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Sample records for ceramography

  1. Metallography and thermal analysis of ceramic nuclear fuels

    International Nuclear Information System (INIS)

    The book contains two parts: the ceramography laboratory and the thermal treatment laboratory. After general remarks on sintering the first part includes sample preparation for ceramography (grinding, polishing, etching), microscopic examination and quantitative image analysis. The second part deals with temperature measurement, oxide/metal ratio determination, thermogravimetry, differential thermal analysis (DTA), melting point determination and constitution of phase diagrams. Installation of a Pu laboratory, sample decontamination, and research with a microprobe are described. 188 photomicrographs present the microstructure of ceramics based on U, Pu and higher actinides

  2. Contribution to the study of the microstructure of uranium dioxide (1962)

    International Nuclear Information System (INIS)

    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)

  3. 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.)

  4. Measurement of krypton grain-boundary inventories in CANDU fuel

    International Nuclear Information System (INIS)

    A technique for measuring the Kr-85 grain-boundary inventory in irradiated fuel based on the conversion of UO2 to U3O8 at low temperatures has been improved. The improvements include: 1) the use of a tracer isotope to account for release from the matrix during measurement of the grain-boundary inventory and 2) the cutting of samples from known locations. With these improvements it is possible to measure radial variations in the grain-boundary inventory. The measurements of Kr-85 grain-boundary inventory can be combined with gamma mapping and ceramography to allow investigation of the connection between microstructure and fission-product distribution. (author)

  5. Status report March 1981 thermal-gradient-test LEU fuel

    International Nuclear Information System (INIS)

    Samples of candidate LEU fuels have been heated out-of-pile at approx. 15000C in a thermal gradient test. Silicon carbide failure in TRISO UO2 and TRISO UC2 was detected by x-radiography and by loss of Cs-137 after 163.5 hr. End-of-test ceramography showed that the SiC layer in the TRISO UC2 (6151-21-0111-5) had been corroded on both the hot and cold sides of the particles. No fission products were associated with the hot side corrosion, but Pd, U, Ru, Rh and rare-earth metals were detected in the corrosion zone on the cold side of the particle. Ceramography of the TRISO UO2 (6152-01-0111-3) showed that large voids had been formed in the SiC layer completely around the particle. Pd had concentrated at the interface between the inner PyC and the SiC. In addition, free Si had accumulated in the SiC. No SiC corrosion was observed in the TRISO UO2 (6152-03-0111-6). Palladium had accumulated at the SiC interface on the cool side of the particles and had diffused at least two-thirds of the way through the SiC without any visual SiC attack

  6. 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)

  7. Synthesis of UO2 and ThO2 doped with Gd2O3

    International Nuclear Information System (INIS)

    Uranium dioxide (urania, UO2) and thorium dioxide (thoria, ThO2) doped with gadolinium oxide (gadolinia, Gd2O3) were prepared via solid-state synthesis. For Gd2O3-doped ThO2, also an alternative, semi-dry process (“suspension coating”) was applied in which Gd2O3-coated ThO2 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 Gd2O3-doped UO2. However, this route was found to be inappropriate to obtain Gd2O3-doped ThO2 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. 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)

  9. Development of low decontaminated MOX fuel containing MA. 1. Influence of Np on sintering behavior and phase separation for (Pu, Np, U)O2-x

    International Nuclear Information System (INIS)

    MOX fuel containing Neptunium is being developed as candidate fuel for an advanced nuclear fuel recycle. In this report, influence of Np on the sintering behavior, phase separation behavior of MOX fuel pellets and the homogeneity of MOX fuel pellets were evaluated. It was observed that the high Np containing pellets had a low sintered density and the microstructure changes of the pellets during the sintering were slow compared with MOX without Np. The pellets were also analyzed via Ceramography, X-ray diffraction measurement and an electron probe microanalysis. The phase separation behavior of MOX with Np was similar to that of MOX. the homogeneity of the pellet produced with this experiment was acceptable to the fuel specification. (author)

  10. High-temperature transient fission-gas release from UO2 fuel

    International Nuclear Information System (INIS)

    Four of the in-reactor coolant transient tests performed at Chalk River Nuclear Laboratories have provided data on fission-gas release during dryout and LOCA conditions. On-line release measurements and fuel operating conditions have been used to deduce physical mechanisms involved in the transient release of fission products under these high-temperature conditions. Post-irradiation examination of the fuel from these four tests, including ceramography and scanning electron microscopy, are consistent with the on-line release data, and with the release mechanisms postulated from that data. In particular, the microstructural evidence confirms the presence of a stored fission gas inventory in these low burnup fuels, which is available for release by fuel cracking. Measured releases under these simulated accident conditions were less than 1.5% of inventory, indicating that transient fission-gas release during LOCA accidents with fuel temperatures up to 2000 degrees Celsius are not likely to be significant

  11. Study of the properties of the Am-O system in view of the transmutation of Am 241 in fast reactors

    International Nuclear Information System (INIS)

    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 AmO2-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 Am2O3-AmO2 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

  12. High burnup performance of Mg, Mg-Nb and Ti doped UO2 fuels

    International Nuclear Information System (INIS)

    In order to control irradiation performance of fuel swelling and FP gas release etc. at high burnups of light water reactor fuels, doped UO2 pellet fuels were prepared and their irradiation behavior was examined. The UO2 pellets doped 2.5 to 15mol%Mg, 5mol%Mg - 5mol%Nb, and 3.5mol%Ti and undoped UO2 pellets as a reference fuel were loaded together in a capsule and irradiated to the maximum burnups of 94GWd/t(U) below temperature of 1000degC in the JRR-3M reactor of JAERI. As results of post-irradiation examinations such as visual inspection, dimensional and density change measurements, thermal diffusivity and ceramography with optical microscope and EPMA, no difference was observed between the doped and the reference UO2 fuels. And valuable results were obtained on high burnup properties for swelling rates, thermal conductivities, structure changes and so on. (author)

  13. U3O8 microspheres sintering kinetics

    International Nuclear Information System (INIS)

    U3O8 microspheres sintering kinetics was determined using a hot-stage optical microscopy apparatus, able to reach temperature up to 13500C in controlled atmospheres. The sintered material had its microstructure analysed by optical and electron microscopy. The microspheres were characterized initialy utilizing X-ray diffractometry and thermogravimetry. The equation which describes the microspheres shrinkage in function of the time was obtained using finite difference analysis X-ray diffractometry indicated hexagonal structure for the microspheres main starting material, ammonium diuranate thermogravimetric analysis showed reduction of this material to U3O8 at 6000C. Ceramography results showed 5 hours sintered microspheres grain sizes G vary with the temperature. Sintered U3O8 micrographs compared with published results for UO2, indicate similar homogeneity microstructural characteristics and suggest the processed micorspheres to be potentially useful as nuclear fuels. (Author)

  14. 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)

  15. 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)

  16. 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

  17. Experiences from Refurbishment of Metallography Hot Cells and Application of a New Preparation Concept for Materialography Samples

    International Nuclear Information System (INIS)

    After more than 30 years of operation the lead shielded metallography hot cells needed a basic renewal and modernisation not least of the specimen preparation equipment. Preparation in hot cells of radioactive samples for metallography and ceramography is challenging and time consuming. It demands a special design and quality of all in-cell equipment and skill and patience from the operator. Essentials in the preparation process are: simplicity and reliability of the machines, and a good quality, reproducibility and efficiency in performance. Desirable is process automation, flexibility and an alara amounto of radioactive waste produced per sample prepared. State of the art preparation equipment for materialography seems to meet most of the demands, however, it cannot be used in hot cells without modifications. Therefore. IFE and Struers in Copenhagen modified a standard model of a Strues precision cutting machine and a microprocessor controlled grinding and polishing machine for Hot Cell application. Hot cell utilisation of the microcomputer controlled grinding and polishing machine and the existing automatic dosing equipment made the task of preparing radioactive samples more attractive. The new grinding and polishing system for hot cells provides good sample preparation quality and reproductibility at reduced preparation time and reduced amount of contaminated waste produced per sample prepared. the sample materials examined were irradiated cladding materials and fuels

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. 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)

  1. Microstructure Analysis of Thermally Etched Alumina Ceramics

    Directory of Open Access Journals (Sweden)

    Fudurić Jelača, M.

    2008-12-01

    Full Text Available Ceramography is the art and science of preparation, examination, and evaluation of ceramic microstructures. Microstructure is the structure level approximately 0.1 to 100 μ m between the wavelength of visible light and the resolution limit of the naked eye. The microstructure includes most grains, secondary phases, grain boundaries, pores, microcracks, hardness microindentations. Investigation and evaluation of ceramic microstructure is very important because a number of mechanical, optical, thermal, electrical and other properties of ceramics are significantly affected by the microstructure. The techniques for ceramographic preparation are divided into five parts: sawing, mounting, grinding, polishing and etching.In this paper a method for preparation of a cold isostatically pressed high purity alumina ceramics (α-Al2O3 is described. Microstructure analysis of prepared ceramics was performed by means of optical microscopy (OM, scanning electron microscopy (SEM and atomic force microscopy (AFM. Porosity is determined on the polished sample; grain size is measured after thermal etching. The mean grain diameter is determined by means of lineal-intercept method, circular-intercept method and image analysis.

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

    International Nuclear Information System (INIS)

    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

  3. Oxidation of spent fuel in air at 175 degree to 195 degree C

    International Nuclear Information System (INIS)

    Oxidation tests in dry air were conducted on four LWR spent fuels at 175 degrees and 195 degrees C to determine the effect of the fuel characteristics on the oxidation state likely to exist at the time leaching occurs in a potential repository. Weight changes were measured and samples were examined by XRD, ceramography, TEM, and TGA. Despite local variations in the grain boundary susceptibility to oxidation, all four fuels progressed toward an apparent endpoint at an oxygen-to-metal (O/M) ratio of 2.4. The sole oxidation product was U4O9+x, a cubic phase structurally related to UO2 but with a slightly smaller lattice constant. The growth of the U4O9+x from the grain boundaries into the UO2 grains followed parabolic kinetics and had an activation energy of 26.6 kcal/mol. Based on the kinetics, the time required at 95 degrees C to completely oxidize LWR spent fuel to U4O9+x would be at least 2000 yr. The next oxidation product to form after the U4O9+x phase may be U3O8, but no U3O8 or other dilatational oxidation product has been detected in these accelerated tests conducted up to 25,000 h

  4. 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.

  5. Utilization of 14 MW TRIGA research reactor integrated in a structure for materials and nuclear fuel characterization and development

    International Nuclear Information System (INIS)

    The Institute for Nuclear Research (ICN) of Pitesti has a set of nuclear facilities consisting of TRIGA 14 MW(th) materials testing reactor and LEPI (Romanian acronym for post-irradiation examination laboratory) which enable to investigate the behaviour of the nuclear fuel and materials under various irradiation conditions. The LEPI is an alpha-gamma hot cell facility able to manipulate and examine radioactive materials having an activity up to 106 Ci (Eaverage ≤ 1 MeV) and a high content of transuranium elements (Pu, Am, Cm). In order to obtain relevant information on CANDU nuclear fuel performance, a significant number of fuel elements manufactured by ICN has been tested to different power histories in the TRIGA 14 MW(th) reactor. Most important tests have been performed in conditions of power ramping, overpower and accident. After testing, the fuel elements have been examined in the hot cells at LEPI using various post-irradiation examination techniques. These techniques include both non-destructive methods (visual inspection and photography, eddy current testing, profilometry, gamma scanning) and destructive methods (fission gas release and analysis, matallography, ceramography, burnup determination by mass spectrometry, mechanical testings). The data obtained from post-irradiation examinations are used on one hand to confirm the integrity, safety and performance of the irradiated fuel and on the other hand for further progress in CANDU fuel development. (author)

  6. Characterization of spent fuel approved testing material: ATM-106

    International Nuclear Information System (INIS)

    The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory 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 (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) 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) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs

  7. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    International Nuclear Information System (INIS)

    To study MgAl2O4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241AmOx in spinel and 25 wt% micro-dispersed enriched UO2 in spinel. During irradiation, the initially present 241Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243Am (4%). The UO2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmOx inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling

  8. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    Science.gov (United States)

    Klaassen, F. C.; Bakker, K.; Schram, R. P. C.; Klein Meulekamp, R.; Conrad, R.; Somers, J.; Konings, R. J. M.

    2003-06-01

    To study MgAl 2O 4 spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed 241AmO x in spinel and 25 wt% micro-dispersed enriched UO 2 in spinel. During irradiation, the initially present 241Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and 243Am (4%). The UO 2 spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO x inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO 2 inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO 2 inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

  9. Post irradiation examination of irradiated americium oxide and uranium dioxide in magnesium aluminate spinel

    Energy Technology Data Exchange (ETDEWEB)

    Klaassen, F.C. E-mail: klaassen@nrg-nl.com; Bakker, K.; Schram, R.P.C.; Klein Meulekamp, R.; Conrad, R.; Somers, J.; Konings, R.J.M

    2003-06-01

    To study MgAl{sub 2}O{sub 4} spinel as inert matrix material for the transmutation of minor actinides, two capsules were irradiated at the high flux reactor in Petten, containing 12.5 wt% micro-dispersed {sup 241}AmO{sub x} in spinel and 25 wt% micro-dispersed enriched UO{sub 2} in spinel. During irradiation, the initially present {sup 241}Am was converted for 99.8% to fission products (50%), plutonium (30%), curium (16%) and {sup 243}Am (4%). The UO{sub 2} spinel target experienced a burn-up of 32% fission per initial metal atom. The post irradiation examination of the AmO{sub x} inert matrix target showed swelling of 27 vol.%, and a gas release of 48% for He and 16% for Xe and Kr. The UO{sub 2} inert matrix target also showed a large volumetric swelling of 11%, directed mainly radially. Ceramography on the UO{sub 2} inert matrix target revealed a complete restructuring of the spinel grains upon irradiation and the absence of porosity, suggesting that amorphisation is the main cause of the swelling.

  10. 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

  11. 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).

  12. 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)

  13. 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

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ionescu, S. [Institute for Nuclear Research Pitesti, Campului Str., 1, 115400 Mioveni (Romania)], E-mail: silviu.ionescu@nuclear.ro; Uta, O.; Parvan, M.; Ohai, D. [Institute for Nuclear Research Pitesti, Campului Str., 1, 115400 Mioveni (Romania)

    2009-03-31

    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.

  16. 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

  17. Design, irradiation and post-irradiation examination of the (U,Pu) C fuel pins of the test group FR 2-6d

    International Nuclear Information System (INIS)

    In the frame of the capsule group 6d three mixed carbide pins (pellet-density 92% T.D.) with a big radial gap (250 μm) were irradiated in the thermal neutron flux of FR 2. The cladding material consisted of the austenitic steel No. 1.4988. The exposure time in the reactor was up to 680 days, the burnup was 7.4 at %. The pins were instrumented on their surface with 6 thermocouples each. During irradiation in the NaK capsule no bigger irregularities in temperature readings were found. During dismantling in Karlsruhe Hot Cells the capsules it was found that all 3 pins showed cracks on their clads. The profilometry of the pins showed diameter increases from 4.5 to 6.0%. The carburization of the cladding proved the same tolerable magnitude as found for pins irradiated with moderate rod powers (Tsub(clad) 0C). Comparing ceramography with that of other pins of the same capsule group (KfK 2577) no bigger differences in structure were found. (orig.)

  18. In-reactor performance of prototype SBR MOX fuel

    International Nuclear Information System (INIS)

    As part of the international Callisto experiment, BNFL have undertaken the base irradiation, ramp testing and post-irradiation examination of two fuel rods, manufactured by BNFL's Short Binderless Route (SBR) for MOX fuel. Although of only short (lm) length, the rods were in other respects of standard PWR geometry. Base irradiation was performed in the Callisto loop of the BR2 reactor and achieved burn-ups of 17 GWd/tM, 28 GWd/tM peak pellet, at moderate to high power levels. No problems were encountered during base irradiation, confirmed by intermediate examination. In a separate BR2 capsule, a ramp test was then performed on one of the two rods. Again, the rod showed no problems and was discharged intact. The ramp test conditions employed were somewhat more onerous than the level corresponding to the best-estimate failure level for standard UO2 fuel. The survival of the rod confirms the general observation that the PCI failure resistance of MOX fuel is superior to that of standard UO2. The irradiation has been followed by a detailed programme of non-destructive and destructive hot-cell examinations on both of the rods. Profilometry, ECT and gamma-scanning confirmed the overall satisfactory performance of the rods and the absence of incipient damage. Puncture testing and gas analysis showed the fission gas release levels in both the ramped and unramped rods to be in line with what would be expected for a UO2 rod. The satisfactory microstructure of the fuel was confirmed via optical ceramography and autoradiography. This paper will describe the fuel manufacture, base irradiation and ramp test conditions, and will provide a summary of the results of the PIE programme. (author)

  19. Investigation on uranium and plutonium nitrides with low oxygen and carbon contents

    International Nuclear Information System (INIS)

    Uranium nitride (UN) and uranium-plutonium nitride (UO.8Pu0.2N) with various oxygen impurity levels, up to 20.000 10-6 weight ratio, was studied. The strong affinity of these nitrides for the oxygen avoids to synthesize pure compounds (no oxygen) by direct combination of the elements or from hydride. The process expected to be used in nuclear fuel industry was chosen. The nitride was prepared by carboreduction and nitridation of the oxide, then ground with different amounts of oxide. The powder obtained was cold pressed and sintered (T = 17200C - 18000C ; t > 15 hours). Analysis of carbon and oxygen content, X ray diffraction measurements, ceramography and electronprobe microanalysis were used to characterize the pellets. The main results are: The oxide (UO2 or MO2) forms at temperatures higher than about 11500C, an oxinitride in contact with nitride matrix (UN or MN), only under nitrogen. This oxinitride, isomorphous with UO2 crystal, is stable up to 17500C with nitride matrix, under a pressure of 1 bar. During the cooling the oxinitride is decomposed in UO2 and U2N3+x. This mixed oxinitride of U and Pu was observed for the first time. The plutonium content of this solid solution is twice smaller than in the nitride matrix. The solubility limit of oxygen in the UN and U0.8Pu0.2N is less than 1000.10-6 weight ratio. This value is lower than published results. The lattice parameter of UN increases in ratio with carbon content, but no noticeable influence of oxygen was detected. This lattice parameter, for UN saturated with oxygen, is 0.48887 ± 5.10-5 nm

  20. 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)

  1. Improvements in PIE-techniques at the IFE hot-laboratory. 'Neutron radiography, three dimensional profilometry and image compilation of PIE data for visualization in an image based user-interface'

    International Nuclear Information System (INIS)

    The PIE-techniques used at IFE are continuously improved through upgrading of equipment and methods, e.g. image handling techniques and components utilized in data acquisition and editing techniques. To improve the quality or spatial resolution of neutron radiographs the normal technique was complemented with another method, i.e. the dysprosium foil/X ray film technique is supplemented with a track-etch recorder consisting of a cellulose nitrate film. For further examination of the neutron radiographs the cellulose nitrate film can be digitized to allow electronic image treatment. Promising results were obtained with this technique on neutron radiographs, namely higher spatial resolution compared to the normal technique, high contrast and sharp neutron radiography images. The traditional uniaxial profilometry of fuel rods was modified so that diameter/bow measurements are possible at several angular orientations during one acquisition sequence. This extension is very useful in several ways, for instance the built-in data symmetry of the method is used to check the correctness of the measurement results. Diameter and bow measurements give information of cladding irregularities and fuel rod profiles. Implementation of electronic image handling techniques is particularly useful in PIE when data are collected and compiled in an image file. Inspection and examination of the file contents (examination results) are possible through an ideal user-interface, i.e. Adobe Photoshop software with navigator possibilities. Examples incorporating PIE data acquired from neutron radiography, visual inspection and ceramography are utilized for illustration of the user-interface and some of its possibilities. (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. 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

  4. Fabrication and characterisation of composite targets for the transmutation of actinides

    International Nuclear Information System (INIS)

    Transmutation of transuranic elements separated from spent fuel is a way to reduce the toxicity of long-lived nuclides in the waste before disposal. Plutonium and the minor actinides (MA) are reintroduced into the fuel cycle for further irradiation and incineration. Currently CERMET fuel forms, in which a ceramic actinide is dispersed in a matrix, are considered for MA transmutation. In a first step, PuO2 beads are produced by a sol gel method in which a Pu nitrate solution is converted to solid, dust-free, particles. These porous beads are then infiltrated with an americium nitrate solution to the incipient wetness point and calcined to give the (PuAm)O2 beads, which are blended with a metal matrix and compacted and sintered to form the final fuel pellet. The matrix used is molybdenum due to its high thermal conductivity and low neutron capture cross section, if it is enriched in 92Mo. In this work, optimization of the bead porosity is investigated to achieve a higher Am content by infiltration. Addition of carbon to the mother solution in the sol gel step increases the bead porosity but it also changes both bead and final fuel pellet microstructure. A surrogate fuel, with cerium simulating the actinides has been fabricated and its mechanical stability and bead characteristics investigated as a function of carbon content and thermal treatment. The characterization of the surrogate fuel by ceramography, density, porosity, bead-quality, etc., is a necessary step in the process optimization, to be transferred to the production of the actinide samples. This process is now at an advanced stage and is being used for the production of fuels for irradiation tests in the Phenix (Futurix) and HFR-Petten (HELIOS) reactors. In parallel, studies on the dissolution of the fuel pellets, with the aim of dissolving the Mo-matrix while keeping the CeO2 beads intact, have been initiated. Thus, Mo can be recycled for further fuel fabrication either from production scraps or from the

  5. 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

  6. Overview of Post-Irradiation Examination Techniques Applied at PSI for Light Water Reactor Fuel Characterization

    International Nuclear Information System (INIS)

    Within long term cooperation agreements, the PSI Laboratory for Material Behaviour (LWV) in the Department for Nuclear Energy and Safety (NES) has characterized pathfinder PWR- and BWR fuel (meaning fuel and cladding) of the Swiss nuclear power stations Goesgen (KKG) and Leibstadt (KKL) in many PIE sequences. Based on these pathfinder fuel pin and lead assembly tests the fuels in these reactors has been continuously improved over the years and reaches today a batch burnup which is nearly twice as high as at the beginning of 1980. Additionally to providing scientific analytical services with respect to accurate engineering fuel and cladding PIE data, PSI itself as a research organization undertakes basic and applied scientific research. We focus in this environment in elucidating the cladding corrosion and hydriding mechanisms, the cladding mechanical aging processes and the fission gas diffusion process in the fuel. The PIE tools available at PSI consist of non-destructive methods (visual examination, gamma scanning, profilometry and EC-defect and -oxide thickness measurements), puncturing and fission gas analysis, and destructive investigations of cut samples (metallography/ceramography, hydrogen hot gas extraction, electron probe micro-analysis (EPMA), secondary ion mass spectroscopy (SIMS), scanning- and transmission- electron microscopy (SEM and TEM), lately also laser ablation inductively coupled plasma mass spectroscopy (LA-ICPMS), x-ray absorption spectroscopy, and mechanical testing). While commercial high burnup programs often request standard engineering data like cladding oxide thickness values and hydrogen contents or fuel pin fission gas release values, PSI has improved such analytical techniques. Additionally, we have performed independent research in order to improve the fundamental understanding of the irradiation behaviour e.g. by elucidating the corrosion process with detailed characterization of the cladding metal-oxide interface by TEM, by local

  7. 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

  8. IFPE/US-PWR-16 X 16 Lead Test Assembly Extended Burnup Demonstration Program

    International Nuclear Information System (INIS)

    Description: US-PWR 16 x 16 LTA (lead test assembly) extended burnup demonstration program conducted during the 1980's. Relevant program data was obtained from the project final report and other supporting documents. The objective of this program was to demonstrate improved nuclear fuel utilization through more efficient fuel management and increased discharge burnup. The use of the 16 x 16 LTAs with Zr-4 cladding in this program demonstrated the capability to achieve peak fuel rod average burnups of ∼ 60 GWd/MTU. Both pool side (non-destructive) and hot cell (destructive) post irradiation examinations (PIE) of selected rods from the two LTAs were conducted. These examinations included rods irradiated for 3 and 5 cycles. Pool side examinations of the LTAs included visual inspection, dimensional measurements, eddy currant testing (ECT), and waterside corrosion thickness measurement. Hot cell fuel rod PIE included void volume measurements, fill gas analyses, cladding visual inspections, dimensional measurements, neutron radiography, and gamma scanning. Fuel pellet examinations included fuel densification and swelling measurements, fuel burnup analyses, and ceramography. Cladding examinations included metallography, hydrogen concentration measurement, and mechanical property testing. The irradiation of two 16 x 16 LTAs was completed in a US commercial PWR. LTA D039 was irradiated during reactor cycles 2 through 4. The irradiation of LTA D040 was extended through reactor cycle 6 to achieve a lead rod, axial average burnup of 58 GWd/MTU. The fuel assembly design consisted of 236 rods in a 16 x 16 array, five control element guide tubes, 12 fuel rod spacer grids, upper and lower end fittings, and a hold-down device. The bottom spacer grid is Inconel 625. All other spacer grids and all guide tubes are Zr-4. The standard fuel rod design consists of enriched UO2, solid cylindrical pellets, a round wire Type 302 stainless steel compression spring, and an alumina spacer