WorldWideScience

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. Determination of porosity of pyrocarbon by means of the automatic quantitative image analysis

    International Nuclear Information System (INIS)

    For a long time, the quantitative image analysis is well known as a method for quantifying the results of material investigation basing on ceramography. The development of the automatic image analysers has made it a fast and elegant procedure for evaluation. It is used to determine easily and routinely the macroporosity and by this the density of the pyrocarbon coatings of nuclear fuel particles. This report describes the definition of measuring parameters, the measuring procedure, the mathematical calculations, and first experimental and mathematical results. (orig.)

  3. Contribution to the study of the microstructure of uranium dioxide (1962); Contribution a l'etude de la microstructure du dioxyde d'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Porneuf, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-05-15

    The microstructure of sintered uranium dioxide is studied in relation with several parameters, specially the sintering temperatures and atmospheres. The external surface and the internal microstructure of the sintered are examined, using fractography and ceramography. Various techniques for preparing surfaces (mechanical and electrolytic polishing) and for revealing the structure (chemical and anodic attack, ionic bombardment oxidation) have been experienced and compared. Patterns similar to those revealed in metals and probably related with interactions between dislocations and vacancies have been observed. (author) [French] La microstructure de frittes d'oxyde d'uranium est etudiee en fonction de divers parametres, en particulier de la temperature et de l'atmosphere de frittage, par examen de la surface externe des frittes, puis de leur microstructure interne (fractographie, ceramographie). Differentes techniques de preparation des surfaces (polissage mecanique ou electrolytique) et de revelation de la structure (attaque chimique ou anodique, bombardement ionique, oxydation preferentielle) ont ete experimentees et comparees. Des figures comparables a celles revelees dans les metaux et liees probablement a des interactions entre dislocations et lacunes ont ete observees. (auteur)

  4. 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)

  5. Irradiation behavior of FBTR mixed carbide fuel at various burn-ups

    International Nuclear Information System (INIS)

    The fast breeder test reactor at Kalpakkam has completed nearly 25 years of operation and is now operating at 18 MWt capacity with 46 fuel subassemblies (FSA) in the core consisting of 27 Mark-I (70% PuC + 30% UC), 13 Mark-II (55% PuC + 45% UC) and 6 MOX (44% PuO2 + 56% UO2) and one test PFBR FSA. Post Irradiation Examination (PIE) campaigns on FSAs at different burnup levels has provided valuable information about the irradiation behavior of the carbide fuel. This paper gives a summary of the irradiation performance of the carbide fuel evaluated through some of the investigations such as neutron radiography, x-radiography, gamma scanning, fission gas analysis and ceramography. Burnup of the carbide fuel could be enhanced from the initial design burnup limit of 50 GWd/t to 165 GWd/through systematic PIE. (author)

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

    International Nuclear Information System (INIS)

    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)

  7. Fission-product behaviour during irradiation of TRISO-coated particles in the HFREU1bis experiment - HTR2008-58125

    International Nuclear Information System (INIS)

    The irradiation experiment HFR-EU1bis, coordinated by the European Joint Research Centre - Inst. for Energy, was performed in the High Flux Reactor (HFR) at Petten to test five spherical HTR fuel pebbles of former German production with TRISO coated particles in conditions beyond the specifications of current HTR reactor designs (central temperature of 1250 deg. C). In this paper, the behaviour of the fission products (FPs) and kernel micro-structure evolution during the test are investigated. While FP behaviour is a key issue for potential source term evaluation it also determines the evolution of the oxygen potential in the oxide kernel which in turn is important for formation of carbon oxides (amoeba effect and pressurization). Fission-gas release from the kernel can induce additional mechanical loading and finally some FPs (Ag, Cs, Sr) might alter the mechanical integrity of the coatings. This study is based on post- irradiation examinations (ceramography + EPMA) performed both on UO2 kernels and on coatings. Significant evolutions of the kernel as a function of temperature are shown (grain structure, porosity, size of metallic inclusions). The quality of the ceramography results allows characteristics of the intergranular bubbles in the kernel (and estimation of swelling) to be determined. Remarkable results considering FP release from the kernel have been observed and will be presented. Examples are the significant release of Cs out of the kernel as well as Pd, whereas Zr remains trapped. Mo and Ru are mainly incorporated in metallic precipitates. These observations are interpreted and mechanisms for FP and micro-structural evolutions are proposed. These results are coupled to the results of calculations performed with the mechanistic code MFPR (Module for Fission Product Release) and the thermodynamic database MEPHISTA (Multiphase Equilibria in Fuels via Standard Thermodynamic Analysis). The effect of high flux rate and high temperature on fission gas

  8. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    Science.gov (United States)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  9. Operation Procedure of Inspection Equipment for TRISO-coated Fuel Particle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Kim, Y. K.; Cho, M. S.; Kim, Y. M.; Park, J. Y.; Kim, W. J.; Jeong, K. C.; Oh, S. C.; Lee, Y. W

    2007-03-15

    TRISO-coated fuel particle for HTGR(high temperature gas cooled reactor) is composed of fuel kernel and coating layers. The kernel and coated particle are characterized by inspection processes for inspection items such as diameter of kernel, thickness, density and an-isotropy of coating layer. The coating thickness can be nondestructively measured by X-ray inspection equipment. The coating thickness as well as the sphericity can be also measured by optical inspection system as a ceramography method. The an-isotropy can be characterized by photometer. The density of coating layer can be measured by density column. The size and sphericity of particles can be measured by PSA(particle size analyzer). The thermo-chemical characteristics of kernel can be analyzed by TG/DTA(Thermogravimetric/Differential Thermal Analyzer). The inspection objective, equipment composition, operation principle, operation manual for each equipment was described in this operation procedure, which will be used for the characterization of inspection items described above.

  10. 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.

  11. 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.

  12. Post irradiation examination of thoria-plutonia mixed oxide fuel in Indian hot cells

    International Nuclear Information System (INIS)

    Mixed oxide (MOX) fuel clusters containing ThO2+4%PuO2, and ThO2+6.75%PuO2 fuel pins were irradiated in the pressurized water loop of the Indian research reactor CIRUS, to burn up in the range of 20 GWd/T(HM). The ThO2+4%PuO2 fuel elements had free standing cladding made of Zircaloy-2 and the ThO2+6.75%PuO2 had collapsible Zircaloy-2 cladding. The fuel clusters had performed well during irradiation with no apparent indications of failure. The techniques used for the post irradiation examination (PIE) of these fuels in the hot cells included visual examination, fuel pin diameter measurements, leak testing, gamma scanning, gamma spectrometry, ultrasonic testing, eddy current testing, ceramography, metallography, beta gamma autoradiography and measurement of released fission gases. Micro hardness measurement of cladding and evaluation of mechanical properties using ring tension test were also carried out. This paper elaborates on the techniques and the results of the PIE carried out on ThO2+4%PuO2 fuel. (author)

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

    International Nuclear Information System (INIS)

    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

  14. Plutonium doping of SYNROC-D

    International Nuclear Information System (INIS)

    The purpose of this work was to perform an experimental simulation of the radiation effects that SYNROC-D (a ceramic waste form and the alternate to borosilicate glass for US defense high-level nuclear waste) will experience during the first million years of storage. Technology was developed for doping SYNROC-D with 238Pu and performing external gamma irradiation to simulate both actinide and fission product decay. The doping technique was tested using both Ce and U as stand-ins to simulate the +3 and +4 oxidation states of Pu, respectively. Samples were characterized by ceramography, density measurements, x-ray diffraction, scanning electron microscope-energy dispersive x-ray analysis, electron microprobe, scanning transmission electron microscope, gamma ray spectrometry, and leaching; equipment was fabricated for dilatation measurements. An early decision by the Department of Energy (DOE) to select borosilicate glass and terminate SYNROC-D development prevented doping with 238Pu or external gamma irradiation. However, a sample was doped with 239Pu in order to study the Pu distribution, and characterization of this sample was completed. Although conclusive proof was not developed, all indications from this work are that Pu will go into the zirconolite and perovskite phases in SYNROC-D, favoring perovskite under the redox conditions prevailing in a graphite die. Technology development and results of Ce, U, and 239Pu doping studies are described in this report

  15. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Kikuchi, Hironobu; Fukuda, Kousaku; Kikuchi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saruta, Tohru; Kitajima, Toshio

    1994-10-01

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of {sup 88}Kr, was relatively high, 1.5x10{sup -5}, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of {sup 88}Kr for the 12th assembly was reduced to an excellent value of 2x10{sup -6}. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author).

  16. Irradiation experiments of the 6th-12th OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    The Oarai Gas Loop-1, OGL-1, is an in-pile helium gas loop, installed in the Japan Materials Testing Reactor (JMTR), for irradiation of high-temperature gas-cooled reactor fuels at high pressure and temperature. The present report describes results of fabrication, irradiation and post-irradiation examinations (PIE) of the 6th-12th OGL-1 fuel assemblies. The 6th-8th assemblies used coated fuel particles produced by a small-scale fluidized bed. On the other hand, the 9th-12th assemblies used trial manufacturing fuels, produced with a large-scale fluidized bed for mass production of the fuel for the High Temperature Engineering Test Reactor (HTTR) being constructed. For the 9th assembly loaded with the first mass-product fuel, the fission gas release, R/B of 88Kr, was relatively high, 1.5x10-5, and various defects were observed in the ceramography of the irradiated coating layers. Afterwards, a decrease was achieved in the through-coating failure fractions at the fabrication. Correspondingly, the R/B of 88Kr for the 12th assembly was reduced to an excellent value of 2x10-6. Thus, the production technology and the irradiation performance of the HTTR design fuels were successfully demonstrated. (author)

  17. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Postirradiation heating tests of TRISO-coated UO2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85Kr, 110mAg, 134Cs, 137Cs, and 154Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110mAg, 134Cs, 137Cs, and 154Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

  18. 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.

  19. 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.

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

    International Nuclear Information System (INIS)

    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

  1. Suitability of a thermal design method for FBR oxide fuel rods

    International Nuclear Information System (INIS)

    To study suitability of the thermal design method for fast breeder reactor (FBR) oxide fuel rods, the B14 irradiation test with four fuel rods was carried out in the experimental fast reactor 'JOYO'. Pellet-cladding gap width and O/M ratio of oxide fuels were specified as experimental parameters. In addition, by taking into account the actual design conditions for FBR oxide fuel, the conditions in the B14 irradiation test, i.e. linear power and cladding temperature, were planned to include the hottest design conditions. The maximum Pu content and the maximum Am content of fuel pellet were 31 wt% and 2.4 wt%, respectively. The B14 fuel rods were irradiated with the maximum linear power of ∼ 47 kW/m in the test. After irradiation, ceramography samples were taken from the axial position of each fuel rod where the fuel centerline temperature reached the maximum during irradiation. The result was that the influences of both pellet-cladding gap width and O/M ratio on the fuel restructuring were observed, but the fuel melting was not observed. In addition, thermal analysis code 'DIRAD' would be suitable to evaluate the thermal behavior of oxide fuels containing several percent Am, from the result of verification by using the result of the B14 irradiation test. Moreover, from the computation result of DIRAD, the power to melt for the B14 oxide fuels was evaluated as 55-57 kW/m. It could be mentioned that the margin of the conventional oxide fuel design would be at least 10 kW/m at the transient. Consequently, the margin to the criterion in the thermal design would be suitable and the fuel melting would be prevented under the conditions designated in the conventional FBR design.(author)

  2. Post irradiation examination of thoria-plutonia MOX fuel

    International Nuclear Information System (INIS)

    Thoria based mixed oxide (MOX) fuel pin clusters containing (Th-4% Pu)O2 fuel pins were irradiated in the pressurised water loop (PWL) of the research reactor CIRUS to a burnup of 18.5 GWd/Te. The fuel pins had fuel pellets, encapsulated in free standing cladding made of Zircaloy-2. The fuel pin clusters had performed well during irradiation, with no apparent indications of failure. Non-destructive and destructive post irradiation examinations were carried out which involved visual examination, fuel pin diameter measurement, leak testing, ultrasonic testing, eddy current testing, gamma spectroscopy and gamma scanning, measurement of released fission gases, ceramography, metallography, alpha autoradiography, beta-gamma autoradiography and microhardness measurements. No abnormality was observed on the surface of the fuel pins during visual examination and there was no significant change in the dimensions of the fuel pins. Ultrasonic testing of fuel pin cladding did not reveal any defect. The spectra obtained from gamma scanning of the fuel pins revealed the presence of 137Cs, 134Cs, 154Eu- and 208Tl. Results of the metallographic examination using optical metallography of transverse sections of the fuel pins and SEM studies on fractured fuel grain surfaces has provided valuable information on fuel cracking, grain morphology and distribution of porosity and fission gas bubbles in the fuel. During eddy current testing, a defect signal was obtained from surface of the cladding of one of the fuel pins. Leak testing of the fuel pin by liquid nitrogen-alcohol leak test did not reveal any leaks. During the measurement of released fission gases, the pin held vacuum and the plenum gas did not contain any fission gas. On subsequent sectioning and metallography of the cladding at the defect location, formation of massive Zirconium hydride was observed. This paper elaborates the irradiation performance of Thoria-Plutonia MOX fuel and also enumerates the probable reasons of

  3. LWR MOX fuel irradiation tests. HBWR irradiation with the instrument rig, IFA-514/565

    International Nuclear Information System (INIS)

    IFA-514 irradiation test was performed in Halden Reactor (HBWR) in Norway to study the irradiation performance of LWR MOX fuels. The fuel specifications for this irradiation test were decided in accordance with those of BWR 8x8 fuels, and plutonium content of MOX fuels was set to be 5.8 wt.%. Six MOX fuel rods, of which parameters were pellet geometry (solid or annular) and surface roughness (grinded or as-sintered), were irradiated to the assembly average burn-up of ∼45 GWd/t, and the instrument data during irradiation, i.e. cladding elongation, fuel stack elongation, fuel center temperature, and rod inner pressure, were obtained, and subsequent post-irradiation examinations provided the following results: No remarkable corrosion and deformation was observed on the irradiated fuel rods. The averaged irradiation growth was 0.13%, no nodular corrosion was observed, and the maximum thickness of oxide layer was ∼60 μm. The FP gas release rate was ∼21%. The irradiation for three of six fuel rods irradiated in IFA-514 irradiation tests were continued to the assembly average burn-up of ∼56 GWd/t in IFA-565 irradiation tests, and the results of this irradiation test were as follows: No remarkable corrosion and deformation was observed on the irradiated fuel rods. The FP gas release behavior of LWR MOX fuels was similar to that of BWR UO2 and ATR MOX fuels, and no difference was confirmed in the FP gas release behavior. Also FP gas release rate of annular pellets (∼13%) was lower than that of solid ones (∼16%). No remarkable difference in the effect of pellet geometry, i.e. solid and annular, on PCMI behavior was observed in any fuel rod from the instrument data of fuel stack elongation and the results of ceramography. However, the reduction of cladding diameter change in pellet geometry is expected since the cladding diameter change of annular fuel rods was less than that of solid ones. The disappearance of as-fabricated granular boundaries during irradiation

  4. Experience of Integrated Safeguards Approach for Large-scale Hot Cell Laboratory

    International Nuclear Information System (INIS)

    The Japan Atomic Energy Agency (JAEA) has been operating a large-scale hot cell laboratory, the Fuels Monitoring Facility (FMF), located near the experimental fast reactor Joyo at the Oarai Research and Development Center (JNC-2 site). The FMF conducts post irradiation examinations (PIE) of fuel assemblies irradiated in Joyo. The assemblies are disassembled and non-destructive examinations, such as X-ray computed tomography tests, are carried out. Some of the fuel pins are cut into specimens and destructive examinations, such as ceramography and X-ray micro analyses, are performed. Following PIE, the tested material, in the form of a pin or segments, is shipped back to a Joyo spent fuel pond. In some cases, after reassembly of the examined irradiated fuel pins is completed, the fuel assemblies are shipped back to Joyo for further irradiation. For the IAEA to apply the integrated safeguards approach (ISA) to the FMF, a new verification system on material shipping and receiving process between Joyo and the FMF has been established by the IAEA under technical collaboration among the Japan Safeguard Office (JSGO) of MEXT, the Nuclear Material Control Center (NMCC) and the JAEA. The main concept of receipt/shipment verification under the ISA for JNC-2 site is as follows: under the IS, the FMF is treated as a Joyo-associated facility in terms of its safeguards system because it deals with the same spent fuels. Verification of the material shipping and receiving process between Joyo and the FMF can only be applied to the declared transport routes and transport casks. The verification of the nuclear material contained in the cask is performed with the method of gross defect at the time of short notice random interim inspections (RIIs) by measuring the surface neutron dose rate of the cask, filled with water to reduce radiation. The JAEA performed a series of preliminary tests with the IAEA, the JSGO and the NMCC, and confirmed from the standpoint of the operator that this

  5. 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