to increasing interest in high-temperature, gas-cooled reactors (HTGRs) in many countries and the need for improved materials for nuclear applications in high-temperature environments, the NEA organised a Second Information Exchange Meeting on Basic Studies in the Field of High-temperature Engineering. These proceedings provide an overview of the activities being carried out in eight countries, the improvement of material properties for HTGR application, in-core monitoring methods and properties of irradiated graphite, and HTGR fuel fabrication and performance.
Author: GP. Airey Publisher: ISBN: Category : Anodic coatings Languages : en Pages : 14
Book Description
Microstructural characterization of oxide films formed on zirconium-based alloys was performed by use of scanning and transmission electron microscopy. Examination of pre-transition films formed on Zircaloy-4 oxidized in 360°C (680°F) water revealed a small grain size of approximately 100 Å (10 nm) diameter. In addition, a gradation of grain size was observed throughout the film thickness, such that at the oxide-water surface (oldest oxide), a grain size of less than 50 Å (5nm) was observed, and at the metal-oxide surface (newest oxide), the grain size was approximately 200 Å (20 nm). In post-transition films the outermost oxide still possessed the very fine 50 Å (5 nm) diameter grain size. However, the newest oxide of post-transition films consisted of relatively large grains, with grain diameters of 1000 to 5000 Å (100 to 500 nm). At the midthicknesses of these oxides intermediate grain sizes were observed. The bulk of the post-transition films was highly porous. Pore sizes ranged from approximately 10 to 150 Å (1 to 15 nm), and many connected pores were concentrated at the grain boundaries. Under more severe oxidizing conditions, imposed by increasing the corrosion temperature to 427°C (800°F), the growth of large grains at the metal-oxide interface was unstable and film growth proceeded by the nucleation of finer grains.
Author: P.A. Schubiger Publisher: Springer Science & Business Media ISBN: 3540495274 Category : Science Languages : en Pages : 348
Book Description
Personalized medicine employing patient-based tailor-made therapeutic drugs is taking over treatment paradigms in a variety of ?elds in oncology and the central nervous system. The success of such therapies is mainly dependent on ef?cacious therapeutic drugs and a selective imaging probe for identi?cation of potential responders as well as therapy monitoring for an early bene?t assessment. Molecular imaging (MI) is based on the selective and speci?c interaction of a molecular probe with a biological target which is visualized through nuclear, magnetic resonance, near infrared or other methods. Therefore it is the method of choice for patient selection and therapy monitoring as well as for speci?c e- point monitoring in modern drug development. PET (positron emitting tomography), a nuclear medical imaging modality, is ideally suited to produce three-dimensional images of various targets or processes. The rapidly increasing demand for highly selective probes for MI strongly pushes the development of new PET tracers and PET chemistry. ‘PET chemistry’ can be de?ned as the study of positron-emitting compounds regarding their synthesis, structure, composition, reactivity, nuclear properties and processes and their properties in natural and - natural environments. In practice PET chemistry is strongly in?uenced by the unique properties of the radioisotopes used (e. g. , half-life, che- cal reactivity, etc. ) and integrates scienti?c aspects of nuclear-, organic-, inorganic- and biochemistry.
Author: International Atomic Energy Agency Publisher: IAEA ISBN: Category : Business & Economics Languages : en Pages : 100
Book Description
The reactors around the world have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts worldwide. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). This publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term.
Author: Sousan Abolhassani
Author: Gerry D. Moan
Author: OECD Nuclear Energy Agency
Author: Joshua Samuel Bowman
Author: International Atomic Energy Agency
Author: Nihon Genshiryoku Kenkyūjo
Author: GP. Airey
Author: P.A. Schubiger
Author: International Atomic Energy Agency
Author: United States. Congress. Senate. Committee on Foreign Relations