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Sample records for ceramography

  1. Ceramography and segmentation of polycristalline ceramics: application to grain size analysis by automatic methods

    Energy Technology Data Exchange (ETDEWEB)

    Arnould, X.; Coster, M.; Chermant, J.L.; Chermant, L. [LERMAT, ISMRA, Caen (France); Chartier, T. [SPCTS, ENSCI, Limoges (France)

    2002-07-01

    The knowledge of the mean grain size of ceramics is a very important problem to solve in the ceramic industry. Some specific methods of segmentation are presented to analyse, by an automatic way, the granulometry and morphological parameters of ceramic materials. Example presented concerns cerine materials. Such investigations lead to important information on the sintering process. (orig.)

  2. Metallography. With an introduction to ceramography. 15. rev. and enl. ed.; Metallografie. Mit einer Einfuehrung in die Keramografie

    Energy Technology Data Exchange (ETDEWEB)

    Oettel, Heinrich [Technische Univ. Bergakademie Freiberg (Germany). Inst. fuer Werkstoffwissenschaft; Schumann, Hermann (eds.)

    2011-07-01

    This revised primer comprises the following chapters: Structures of inorganic materials (metals and ceramics); Metallographic techniques; Phase equilibria and state diagrams; Influence of processing and treatment on the structure of metals and alloys; Iron and iron alloys; Structures of technical nonferrous metals and their alloys; High-performance ceramics. [German] Dieses aktualisierte Grundlagenwerk ist in folgende Kapitel aufgeteilt: Strukturen anorganischer Werkstoffe (Metalle und Keramiken); Metallografische Arbeitsverfahren; Phasengleichgewichte und Zustandsdiagramme; Einfluss der Verarbeitung und Behandlung auf die Gefuegeausbildung von Metallen und Legierungen; Eisen und Eisenlegierungen; Gefuege der technischen Nichteisenmetalle und ihrer Legierungen; Hochleistungskeramik.

  3. Non destructive localization by gamma tomography of fission products emission in a fuel pencil area after an accidental sequence; Localisation non destructive par tomographie gamma d`emission des produits de fission dans une section de crayon combustible ayant subi une sequence accidentelle

    Energy Technology Data Exchange (ETDEWEB)

    Ducros, G.; Confort, E.; Drevon, P.

    1994-12-31

    The transverse gamma spectrometry of an irradiated fuel pencil, associated to a tomography method of gamma emitters reconstruction in a straight section of the pencil, is a non destructive technique which proves efficient when there are enough projections. For a little irradiated fuel, four to six projections equally distributed on 180 degrees and associated to an activity repartition initial model with symmetry of revolution are generally sufficient. For a very damaged fuel, it is necessary to practice eighteen projections every ten degrees; the initial repartition model is no more a noticeable parameter. Applied to the VERCORS safety programme, it allows to know the fuel behavior in situation of serious accident and the radial migrations of fission products. The localization of gamma emitters fission products brings informations supplementary to others optical techniques (ceramographies), or chemical techniques (micro-probe). Then, by its non destructive aspect, it can be in certain cases, unavoidable in front of usual destructive techniques. (N.C.). 4 refs.

  4. Study of the properties of the Am-O system in view of the transmutation of Am 241 in fast reactors; Etude des proprietes du systeme Am-O en vue de la transmutation de l`americium 241 en reacteur a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Casalta, S.

    1996-04-01

    To reduce the long term toxicity of Am 241 it was considered to transmute this isotope in fast reactor. The first part of this thesis is an introduction at this problem. In the second part we give the experimental techniques used for the realisation of an AmO{sub 2}-MgO target (powder metallurgy under inert, oxidizing or reducing atmosphere). The properties of the Am-O system has been analyzed by X diffraction, thermodynamic and ceramography, in the Am{sub 2}O{sub 3}-AmO{sub 2} field. In the third part we study the external exposure risk created by the manufacturing of this target and in the last part the behavior of this target in a fast reactor. 66 refs., 28 figs., 25 tabs., 1 append.

  5. HRB-22 capsule irradiation test for HTGR fuel. JAERI/USDOE collaborative irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo; Sawa, Kazuhiro; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-03-01

    As a JAERI/USDOE collaborative irradiation test for high-temperature gas-cooled reactor fuel, JAERI fuel compacts were irradiated in the HRB-22 irradiation capsule in the High Flux Isotope Reactor at the Oak Ridge National Laboratory (ORNL). Postirradiation examinations also were performed at ORNL. This report describes 1) the preirradiation characterization of the irradiation samples of annular-shaped fuel compacts containing the Triso-coated fuel particles, 2) the irradiation conditions and fission gas releases during the irradiation to measure the performance of the coated particle fuel, 3) the postirradiation examinations of the disassembled capsule involving visual inspection, metrology, ceramography and gamma-ray spectrometry of the samples, and 4) the accident condition tests on the irradiated fuels at 1600 to 1800degC to obtain information about fuel performance and fission product release behavior under accident conditions. (author)

  6. AFC-1 Transmutation Fuels Post-Irradiation Hot Cell Examination 4-8 at.% - Final Report (Irradiation Experiments AFC-1B, -1F and -1Æ)

    Energy Technology Data Exchange (ETDEWEB)

    Bruce Hilton; Douglas Porter; Steven Hayes

    2006-09-01

    The AFC-1B, AFC-1F and AFC-1Æ irradiation tests are part of a series of test irradiations designed to evaluate the feasibility of the use of actinide bearing fuel forms in advanced fuel cycles for the transmutation of transuranic elements from nuclear waste. The tests were irradiated in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) to an intermediate burnup of 4 to 8 at% (2.7 - 6.8 x 1020 fiss/cm3). The tests contain metallic and nitride fuel forms with non-fertile (i.e., no uranium) and low-fertile (i.e., uranium bearing) compositions. Results of postirradiation hot cell examinations of AFC-1 irradiation tests are reported for eleven metallic alloy transmutation fuel rodlets and five nitride transmutation fuel rodlets. Non-destructive examinations included visual examination, dimensional inspection, gamma scan analysis, and neutron radiography. Detailed examinations, including fission gas puncture and analysis, metallography / ceramography and isotopics and burnup analyses, were performed on five metallic alloy and three nitride transmutation fuels. Fuel performance of both metallic alloy and nitride fuel forms was best correlated with fission density as a burnup metric rather than at.% depletion. The actinide bearing transmutation metallic alloy compositions exhibit irradiation performance very similar to U-xPu-10Zr fuel at equivalent fission densities. The irradiation performance of nitride transmutation fuels was comparable to limited data published on mixed nitride systems.

  7. Synthesis of UO{sub 2} and ThO{sub 2} doped with Gd{sub 2}O{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Baena, Angela [KU Leuven, Department of Chemistry, Celestijnenlaan 200F, P.O. Box 2404, B-3001 Heverlee (Belgium); Belgian Nuclear Research Centre (SCK-CEN), Institute for Nuclear Materials Science, Boeretang 200, B-2400 Mol (Belgium); Cardinaels, Thomas, E-mail: thomas.cardinaels@sckcen.be [Belgian Nuclear Research Centre (SCK-CEN), Institute for Nuclear Materials Science, Boeretang 200, B-2400 Mol (Belgium); Vos, Benedict [Belgian Nuclear Research Centre (SCK-CEN), Institute for Nuclear Materials Science, Boeretang 200, B-2400 Mol (Belgium); Binnemans, Koen [KU Leuven, Department of Chemistry, Celestijnenlaan 200F, P.O. Box 2404, B-3001 Heverlee (Belgium); Verwerft, Marc [Belgian Nuclear Research Centre (SCK-CEN), Institute for Nuclear Materials Science, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-15

    Uranium dioxide (urania, UO{sub 2}) and thorium dioxide (thoria, ThO{sub 2}) doped with gadolinium oxide (gadolinia, Gd{sub 2}O{sub 3}) were prepared via solid-state synthesis. For Gd{sub 2}O{sub 3}-doped ThO{sub 2}, also an alternative, semi-dry process (“suspension coating”) was applied in which Gd{sub 2}O{sub 3}-coated ThO{sub 2} powder was produced via suspension drying followed by calcination. The microstructure and homogeneity of the materials were investigated by ceramography, EPMA and XRD. Solid-state synthesis is a convenient method to produce Gd{sub 2}O{sub 3}-doped UO{sub 2}. However, this route was found to be inappropriate to obtain Gd{sub 2}O{sub 3}-doped ThO{sub 2} with an acceptable microstructure and homogeneity. The suspension coating process reported in this work is a simple and practical method to overcome these issues.

  8. The sintering of uranium carbide and of uranium-plutonium carbide, and the role of nickel as a sintering additive

    Science.gov (United States)

    Pickles, S.; Yates, G.; Bramman, J. I.; Finlayson, Moira B.

    1980-04-01

    A comparison of the experimentally determined sintering kinetics for uranium and uranium-plutonium carbides of different stoichiometries with calculations for various theoretical models has been used to indicate probable sintering mechanisms. A bulk diffusion model with activation energies approximating to those for chemical diffusion under a concentration gradient is thought to apply. Ceramography has been used to study the influence of changes in composition and sintering atmosphere on grain size and microstructure, with the conclusion that grain growth is impeded by the presence of a grain-boundary second phase. The role of nickel as a sintering aid has also been investigated using, in addition to the above techniques, electron microprobe analysis and X-ray diffraction for chemical identification of phases. It is concluded that the first stage of sintering is one of particle rearrangement in a binary metallic liquid phase (U-Ni), followed by a solution-precipitation process. On prolonged annealing ternary U-C-Ni phases are produced, dominated by the composition U 2NiC 3.

  9. Characterization of spent fuel approved testing material--ATM-104

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  10. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  11. Restructuring and redistribution of actinides in Am-MOX fuel during the first 24 h of irradiation

    Science.gov (United States)

    Tanaka, Kosuke; Miwa, Shuhei; Sekine, Shin-ichi; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shin-ichi

    2013-09-01

    In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.

  12. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    accordance with the results of ceramography and density measurements. The possible fuel loss and Cs yield from the failed rod were evaluated according to the results of gamma scanning of fuel rod meats. As a result of the spent fuel assembly examinations it was established that in most cases the failures of the VVER fuel rods are related to their operation features. The main mechanisms of fuel rod failures are the fretting corrosion in the spacer grid and fretting corrosion by foreign objects that foul the coolant. Sizes of the cladding defects often reach some millimeters. Under the VVER typical conditions the secondary defects in the fuel rod claddings may arise in a time lesser than the fuel cycle duration. The failed fuel rods are characterized by increased diameter of cladding in the defect area induced by hydrogenation. For certain failed fuel rods the local and lengthy increases of cladding diameter ('inverse deformation') induced by mechanical impact of fuel were detected. The fuel loss {approx}10% was registered in the area of large cladding defects as a result of pellet abrasion by solid objects (spacer grid, foreign object) and small particle fallout in consequence of high fragmentation of the fuel pellets. In the area of small cladding defects as well as outside the defect area, the fuel loss was not found out (within the limits of measurement error {approx}5%). In the failed rod sections operated at the linear power less than {approx}150 W/cm (conservative estimate) the fuel structure is similar to one of the intact fuel rods. As for the failed rod sections operated at the linear power more than {approx}150 W/cm, the central regions of the pellets had increased equiaxed and un-equiaxed grains. In this case the volume of the recrystallized fuel sharply rises with increase of the fuel rod linear power. The Cs yield from the fuel pellets of the failed rods sharply rises due to the recrystallization at the linear power more than 150 W/cm. More than 80% of