A Characterization Research of UO2 Powder for UO2 Pellet Fabrication of Candu Type

International Nuclear Information System (INIS)

Rachmawati, M.

1998-01-01

A characterization research of of UO 2 powder for UO 2 pellet fabrication of Candu type is reported in this paper. The research has been conducted by characterizing sinterability, compactibility, and compressibility of UO 2 (Cameco) without a pre-compacting and UO 2 powder the result of a pre-compacting. The pre-compacting UO 2 powder has been done to have particle size to less than 150 mu (150-800) mu, and more than 800 mu with distribution varied. Sinterability of each group of particle sizes is analyzed using Thermogravimetric-Differential Thermal Analysis (TG-DTA). Then the final compacting to the powder is done using compaction pressure varied from 1 MP to 4 MP to the all groups of the particle sizes to find the optimum pressure by measuring the density and mechanical strength of the UO 2 green pellet. Both measurements are performed using Micrometer and Universal Testing Machine respectively. The result of this investigation shows that the group of UO 2 powder with no pre-compacting with particle size of less than 150 mu with 60% distribution and (150-800) mu size with 40% distribution are the UO 2 pellets which are eligible in terms of their density and mechanical strength

Heat conductance of sintered UO{sub 2}; Toplotna provodljivost sinterovanog UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Katanic-Popovic, J; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

Phenomena influencing the heat conductance of the sintered UO{sub 2} were analyzed, first of all when used as nuclear fuel. Influence of temperature, density and porosity, additives and irradiation in the reactor are shown. Based on the available literature, the measured heat conductance values were analyzed for the sintered UO{sub 2} outside the reactor and in the reactor during irradiation. Analizirane su pojave koje uticu na toplotnu provodljivost sinterovanog UO{sub 2}, pre svega, sa aspekta njegove primene kao goriva. Izlozen je uticaj temperature, gustine i poroznosti, aditiva i ozracivanja u reaktoru. Na osnovu pregleda dostupne literature kriticki su prikazani rezultati merenja toplotne provodljivosti sinterovanog UO{sub 2} van reaktora i u reaktoru pri ozracivanju (author)

Etching of UO2 in NF3 RF Plasma Glow Discharge

International Nuclear Information System (INIS)

John M. Veilleux

1999-01-01

A series of room temperature, low pressure (10.8 to 40 Pa), low power (25 to 210 W) RF plasma glow discharge experiments with UO 2 were conducted to demonstrate that plasma treatment is a viable method for decontaminating UO 2 from stainless steel substrates. Experiments were conducted using NF 3 gas to decontaminate depleted uranium dioxide from stainless-steel substrates. Depleted UO 2 samples each containing 129.4 Bq were prepared from 100 microliter solutions of uranyl nitrate hexahydrate solution. The amorphous UO 2 in the samples had a relatively low density of 4.8 gm/cm 3 . Counting of the depleted UO 2 on the substrate following plasma immersion was performed using liquid scintillation counting with alpha/beta discrimination due to the presence of confounding beta emitting daughter products, 234 Th and 234 Pa. The alpha emission peak from each sample was integrated using a gaussian and first order polynomial fit to improve quantification. The uncertainties in the experimental measurement of the etched material were estimated at about ± 2%. Results demonstrated that UO 2 can be completely removed from stainless-steel substrates after several minutes processing at under 200 W. At 180 W and 32.7 Pa gas pressure, over 99% of all UO 2 in the samples was removed in just 17 minutes. The initial etch rate in the experiments ranged from 0.2 to 7.4 microm/min. Etching increased with the plasma absorbed power and feed gas pressure in the range of 10.8 to 40 Pa. A different pressure effect on UO 2 etching was also noted below 50 W in which etching increased up to a maximum pressure, approximately23 Pa, then decreased with further increases in pressure

Thermal conductivity of the sintered UO{sub 2}; Toplotna provodljivost sinterovanog UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Katanic-Popovic, J; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1967-04-15

Phenomena influencing the thermal conductivity of the sintered UO{sub 2} fuel were analyzed. Influence of temperature, density and porosity, additives and irradiation in the reactor core are presented. Thermal conductivity of sintered UO{sub 2} was measured both outside the reactor and during the irradiation in the reactor. Results are discussed and analyzed based on the available literature. Analizirane su pojave koje uticu na toplotnu provodljivost sinterovanog UO{sub 2}, pre svega, sa aspekta njegove primene kao goriva. Izlozen je uticaj temperature, gustine i poroznosti, aditiva i ozracivanja u reaktoru. Na osnovu pregleda dostupne literature kriticki su prikazani rezultati merenja toplotne provodljivosti sinterovanog UO{sub 2} van reaktora i u reaktoru pri ozracivanju (author)

Modeling of UO2 aqueous dissolution over a wide range of conditions

International Nuclear Information System (INIS)

Steward, S.A.; Weed, H.C.

1993-11-01

Previously it was not possible to predict reliably the rate at which spent fuel would react with groundwater because of conflicting data in the literature. The dissolution of the UO 2 spent fuel matrix is a necessary step for aqueous release of radioactive fission products. Statistical experimental design was used to plan a set of UO 2 dissolution experiments to examine systematically the effects of temperature (25--75C), dissolved oxygen (0.002--0.2 atm overpressure), pH (8--10) and carbonate (2-200x10 -4 molar) concentrations on UO 2 dissolution. The average uranium dissolution rate was 4.3 mg/m 2 /day. The regression fit of the data indicate an Arrhenius type activation energy of 8750 cal/mol and a half-power dependence on dissolved oxygen in the simulated groundwater

Irradiation of UO2

International Nuclear Information System (INIS)

Stevanovic, M.

1965-10-01

Based on the review of the available literature concerned with UO 2 irradiation, this paper describes and explains the phenomena initiated by irradiation of the UO 2 fuel in a reactor dependent on the burnup level and temperature. A comprehensive review of UO 2 radiation damage studies is given as a broad research program. This part includes the abilities of our reactor as well as needed elements for such study. The third part includes the definitions of the specific power, burnup level and temperature in the center of the fuel element needed for planning and performing the irradiation. Methods for calculating these parameters are included [sr

Development of ceramics based fuel, Phase I, Kinetics of UO{sub 2} sintering by vibration compacting of UO{sub 2} powder (Introductory report); Razvoj goriva na bazi keramike, I faza, Kinetika sinterovanja UO{sub 2} vibraciono kompaktiranje praha UO{sub 2} (Uvodni izvestaj)

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

After completing the Phase I of the task related to development of ceramics nuclear fuel the following reports are presented: Kinetics of UO{sub 2} sintering; Vibrational compacting and sintering of UO{sub 2}; Characterisation of of UO{sub 2} powder by DDK and TGA methods; Separation of UO{sub 2} powder.

Separation of UO{sub 2} powder; Separacija praha UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

This report deals with theoretical approach to separation process and describes the constructed separator with liquid medium. The separator was calibrated and tested with Al{sub 3}O{sub 3} and UO{sub 2}. it has been concluded that it can be used for separation of powders with sufficient accuracy if the separation is performed for a longer period of time. The separated fractions were characterised by microscopic method and the UO{sub 2} fraction additionally by sedimentation method.

Kinetics of UO{sub 2} sintering; Kinetika sinterovanja UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-10-15

Detailed conclusions related to the UO{sub 2} sintering can be drawn from investigating the kinetics of the sintering process. This report gives an thorough analysis of the the data concerned with sintering available in the literature taking into account the Jander and Arrhenius laws. This analysis completes the study of influence of the O/U ratio and the atmosphere on the sintering. Results presented are fundamentals of future theoretical and experimental work related to characterisation of the UO{sub 2} sintering process.

Cation interdiffusion in the UO2 - (U, Pu)O2 and UO2 - PuO2 systems

International Nuclear Information System (INIS)

Leme, D.G.

1985-01-01

The interdiffusion of U and Pu ions in UO sub(2 +- x) - (U sub(0,83) Pu sub(0,17))O sub(2 + - x) and UO sub(2 + - x) -PuO sub(2 - x) sintered pellets and UO sub(2 +- x) -(U sub(0,82) Pu sub(0,18))O sub(2 + - x) single crystals has been studied as a function of the oxygen potential ΔG sup(-) (O 2 ) or the stoichiometric ratio O/M. The diffusion profiles of UO 2 /(U,Pu)O 2 and UO 2 /PuO 2 couples of different O/M ratios have been measured using high resolution α-spectrometer and microprobe. Thermal annealing of the specimens was performed in controlled atmospheres using either CO-CO 2 gas mixtures for constant O/M ratios or purified argon. The interdiffusion profiles have been analysed by means of the Boltzmann-Matano and Hall methods. The interdiffusion coefficient D sus(approx.) increases with increasing Pu content in sintered pellets (up to 17 wt. %PuO 2 ) showing a strong dependence of D sup(approx.) on the O/M ratio. The micropobe results show that the interdiffusion along grain boundaries is the main diffusion mechanism in the pellets. Experiments have also been carried out in single cristals to measure just the bulk-interdiffusion and avoiding effects due to grain boundaries. A marked dependence of D sup(approx.) on O/M ratio or on oxygen potential ΔG sup(-) (O 2 ), similar to the dependence already reported for self diffusion by means of radioactive tracers, has also been observed. (Author) [pt

Inter-atomic interaction between electrons, 2

International Nuclear Information System (INIS)

Haga, Eijiro; Kato, Tomohiko; Aisaka, Tsuyoshi.

1978-01-01

Intra- and inter-atomic interactions in the exchange process are defined with respect to the Wannier function rather than the atomic function. In relation to the neutron scattering data for nickel, the behavior for the effective exchange parameter I(q) in the q-dependent susceptibility is, in RPA, investigated by taking into account the main types of the nearest neighbor interactions and by extending our previous treatment. The different types of interactions lead to different behavior for the q-dependence of I(q). The contribution to I(q) from inter-atomic interactions other than the exchange type decreases as the surface area of the Fermi surface becomes large. For the exchange type, the l-th neighbor interaction with l<=4 is taken into account, and, from the comparison with the empirical result for I(q), it is found that the inter-atomic contribution to I(0) is about thirty percent with a reasonable decrease against l. (author)

Phonon density of states and anharmonicity of UO2

Science.gov (United States)

Pang, Judy W. L.; Chernatynskiy, Aleksandr; Larson, Bennett C.; Buyers, William J. L.; Abernathy, Douglas L.; McClellan, Kenneth J.; Phillpot, Simon R.

2014-03-01

Phonon density of states (PDOS) measurements have been performed on polycrystalline UO2 at 295 and 1200 K using time-of-flight inelastic neutron scattering to investigate the impact of anharmonicity on the vibrational spectra and to benchmark ab initio PDOS simulations performed on this strongly correlated Mott insulator. Time-of-flight PDOS measurements include anharmonic linewidth broadening, inherently, and the factor of ˜7 enhancement of the oxygen spectrum relative to the uranium component by the increased neutron sensitivity to the oxygen-dominated optical phonon modes. The first-principles simulations of quasiharmonic PDOS spectra were neutron weighted and anharmonicity was introduced in an approximate way by convolution with wave-vector-weighted averages over our previously measured phonon linewidths for UO2, which are provided in numerical form. Comparisons between the PDOS measurements and the simulations show reasonable agreement overall, but they also reveal important areas of disagreement for both high and low temperatures. The discrepancies stem largely from a ˜10 meV compression in the overall bandwidth (energy range) of the oxygen-dominated optical phonons in the simulations. A similar linewidth-convoluted comparison performed with the PDOS spectrum of Dolling et al. obtained by shell-model fitting to their historical phonon dispersion measurements shows excellent agreement with the time-of-flight PDOS measurements reported here. In contrast, we show by comparisons of spectra in linewidth-convoluted form that recent first-principles simulations for UO2 fail to account for the PDOS spectrum determined from the measurements of Dolling et al. These results demonstrate PDOS measurements to be stringent tests for ab inito simulations of phonon physics in UO2 and they indicate further the need for advances in theory to address the lattice dynamics of UO2.

Densification Behavior of BN-added UO2

International Nuclear Information System (INIS)

Rhee, Young Woo; Kim, Keonsik; Kim, Dong Joo; Kim, Jong Hun; Oh, Jang Soo; Yang, Jae Ho

2013-01-01

Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study a pulsed eddy current (PEC) technology to detect the wall thing of carbon steel pipe covered with insulation is developed. Boron is commercially used as a neutron absorber fuel. A neutron absorber fuel is burned out or depleted during reactor operation. Westinghouse have been produced the Integral Fuel Burnable Absorber (IFBA) which is enriched UO 2 fuel pellets with a thin coating of zirconium diboride (ZrB 2 ) on the outer surface. Standard sintered fuel pellets are sputter coated with ZrB 2 . It is known that IFBA fuel can incur 20% to 30% additional fabrication costs. Boron-dispersed UO 2 fuel pellet made by the conventional pressing and sintering process of a powder mixture of UO 2 and B compound might be more cost-effective than IFBAs. M. G. Andrew et al. tried to sinter boron-dispersed UO 2 green pellet. However, they reported that boron-dispersed UO 2 fuel pellet is very difficult to be fabricated with a sufficient level of boron retention and high sintered density (greater than 90 % of theoretical density) because of the volatilization of boron oxide. We have investigated the densification behavior of mixtures of UO 2 and various boron compounds, such as B 4 C, BN, TiB 2 , ZrB 2 , SiB 6 , and HfB 2 . Boron compounds seemed to act as a sintering additive for UO 2 at a certain low temperature range. In this study, the densification behavior of BN-added UO 2 pellet has been investigated by sintering green pellets of a mixture of UO 2 powder and BN powder in H 2 atmosphere. A high density BN-added UO 2 pellet can be fabricated after sintering at 1200 .deg. C for more than 1 h in a H 2 atmosphere. The sintered density of BN-added UO 2 pellet can be increased up to about 95 %TD

Characterization of UO2 by infrared spectroscopy

International Nuclear Information System (INIS)

Faeda, Kelly C.M.; Machado, Geraldo C.; Lameiras, Fernando S.

2011-01-01

The characterization of nuclear fuel is of great importance to minimize the effects related to burnup and temperature and to achieve stability during in-core operation. The understanding the U-O system and its thermodynamic properties has fundamental importance in nuclear industry. Many physical properties of UO 2±x depend on the ratio O / U, such as the electrical conductivity and thermal properties, as well as the diffusivities of its constituents and solutes. The U-O system presents various oxides such as UO 2±x , U 4 O 9 , U 3 O 8 , and UO 3 . The control of the O/U relation is critical to the manufacturing process of UO 2 . In this work, the infrared spectroscopy was used to identify the presence of phases in UO 2 powder samples that cannot be identified by thermogravimetry and X-ray diffraction. (author)

Study of UO2 radioinduced densification

International Nuclear Information System (INIS)

Stora, J.P.; Bruet, M.

1975-01-01

Measurements of radioinduced densification were performed on UO 2 DCN (intergranular fine porosity) and UO 2 DCI (interaggregate coarse porosity) in the Anemone device. The densification kinetics was followed by measuring the shrinkage of the oxide column on neutron radiographic plates. UO 2 DCI was found stable in regard to densification. At power near 450Wcm -1 , densification is hitten by restructuring phenomena [fr

Migration behavior of palladium in UO2, (3)

International Nuclear Information System (INIS)

Yoneyama, Mitsuru; Sato, Seichi; Ohashi, Hiroshi; Ogawa, Toru; Ito, Akinori; Fukuda, Kousaku.

1992-08-01

Palladium (Pd) is easily released from UO 2 kernels in HTGR coated fuel particles, and reacts with SiC coating layer. In addition, Pd is one of metallic fission products in irradiation UO 2 , which constitutes in dissoluble residue in reprocessing of LWR fuels. In the present investigation, the migration of palladium in UO 2 was examined by heating diffusion pairs sandwiched Pd foil between UO 2 wafers at 1300 ∼ 1800degC. Experiments were also carried out on affinity of Pd to UP 2 and a formation of U-Pd alloy. Pd was found mainly in the pores of UO 2 . The maximum depth intruded by Pd in fairly large amount was more than 100 μm for UO 2 with 90%TD and 50μm for UO 2 with 95%TD, while the maximum length of open pores was 330 μm for UO 2 with 90%TD, and 50 m for that with 95%TD. Fused Pd wetted UO 2 very much. Pd intruded deeply into UO 2 , especially in the edge of Pd droplet. Furthermore, U was detected either in precipitates or the Pd source with α-Pd phase of U-Pd alloy containing Pd at about 10at%. This fact indicates that Pd highly reacts with UO 2 . From the above results, the transport of Pd in UO 2 was explained by the model of gaseous diffusion through pores in UO 2 , which is retarded by formation of U-Pd alloy. It is also indicated that UPd 3 forms even at the oxygen potential condition of O/U ratio, which is a little higher than 2.00 on the basis of thermodynamic calculation. (author)

Parameterization of interatomic potential by genetic algorithms: A case study

Energy Technology Data Exchange (ETDEWEB)

Ghosh, Partha S., E-mail: psghosh@barc.gov.in; Arya, A.; Dey, G. K. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai-400085 (India); Ranawat, Y. S. [Department of Ceramic Engineering, Indian Institute of Technology (BHU), Varanasi-221005 (India)

2015-06-24

A framework for Genetic Algorithm based methodology is developed to systematically obtain and optimize parameters for interatomic force field functions for MD simulations by fitting to a reference data base. This methodology is applied to the fitting of ThO{sub 2} (CaF{sub 2} prototype) – a representative of ceramic based potential fuel for nuclear applications. The resulting GA optimized parameterization of ThO{sub 2} is able to capture basic structural, mechanical, thermo-physical properties and also describes defect structures within the permissible range.

New UO2 fuel studies

International Nuclear Information System (INIS)

Dehaudt, P.; Lemaignan, C.; Caillot, L.; Mocellin, A.; Eminet, G.

1998-01-01

With improved UO 2 fuels, compared with the current PWR, one would enable to: retain the fission products, rise higher burn-ups and deliver the designed power in reactor for longer times, limit the pellet cladding interaction effects by easier deformation at high temperatures. Specific studies are made in each field to understand the basic mechanisms responsible for these improvements. Four programs on new UO 2 fuels are underway in the laboratory: advanced microstructure fuels (doped fuels), fuels containing Er 2 O 3 a burnable absorber, fuels with improved caesium retention, composite fuels. The advanced microstructure UO 2 fuels have special features such as: high grain sizes to lengthen the fission gas diffusion paths, intragranular precipitates as fission gas atoms pinning sites, intergranular silica based viscoplastic phases to improve the creep properties. The grain size growth can be obtained with a long time annealing or with corundum type oxide additives partly soluble in the UO 2 lattice. The amount of doping element compared with its solubility limit and the sintering conditions allows to obtain oxide or metallic precipitates. The fuels containing Er 2 O 3 as a burnable absorber are under irradiation in the TANOX device at the present time. Specific sintering conditions are required to improve the erbium solubility in UO 2 and to reach standard or large grain sizes. The improved caesium retention fuels are doped with SiO 2 +A1 2 O 3 or SiO 2 +ZrO 2 additives which may form stable compounds with the Cs element in accidental conditions. The composite fuels are made of UO 2 particles of about 100 μm in size dispersed in a molybdenum metallic (CERMET) or MgA1 2 O 4 ceramic (CERCER) matrix. The CERMET has a considerably higher thermal conductivity and remains ''cold'' during irradiation. The concept of double barrier (matrix+fuel) against fission products is verified for the CERMET fuel. A thermal analysis of all the irradiated rods shows that the thermal

Development of ceramics based fuel, Phase I, Kinetics of UO2 sintering by vibration compacting of UO2 powder (Introductory report)

International Nuclear Information System (INIS)

Ristic, M.M.

1962-10-01

After completing the Phase I of the task related to development of ceramics nuclear fuel the following reports are presented: Kinetics of UO 2 sintering; Vibrational compacting and sintering of UO 2 ; Characterisation of of UO 2 powder by DDK and TGA methods; Separation of UO 2 powder

Summary report on UO₂ thermal conductivity model refinement and assessment studies

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lashley, Jason Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Byler, Darrin David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bell, B. D.C. [Imperial College, London (United Kingdom); Grimes, R. W. [Imperial College, London (United Kingdom); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-02-03

Uranium dioxide (UO₂) is the most commonly used fuel in light water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, therefore, governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models were replaced with models that incorporate explicit thermal conductivity degradation mechanisms during fuel burn-up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO₂ thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham type interatomic potential and a potential that combines the many-body embedded atom method potential with Morse-Buckingham pair potentials. Potential parameters for UO_2+x and ZrO₂ are developed for the latter potential. Physical insights from the resonant phonon-spin scattering mechanism due to spins on the magnetic uranium ions have been introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high

Thermal expansion of UO2-Gd2O3 fuel pellets

International Nuclear Information System (INIS)

Une, Katsumi

1986-01-01

In recent years, more consideration has been given to the application of UO 2 -Gd 2 O 3 burnable poison fuel to LWRs in order to improve the core physics and to extend the burnup. It has been known that UO 2 forms a single phase cubic fluorite type solid solution with Gd 2 O 3 up to 20 - 30 wt.% above 1300 K. The addition of Gd 2 O 3 to UO 2 lattices changes the properties of the fuel pellets. The limited data on the thermal expansion of UO 2 -Gd 2 O 3 fuel exist, but those are inconsistent. UO 2 -Gd 2 O 3 fuel pellets were fabricated, and the linear thermal expansion of UO 2 and UO 2 -(5, 8 and 10 wt.%)Gd 2 O 3 fuel pellets was measured with a differential dilatometer over the temperature range of 298 - 1973 K. A sapphire rod of 6 mm diameter and 15.5 mm length was used as the reference material. After the preheating cycle, the measurement was performed in argon atmosphere. The results for UO 2 pellets showed excellent agreement with the data in literatures. The linear thermal expansion of UO 2 -Gd 2 O 3 fuel pellets showed the increase with increasing the Gd 2 O 3 content. Consideration must be given to this excessive expansion in the fuel design of UO 2 -Gd 2 O 3 pellets. The equations for the linear thermal expansion and density of UO 2 -Gd 2 O 3 fuel pellets were derived by the method of least squares. (Kako, I.)

Thermodynamic Behaviour of Hypostoichiometric UO{sub 2}; Comportement Thermodynamique de UO{sub 2} HypostoeChiometrique; Termodinamicheskoe povedenie gipostekhiometricheskoj UO{sub 2}; Comportamiento Termodinamico del UO{sub 2} Subestequiometrico

Energy Technology Data Exchange (ETDEWEB)

Aitken, E. A.; Brassfield, H. C.; Fryxell, R. E. [General Electric Company, Nuclear Materials and Propulsion Operation, Cincinnati, OH (United States)

1966-02-15

The ability of the UO{sub 2}-type structure to accomodate excess oxygen is well known. Recent evidence has indicated that this structure is stable also in the hypostoichiometric state at high temperatures and low oxygen partial pressures, but its manifestation occurs as a uranium metal precipitate in the oxide after cooling from high temperatures. This paper presents further evidence of the existence, at high temperatures, of a stable hypostoichiometric urania and describes in part the variation in thermodynamic properties across its homogeneity range. Hypostoichiometric UO{sub 2} evaporates congruently during free vaporization in slowly flowing hydrogen (-40 Degree-Sign C dew point) at 2400 Degree-Sign C at a composition having oxygen-to-uranium ratio of 1.88. If the temperature is decreased or the moisture content (oxygen partial pressure) increased, the congruent composition increases. The water content of the hydrogen at 2400 Degree-Sign C must be at least one per cent to maintain stoichiometric uranium dioxide. When UO{sub 2} pellets are sealed in tantalum cans and heated above 1700 Degree-Sign C, the O/U ratio of the pellet changes and reaches an equilibrium value which is governed by the oxygen activity of the atmosphere surrounding the can. UO{sub 2} does not react with tantalum but, because of the high solubility of oxygen in tantalum, the latter functions as a membrane. Using the data from congruent evaporation, and tantalum capsule tests, conducted in various argon-hydrogen mixtures, the oxygen activity in urania as a function of stoichiometry has been determined. The partial molar free energy of oxygen, G(O{sub 2} ), increases almost linearly on the oxygen deficient side with increasing oxygen-to-uranium ratio. Near the stoichiometric composition G(O{sub 2}) rises steeply. Using these results together with estimated G(O{sub 2}) values on the oxygen excess side obtained from the literature, it is shown that the data at a given temperature are consistent

Measurements of the viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 liquid solutions

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Roth, A.

1983-01-01

Adding UO 2 produces an increase of viscosity of borax and sodium metaborate. For temperatures below 920 0 C the measurements with the borax-UO 2 solution show a phase separation. Contrary to borax the sodium metaborate solutions indicate a well defined melting point. At temperatures slightly below the melting point a solid phase is formed. The tested sodium-borates-UO 2 mixtures are in liquid form. (DG)

Tracer surface diffusion on UO2

International Nuclear Information System (INIS)

Zhou, S.Y.; Olander, D.R.

1983-06-01

Surface diffusion on UO 2 was measured by the spreading of U-234 tracer on the surface of a duplex diffusion couple consisting of wafers of depleted and enriched UO 2 joined by a bond of uranium metal

UO2 pellet and manufacturing method

International Nuclear Information System (INIS)

Komada, Kiichi; Nishinaka, Keiji; Adachi, Kazunori; Fujiwara, Shuji.

1995-01-01

The present invention concerns an uranium dioxide pellet having a large crystal grain size. The grain size of the pellet is enlarged to increase the distance of an FP gas generated in the crystal grain to reach the grain boundary and, as a result, decrease the releasing speed of the FP gas. A UO 2 powder having a specific surface area of from 5 to 50m 2 /g is used as a starting powder in a step of forming a molding product, and chlorine or a chlorine compound is added in such an amount that the chlorine content in the UO 2 pellet is from 3 to 25ppm, in one of a production step, a molding step or a sintering step for UO 2 powder. With such procedures, a UO 2 pellet having a large crystal grain size can be prepared with good reproducibility. (T.M.)

Thermal conductivity and thermal diffusivity of solid UO2

International Nuclear Information System (INIS)

Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

1981-06-01

New equations for the thermal conductivity of solid UO 2 were derived based upon a nonlinear least squares fit of the data available in the literature. In the development of these equations, consideration was given to their thermodynamic consistency with heat capacity and density and theoretical consistency with enthalpy and heat capacity. Consistent with our previous treatment of enthalpy and heat capacity, 2670 K was selected as the temperature of a phase transition. A nonlinear equation, whose terms represent contributions due to phonons and electrons, was selected for the temperature region below 2670 K. Above 2670 K, the data were fit by a linear equation

Study of UO{sub 2}F{sub 2} - H{sub 2}O - HF compounds; Etude des composes UO{sub 2}F{sub 2} - H{sub 2}O - HF

Energy Technology Data Exchange (ETDEWEB)

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

1961-07-01

We study various compounds resulting from the interaction of UO{sub 2}F{sub 2} with H{sub 2}O and HF (gas), and various triple compounds UO{sub 2}F{sub 2} - H{sub 2}O - HF; the conditions of decomposition and the thermodynamic limits of stability are specified. (author) [French] Nous etudions divers composes formes par reaction de UO{sub 2}F{sub 2} avec H{sub 2}O et HF (gaz) et divers composes triples UO{sub 2}F{sub 2} - H{sub 2}O - HF, en essayant de preciser les decompositions et domaines d'exisfence thermodynamiques de ces corps. (auteur)

Dissolution of UO2 in redox conditions

International Nuclear Information System (INIS)

Casas, I.; Pablo de, J.; Rovira, M.

1998-01-01

The performance assessment of the final disposal of the spent nuclear fuel in geological formations is strongly dependent on the spent fuel matrix dissolution. Unirradiated uranium (IV) dioxide has shown to be very useful for such purposes. The stability of UO 2 is very dependent on vault redox conditions. At reducing conditions, which are expected in deep groundwaters, the dissolution of the UO 2 -matrix can be explained in terms of solubility, while under oxidizing conditions, the UO 2 is thermodynamically unstable and the dissolution is kinetically controlled. In this report the parameters which affect the uranium solubility under reducing conditions, basically pH and redox potential are discussed. Under oxidizing conditions, UO 2 dissolution rate equations as a function of pH, carbonate concentration and oxidant concentration are reported. Dissolution experiments performed with spent fuel are also reviewed. The experimental equations presented in this work, have been used to model independent dissolution experiments performed with both unirradiated and irradiated UO 2 . (Author)

Photochemical synthesis of UO2 nanoparticles

International Nuclear Information System (INIS)

Rath, M.C.; Keny, Sangeeta; Naik, D.B.

2014-01-01

UO 2 nanoparticles have been recently synthesized by us from aqueous solutions of uranyl nitrate through radiolytic method on high-energy electron beam irradiation. In this study, the synthesis of UO 2 nanoparticles through photochemical method is reported which is a complementary route to radiation chemical method

Etching of UO₂ in NF₃ RF Plasma Glow Discharge

Energy Technology Data Exchange (ETDEWEB)

Veilleux, John M. [Univ. of California, Berkeley, CA (United States)

1999-08-01

A series of room temperature, low pressure (10.8 to 40 Pa), low power (25 to 210 W) RF plasma glow discharge experiments with UO₂ were conducted to demonstrate that plasma treatment is a viable method for decontaminating UO₂ from stainless steel substrates. Experiments were conducted using NF₃ gas to decontaminate depleted uranium dioxide from stainless-steel substrates. Depleted UO₂ samples each containing 129.4 Bq were prepared from 100 microliter solutions of uranyl nitrate hexahydrate solution. The amorphous UO₂ in the samples had a relatively low density of 4.8 gm/cm₃. Counting of the depleted UO₂ on the substrate following plasma immersion was performed using liquid scintillation counting with alpha/beta discrimination due to the presence of confounding beta emitting daughter products, ²³⁴Th and ²³⁴Pa. The alpha emission peak from each sample was integrated using a gaussian and first order polynomial fit to improve quantification. The uncertainties in the experimental measurement of the etched material were estimated at about ± 2%. Results demonstrated that UO₂ can be completely removed from stainless-steel substrates after several minutes processing at under 200 W. At 180 W and 32.7 Pa gas pressure, over 99% of all UO₂ in the samples was removed in just 17 minutes. The initial etch rate in the experiments ranged from 0.2 to 7.4 μm/min. Etching increased with the plasma absorbed power and feed gas pressure in the range of 10.8 to 40 Pa. A different pressure effect on UO₂ etching was also noted below 50 W in which etching increased up to a maximum pressure, ~23 Pa, then decreased with further increases in pressure.

Spent fuel UO2 matrix corrosion behaviour studies through alpha-doped UO2 pellets leaching

International Nuclear Information System (INIS)

Muzeau, B.; Jegou, C.; Broudic, V.

2005-01-01

Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO 2 matrix in aqueous media subjected to α-β-γ radiations. The β-γ emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO 2 matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO 2 matrix, 238/239 Pu doped UO 2 pellets (0.22 %wt. Pu total) were fabricated with different 238 Pu/ 239 Pu ratio to reproduce the alpha activity of a 47 GWd.t HMi -1 UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO 2 pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO 3 1 mM), under Argon (O 2 2 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO 2 batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry (HCO 3 - , pH, Eh,..), the atmosphere (Ar, Ar/H 2 ,..), and the radiolysis strength. The experimental matrix

The heating of UO_2 kernels in argon gas medium on the physical properties of sintered UO_2 kernels

International Nuclear Information System (INIS)

Damunir; Sri Rinanti Susilowati; Ariyani Kusuma Dewi

2015-01-01

The heating of UO_2 kernels in argon gas medium on the physical properties of sinter UO_2 kernels was conducted. The heated of the UO_2 kernels was conducted in a sinter reactor of a bed type. The sample used was the UO_2 kernels resulted from the reduction results at 800 °C temperature for 3 hours that had the density of 8.13 g/cm"3; porosity of 0.26; O/U ratio of 2.05; diameter of 1146 μm and sphericity of 1.05. The sample was put into a sinter reactor, then it was vacuumed by flowing the argon gas at 180 mmHg pressure to drain the air from the reactor. After that, the cooling water and argon gas were continuously flowed with the pressure of 5 mPa with 1.5 liter/minutes velocity. The reactor temperature was increased and variated at 1200-1500 °C temperature and for 1-4 hours. The sinters UO_2 kernels resulted from the study were analyzed in term of their physical properties including the density, porosity, diameter, sphericity, and specific surface area. The density was analyzed using pycnometer with CCl_4 solution. The porosity was determined using Haynes equation. The diameters and sphericity were showed using the Dino-lite microscope. The specific surface area was determined using surface area meter Nova-1000. The obtained products showed the the heating of UO_2 kernel in argon gas medium were influenced on the physical properties of sinters UO_2 kernel. The condition of best relatively at 1400 °C temperature and 2 hours time. The product resulted from the study was relatively at its best when heating was conducted at 1400 °C temperature and 2 hours time, produced sinters UO_2 kernel with density of 10.14 gr/ml; porosity of 7 %; diameters of 893 μm; sphericity of 1.07 and specific surface area of 4.68 m"2/g with solidify shrinkage of 22 %. (author)

Separation of UO2 powder

International Nuclear Information System (INIS)

Ristic, M.M.

1962-01-01

This report deals with theoretical approach to separation process and describes the constructed separator with liquid medium. The separator was calibrated and tested with Al 3 O 3 and UO 2 . it has been concluded that it can be used for separation of powders with sufficient accuracy if the separation is performed for a longer period of time. The separated fractions were characterised by microscopic method and the UO 2 fraction additionally by sedimentation method

First identification and thermodynamic characterization of the ternary U(VI) species, UO2(O2)(CO3)2(4-), in UO2-H2O2-K2CO3 solutions.

Science.gov (United States)

Goff, George S; Brodnax, Lia F; Cisneros, Michael R; Peper, Shane M; Field, Stephanie E; Scott, Brian L; Runde, Wolfgang H

2008-03-17

In alkaline carbonate solutions, hydrogen peroxide can selectively replace one of the carbonate ligands in UO2(CO3)3(4-) to form the ternary mixed U(VI) peroxo-carbonato species UO2(O2)(CO3)2(4-). Orange rectangular plates of K4[UO2(CO3)2(O2)].H2O were isolated and characterized by single crystal X-ray diffraction studies. Crystallographic data: monoclinic, space group P2(1)/ n, a = 6.9670(14) A, b = 9.2158(10) A, c = 18.052(4) A, Z = 4. Spectrophotometric titrations with H 2O 2 were performed in 0.5 M K 2CO 3, with UO2(O2)(CO3)2(4-) concentrations ranging from 0.1 to 0.55 mM. The molar absorptivities (M(-1) cm(-1)) for UO2(CO3)3(4-) and UO2(O2)(CO3)2(4-) were determined to be 23.3 +/- 0.3 at 448.5 nm and 1022.7 +/- 19.0 at 347.5 nm, respectively. Stoichiometric analyses coupled with spectroscopic comparisons between solution and solid state indicate that the stable solution species is UO2(O2)(CO3)2(4-), which has an apparent formation constant of log K' = 4.70 +/- 0.02 relative to the tris-carbonato complex.

Oxidation kinetic changes of UO2 by additive addition and irradiation

International Nuclear Information System (INIS)

You, Gil-Sung; Kim, Keon-Sik; Min, Duck-Kee; Ro, Seung-Gy

2000-01-01

The kinetic changes of air-oxidation of UO 2 by additive addition and irradiation were investigated. Several kinds of specimens, such as unirradiated-UO 2 , simulated-UO 2 for spent PWR fuel (SIMFUEL), unirradiated-Gd-doped UO 2 , irradiated-UO 2 and -Gd-doped UO 2 , were used for these experiments. The oxidation results represented that the kinetic patterns among those samples are remarkably different. It was also revealed that the oxidation kinetics of irradiated-UO 2 seems to be more similar to that of unirradiated-Gd-doped UO 2 than that of SIMFUEL

The compaction and sintering of UO_2-Zr cermet pellets

International Nuclear Information System (INIS)

Tri Yulianto; Meniek Rachmawati; Etty Mutiara

2013-01-01

An innovative fuel pellet of UO_2-Zr cermet has been developed to improve thermal conductivity of UO_2 pellet by adding small amount Zr metal in to UO_2 matrix below 10 % weight. Zirconium powder will serve for the creation of bridges or web structure during compaction and will effectively reduce contact between of UO_2 particles. Based on the theory of phase equilibrium of metals-metal oxides-ceramic, this fabrication technique may produce UO_2 pellets containing continuous metal channel on the grain boundary of UO_2 through sintering in a reduction atmosphere. The fabrication was done by varying process parameters of mixing and compaction. Characterisation of UO_2-Zr cermet pellet involved visual test, dimensional and density measurement, and ceramography test. This advanced cermet fabrication technology may address common issue with cermet fuels such as microstructure with continuous metal channel structure in the UO_2 matrix, which is more effectively than the commonly accepted microstructure involving fraction of UO_2 pellet by standard fabrication route. (author)

Fabrication of nano-structured UO2 fuel pellets

International Nuclear Information System (INIS)

Yang, Jae Ho; Kang, Ki Won; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Heon; Kim, Keon Sik; Song, Kun Woo

2007-01-01

Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

Perovskite phases in the systems AO-SE/sub 2/O/sub 3/-UO/sub 2,x/ with A=alkaline earth metal and SE=rare earths, La, and Y. VII. The systems Ba/sub 2/CaUO/sub 6/-Ba/sub 2/Gd/sub 0. 67/UO/sub 6/ and Ba/sub 2/CaUO/sub 6/-Ba/sub 2/Y/sub 0. 67/UO/sub 6/

Energy Technology Data Exchange (ETDEWEB)

Kemmler-Sack, S; Seemann, I; Schittenhelm, H J [Tuebingen Univ. (F.R. Germany). Institut fuer Anorganische Chemie

1976-05-01

The ordered perovskite Ba/sub 2/CaUO/sub 6/ forms a solid solution series with Ba/sub 2/Gdsub(0.67)UO/sub 6/ and Ba/sub 2/Ysub(0.67)UO/sub 6/, respectively. The deviations from the ideal behaviour are studied by X-ray, diffuse reflectance and vibrational methods.

Perovskite phases in the systems AO-SE/sub 2/O/sub 3/-UO/sub 2,x/ with A=alkaline earth metal and SE=rare earths, La, and Y. IX. The systems Ba/sub 2/SrUO/sub 6/-Ba/sub 2/Gd/sub 0. 67/UO/sub 6/ and Ba/sub 2/SrUO/sub 6/-Ba/sub 2/Y/sub 0. 67/UO/sub 6/

Energy Technology Data Exchange (ETDEWEB)

Kemmler-Sack, S; Seemann, I [Tuebingen Univ. (F.R. Germany). Inst. fuer Anorganische Chemie I

1976-07-01

The ordered perovskite Ba/sub 2/SrUO/sub 6/ forms a solid solution series with Ba/sub 2/Gdsub(0.67)UO/sub 6/ and Ba/sub 2/Ysub(0.67)UO/sub 6/ respectively. The deviations from the ideal behaviour are studied by X-ray, diffuse reflectance and vibrational methods.

High density UO2 powder preparation for HWR fuel

International Nuclear Information System (INIS)

Hwang, S. T.; Chang, I. S.; Choi, Y. D.; Cho, B. R.; Kwon, S. W.; Kim, B. H.; Moon, B. H.; Kim, S. D.; Phyu, K. M.; Lee, K. A.

1992-01-01

The objective of this project is to study on the preparation of method high density UO 2 powder for HWR Fuel. Accordingly, it is necessary to character ize the AUC processed UO 2 powder and to search method for the preparation of high density UO 2 powder for HWR Fuel. Therefore, it is expected that the results of this study can effect the producing of AUC processed UO 2 powder having sinterability. (Author)

Kinetics of UO2 sintering

International Nuclear Information System (INIS)

Ristic, M.M.

1962-01-01

Detailed conclusions related to the UO 2 sintering can be drawn from investigating the kinetics of the sintering process. This report gives an thorough analysis of the the data concerned with sintering available in the literature taking into account the Jander and Arrhenius laws. This analysis completes the study of influence of the O/U ratio and the atmosphere on the sintering. Results presented are fundamentals of future theoretical and experimental work related to characterisation of the UO 2 sintering process

Formation of ternary CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) complexes under neutral to weakly alkaline conditions.

Science.gov (United States)

Lee, Jun-Yeop; Yun, Jong-Il

2013-07-21

The chemical behavior of ternary Ca-UO2-CO3 complexes was investigated by using time-resolved laser fluorescence spectroscopy (TRLFS) in combination with EDTA complexation at pH 7-9. A novel TRLFS revealed two distinct fluorescence lifetimes of 12.7 ± 0.2 ns and 29.2 ± 0.4 ns for uranyl complexes which were formed increasingly dependent upon the calcium ion concentration, even though nearly indistinguishable fluorescence peak shapes and positions were measured for both Ca-UO2-CO3 complexes. For identifying the stoichiometric number of complexed calcium ions, slope analysis in terms of relative fluorescence intensity versus calcium concentration was employed in a combination with the complexation reaction of CaEDTA(2-) by adding EDTA. The formation of CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) was identified under given conditions and their formation constants were determined at I = 0.1 M Na/HClO4 medium, and extrapolated to infinitely dilute solution using specific ion interaction theory (SIT). As a result, the formation constants for CaUO2(CO3)3(2-) and Ca2UO2(CO3)3(aq) were found to be log β113(0) = 27.27 ± 0.14 and log β213(0) = 29.81 ± 0.19, respectively, providing that the ternary Ca-UO2-CO3 complexes were predominant uranium(vi) species at neutral to weakly alkaline pH in the presence of Ca(2+) and CO3(2-) ions.

Development of irradiated UO2 thermal conductivity model

International Nuclear Information System (INIS)

Lee, Chan Bock; Bang Je-Geon; Kim Dae Ho; Jung Youn Ho

2001-01-01

Thermal conductivity model of the irradiated UO 2 pellet was developed, based upon the thermal diffusivity data of the irradiated UO 2 pellet measured during thermal cycling. The model predicts the thermal conductivity by multiplying such separate correction factors as solid fission products, gaseous fission products, radiation damage and porosity. The developed model was validated by comparison with the variation of the measured thermal diffusivity data during thermal cycling and prediction of other UO 2 thermal conductivity models. Since the developed model considers the effect of gaseous fission products as a separate factor, it can predict variation of thermal conductivity in the rim region of high burnup UO 2 pellet where the fission gases in the matrix are precipitated into bubbles, indicating that decrease of thermal conductivity by bubble precipitation in rim region would be significantly compensated by the enhancing effect of fission gas depletion in the UO 2 matrix. (author)

Microstructure study of AUC and UO2

International Nuclear Information System (INIS)

Pan Ying; Gao Dihua; Lu Huaichang

1992-01-01

The microstructures of AUC, UO 2 powder and pellets were investigated with metallo-scope, SEM, TEM, XRD, and image analyzer. The influence of the reduction conditions of AUC on the microstructures of UO 2 powder and pellet were studied

Formation of (Cr, Al)UO{sub 4} from doped UO{sub 2} and its influence on partition of soluble fission products

Energy Technology Data Exchange (ETDEWEB)

Cooper, M.W.D. [Department of Materials, Imperial College London, London (United Kingdom); Gregg, D.J.; Zhang, Y.; Thorogood, G.J.; Lumpkin, G.R. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia); Grimes, R.W. [Department of Materials, Imperial College London, London (United Kingdom); Middleburgh, S.C., E-mail: simm@ansto.gov.au [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales (Australia)

2013-11-15

CrUO{sub 4} and (Cr, Al)UO{sub 4} have been fabricated by a sol–gel method, studied using diffraction techniques and modelled using empirical pair potentials. Cr{sub 2}O{sub 3} was predicted to preferentially form CrUO{sub 4} over entering solution into hyper-stoichiometric UO{sub 2+x} by atomic scale simulation. Further, it was predicted that the formation of CrUO{sub 4} can proceed by removing excess oxygen from the UO{sub 2} lattice. Attempts to synthesise AlUO{sub 4} failed, instead forming U{sub 3}O{sub 8} and Al{sub 2}O{sub 3}. X-ray diffraction confirmed the structure of CrUO{sub 4} and identifies the existence of a (Cr, Al)UO{sub 4} phase for the first time (with a maximum Al to Cr mole ratio of 1:3). Simulation was subsequently used to predict the partition energies for the removal of fission products or fuel additives from hyper-stoichiometric UO{sub 2+x} and their incorporation into the secondary phase. The partition energies are consistent only with smaller cations (e.g. Zr{sup 4+}, Mo{sup 4+} and Fe{sup 3+}) residing in CrUO{sub 4}, while all divalent cations are predicted to remain in UO{sub 2+x}. Additions of Al had little effect on partition behaviour. The reduction of UO{sub 2+x} due to the formation of CrUO{sub 4} has important implications for the solution limits of other fission products as many species are less soluble in UO{sub 2} than UO{sub 2+x}.

Microbes make average 2 nanometer diameter crystalline UO2 particles.

Science.gov (United States)

Suzuki, Y.; Kelly, S. D.; Kemner, K. M.; Banfield, J. F.

2001-12-01

It is well known that phylogenetically diverse groups of microorganisms are capable of catalyzing the reduction of highly soluble U(VI) to highly insoluble U(IV), which rapidly precipitates as uraninite (UO2). Because biological uraninite is highly insoluble, microbial uranyl reduction is being intensively studied as the basis for a cost-effective in-situ bioremediation strategy. Previous studies have described UO2 biomineralization products as amorphous or poorly crystalline. The objective of this study is to characterize the nanocrystalline uraninite in detail in order to determine the particle size, crystallinity, and size-related structural characteristics, and to examine the implications of these for reoxidation and transport. In this study, we obtained U-contaminated sediment and water from an inactive U mine and incubated them anaerobically with nutrients to stimulate reductive precipitation of UO2 by indigenous anaerobic bacteria, mainly Gram-positive spore-forming Desulfosporosinus and Clostridium spp. as revealed by RNA-based phylogenetic analysis. Desulfosporosinus sp. was isolated from the sediment and UO2 was precipitated by this isolate from a simple solution that contains only U and electron donors. We characterized UO2 formed in both of the experiments by high resolution-TEM (HRTEM) and X-ray absorption fine structure analysis (XAFS). The results from HRTEM showed that both the pure and the mixed cultures of microorganisms precipitated around 1.5 - 3 nm crystalline UO2 particles. Some particles as small as around 1 nm could be imaged. Rare particles around 10 nm in diameter were also present. Particles adhere to cells and form colloidal aggregates with low fractal dimension. In some cases, coarsening by oriented attachment on \\{111\\} is evident. Our preliminary results from XAFS for the incubated U-contaminated sample also indicated an average diameter of UO2 of 2 nm. In nanoparticles, the U-U distance obtained by XAFS was 0.373 nm, 0.012 nm

UO2/magnetite concrete interaction and penetration study

International Nuclear Information System (INIS)

Farhadieh, R.; Purviance, R.; Carlson, N.

1983-01-01

The concrete structure represents a line of defense in safety assessment of containment integrity and possible minimization of radiological releases following a reactor accident. The penetration study of hot UO 2 particles into limestone concrete and basalt concrete highlighted some major differences between the two concretes. These included penetration rate, melting and dissolution phenomena, released gases, pressurization of the UO 2 chamber, and characteristics of post-test concrete. The present study focuses on the phenomena associated with core debris interaction with and penetration into magnetite type concrete. The real material experiment was carried out with UO 2 particles and magnetite concrete in a test apparatus similar to the one utilized in the UO 2 /limestone experiment

Heat transfer coefficient between UO2 and Zircaloy-2

International Nuclear Information System (INIS)

Ross, A.M.; Stoute, R.L.

1962-06-01

This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

Oxidative dissolution of ADOPT compared to standard UO2 fuel

International Nuclear Information System (INIS)

Nilsson, Kristina; Roth, Olivia; Jonsson, Mats

2017-01-01

In this work we have studied oxidative dissolution of pure UO 2 and ADOPT (UO 2 doped with Al and Cr) pellets using H 2 O 2 and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO 2 and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO 2 pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO 2. This could be attributed to differences in exposed surface area. However, fission products with low UO 2 solubility display a higher relative release from ADOPT fuel compared to standard UO 2 -fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO 2 which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO 2 fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

Modern x-ray spectral methods in the study of the electronic structure of actinide compounds: Uranium oxide UO2 as an example

Directory of Open Access Journals (Sweden)

Teterin Yury A.

2004-01-01

Full Text Available Fine X-ray photo electron spectral (XPS structure of uranium dioxide UO2 in the binding energy (BE range 0-~č40 eV was associated mostly with the electrons of the outer (OVMO (0-15 eV BE and inner (IVMO (15-40 eV BE valence molecular orbitals formed from the incompletely U5f,6d,7s and O2p and completely filled U6p and O2s shells of neighboring uranium and oxygen ions. It agrees with the relativistic calculation results of the electronic structure for the UO812–(Oh cluster reflecting uranium close environment in UO2, and was confirmed by the X-ray (conversion electron, non-resonance and resonance O4,5(U emission, near O4,5(U edge absorption, resonance photoelectron, Auger spectroscopy data. The fine OVMO and IVMO related XPS structure was established to yield conclusions on the degree of participation of the U6p,5f electrons in the chemical bond, uranium close environment structure and interatomic distances in oxides. Total contribution of the IVMO electrons to the covalent part of the chemical bond can be comparable with that of the OVMO electrons. It has to be noted that the IVMO formation can take place in compounds of any elements from the periodic table. It is a novel scientific fact in solid-state chemistry and physics.

Thermal and Mechanical Properties of UO2 and PuO2

International Nuclear Information System (INIS)

Kato, M.; Matsumoto, T.

2015-01-01

It is important to evaluate basic properties of UO 2 and PuO 2 as fundamental aspects of MA-bearing MOX fuel development. In this work, mechanical properties of UO 2 and PuO 2 were investigated by an ultrasound pulse-echo method. Longitudinal and transversal wave velocities were measured in UO 2 and PuO 2 pellets, and Young's modulus and shear modulus were evaluated, which were 219 MPa and 89 MPa for PuO 2 , and 249 MPa and 95 MPa for UO 2 , respectively. Poisson's ratio was 0.32 in both materials. The relationship between mechanical and thermal properties was described by using thermal expansion data which had been reported previously, and the heat capacity and thermal conductivity were analysed. (authors)

Thermal expansion of UO2 and simulated DUPIC fuel

International Nuclear Information System (INIS)

Ho Kang, Kweon; Jin Ryu, Ho; Chan Song, Kee; Seung Yang, Myung

2002-01-01

The lattice parameters of simulated DUPIC fuel and UO 2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO 2 , and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO 2 . For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO 2 and simulated DUPIC fuel are 10.471x10 -6 and 10.751x10 -6 K -1 , respectively

The bare uranyl(2+) ion, UO22+

International Nuclear Information System (INIS)

Cornehl, H.H.; Heinemann, C.; Marcalo, J.; Pires de Matos, A.; Schwarz, H.

1996-01-01

Ion-molecule reactions between U 2+ and oxygen donors or charge-stripping collisions between singly charged UO 2 2 ions and O 2 collision partners generate uranyl(2+) ions in the gas phase. These do not readily dissociate into singly charged fragments. The standard enthalpy of formation for UO 2 2+ is estimated to be 371±60 kcal mol -1 , in accord with the results of ab initio calculations. (orig.)

The effect of fuel chemistry on UO{sub 2} dissolution

Energy Technology Data Exchange (ETDEWEB)

Casella, Amanda, E-mail: amanda.casella@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-25, Richland, WA 99352 (United States); Hanson, Brady, E-mail: brady.hanson@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, MSIN P7-27, Richland, WA 99352 (United States); Miller, William [University of Missouri Research Reactor, 1513 Research Park Drive, Columbia, MO 65211 (United States)

2016-08-01

The dissolution rate of both unirradiated UO{sub 2} and used nuclear fuel has been studied by numerous countries as part of the performance assessment of proposed geologic repositories. In the scenario of waste package failure and groundwater contact with the fuel, the effects of variables such as temperature, dissolved oxygen, and water and fuel chemistry on the dissolution rates of the fuel are necessary to provide a quantitative estimate of the potential release over geologic time frames. The primary objective of this research was to determine the influence these parameters, with primary focus on the fuel chemistry, have on the dissolution rate of unirradiated UO{sub 2} under oxidizing repository conditions and compare them to the rates predicted by current dissolution models. Both unirradiated UO{sub 2} and UO{sub 2} doped with varying concentrations of Gd{sub 2}O{sub 3}, to simulate used fuel composition after long time periods when radiolysis has minor contributions to dissolution, were examined. In general, a rise in temperature increased the dissolution rate of UO{sub 2} and had a larger effect on pure UO{sub 2} than on those doped with Gd{sub 2}O{sub 3}. Oxygen dependence was observed in the UO{sub 2} samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO{sub 2} matrix resulted in a significant decrease in the dissolution rate. The matrix stabilization effect resulting from the dopant proved even more beneficial in lowering the dissolution rate at higher temperatures and dissolved O{sub 2} concentrations in the leachate where the rates would typically be elevated. - Highlights: • UO{sub 2} dissolution rates were measured for a matrix of repository relevant conditions. • Dopants in the UO{sub 2} matrix lowered the dissolution rate. • Reduction in rates by dopants were increased at elevated temperature and O{sub 2} levels. • UO{sub 2} may be overly

Formation, stability and structural characterization of ternary MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) complexes

Energy Technology Data Exchange (ETDEWEB)

Lee, Jun-Yeop; Yun, Jong-Il [KAIST, Daejeon (Korea, Republic of). Dept. of Nuclear and Quantum Engineering; Vespa, Marika; Gaona, Xavier; Dardenne, Kathy; Rothe, Joerg; Rabung, Thomas; Altmaier, Marcus [Karlsruhe Institute of Technology, Karlsruhe (Germany). Inst. for Nuclear Waste Disposal

2017-06-01

The formation of ternary Mg-UO{sub 2}-CO{sub 3} complexes under weakly alkaline pH conditions was investigated by time-resolved laser fluorescence spectroscopy (TRLFS) and extended X-ray absorption fine structure (EXAFS) and compared to Ca-UO{sub 2}-CO{sub 3} complexes. The presence of two different Mg-UO{sub 2}-C{sub 3} complexes was identified by means of two distinct fluorescence lifetimes of 17±2 ns and 51±2 ns derived from the multi-exponential decay of the fluorescence signal. Slope analysis in terms of fluorescence intensity coupled with fluorescence intensity factor as a function of log [Mg(II)] was conducted for the identification of the Mg-UO{sub 2}-CO{sub 3} complexes forming. For the first time, the formation of both MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) species was confirmed and the corresponding equilibrium constants were determined as log β {sub 113}=25.8±0.3 and β {sub 213}=27.1±0.6, respectively. Complementarily, fundamental structural information for both Ca-UO{sub 2}-CO{sub 3} and Mg-UO{sub 2}-CO{sub 3} complexes was gained by extended EXAFS revealing very similar structures between these two species, except for the clearly shorter U-Mg distance (3.83 Aa) compared with U-Ca distance (4.15 Aa). These results confirmed the inner-sphere character of the Ca/Mg-UO{sub 2}-CO{sub 3} complexes. The formation constants determined for MgUO{sub 2}(CO{sub 3}){sub 3}{sup 2-} and Mg{sub 2}UO{sub 2}(CO{sub 3}){sub 3}(aq) species indicate that ternary Mg-UO{sub 2}-CO{sub 3} complexes contribute to the relevant uranium species in carbonate saturated solutions under neutral to weakly alkaline pH conditions in the presence of Mg(II) ions, which will induce notable influences on the U(VI) chemical species under seawater conditions.

Preparation of UO2 fragments for fuel-debris experiments

International Nuclear Information System (INIS)

Tinkle, M.C.; Kircher, J.A.; Zinn, R.M.; Eash, D.T.

1982-01-01

A unique process was developed for preparing multi-kilogram quantities of > 90% dense fragments of enriched and depleted UO 2 sized 20 mm to 0.038 mm for fuel debris experiments. Precipitates of UO 4 . xH 2 O were treated to obtain UO 2 powders that would yield large cohesive green pieces when isostatically pressed to 206 MPa. The pressed pieces were crushed into fragments that were about 30% oversized, and heated to 1800 0 C for 24 h in H 2 . Oversizing compensates for shrinkage during densification. Effort was dramatically reduced by working on a large scale and by presizing the green UO 2 instead of directly crushing densified pellets

Technological investigation for producing UO2 powder from ADU by using rotary furnace

International Nuclear Information System (INIS)

Pham Duc Thai; Ngo Trong Hiep; Dam Van Tien; Vu Quang Chat; Nguyen Duy Lam; Ngo Xuan Hung; Ngo Quang Hien; Tran Duy Hai; Nguyen Van Sinh

2003-01-01

Uranium dioxide powder UO 2 is main material for producing UO 2 fuel ceramic pellets. The technical characteristics of UO 2 powder directly affect on mechanical and physical characteristics of UO 2 fuel ceramic pellets. Project titled 'Technological investigation for producing UO 2 powder from ADU by using rotary furnace' with the code number BO/01/03-06 for two years 2001 and 2002, on purpose to step by step perfect the technology and equipments for producing UO 2 powder, that is as nuclear fuel. This UO 2 powder may be good material for producing UO 2 fuel ceramic pellets. The results had been achieved as follows: 1. Study on the perfection of the reduction process U 3 O 8 to UO 2 in the gas mixture of 3H 2 + N 2 in inactive condition. 2. Study, design and production of active device system called rotary furnace for manufacturing UO 2 powder from ADU. 3. Study on 4 steps of technology process: drying, calcination, reduction and stabilization of UO 2 powder in the system of rotary furnace from which obtained UO 2 with technical characteristics meeting basic criteria of UO 2 fuel powder. (author)

Vibrational compacting of UO{sub 2} samples in the cladding; Vibraciono kompaktiranje uzoraka UO{sub 2} u zastitnoj kosuljici

Energy Technology Data Exchange (ETDEWEB)

Ristic, M M [Institute of Nuclear Sciences Vinca, Laboratorija za reaktorske materijale, Beograd (Serbia and Montenegro)

1962-12-15

Vibrational compacting was considered as a feasible method for fuel element fabrication. This report describes calibration of the vibrational compacting device. Vibrational compacting of UO{sub 2} was investigated. Obtained densities were not higher than 42% of the theoretical value, i.e. 70% of the possible compacting density. Influence of frequency, acceleration, power and time on the compacted samples was tested. Optimal conditions for UO{sub 2} compacting were as follows: frequency range from 2500 - 4000 Hz; acceleration range from 40 - 100 Hz; maximum power; time of compacting {approx} 5 min. Comparative evaluation of UO{sub 2} and SiO{sub 2} powders was done in order to improve the future development of this method for fabrication of fuel elements.

Studi On Oxidation State Of U In Ba2NdUO6

International Nuclear Information System (INIS)

Firman Windarto, Hendri

1996-01-01

Ba 2 NdUO 6 is not of the important compounds that is formed from a solidification process for high level liquid waste using super high temperature method Ba 2 NdUO 6 has ordered perovskite structure. The objective of this study is to investigate oxidation state of U in Ba 2 NdUO 6 . The properties of Ba 2 NdUO 6 were observed by using Faraday-type torsion magnetometer and X-ray Photoelectron Spectrometer (XPS). The magnetic susceptibility measured in the temperature range of 4K to room temperature showed that the Ba 2 NdUO 6 is paramagnetism that obeys the Curie-Weiss law. The effective moment of Ba 2 NdUO 6 is 3.04 μB. The results of xPs spectrum showed that the peaks of U4f for Ba 2 NdUO 6 appeared exactly between binding energy of UO 2 and UO 3 . It can be concluded that Ba 2 NdUO 6 has binding energy peaks corresponding to pentavalent uranium

Interesting Developments in UO{sub 2} Technology; Progres interessants dans la technologie du bioxyde d'uranium; Interesnye usovershenstvovaniya tekhnologii UO{sub 2}; Recientes progresos en la tecnologia del UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Robertson, J. A.L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

1963-11-15

Now that several UO{sub 2}-fuelled reactors are operating routinely, good irradiation performance of UO{sub 2} is taken for granted. It is therefore stimulating to find that significant developments are still occurring. Most exciting was the recent discovery by Battelle Memorial Institute workers that a particular single crystal of UO{sub 2} had a very high thermal conductivity at elevated temperatures. Following controversy over the matter, an irradiation at Chalk River demonstrated that the large grains formed in operating fuel elements do not necessarily exhibit this enhanced conductivity. Our laboratory experiments have shown that the enhancement is only present in hypostoichiometric compositions and depends little, if any, on the absence of grain boundaries. Indeed, the high conductivity can be obtained in polycrystalline sinters by controlling the stoichiometry. It has long been known that sheath elongation could be reduced by fabricating the UO{sub 2} pellets with depressions in their end faces. Later it was shown that movement of the fuel into a void at the end of the pellet stack was impeded by diametral expansion of the fuel and its mechanical interaction with the sheath. The biggest advance in minimizing sheath distensions has been the realization that longitudinal and diametral expansions are interrelated through the volume expansion of the fuel whose hot core is appreciably plastic. Our empirical knowledge of the factors determining the release of fission-product gases from UO{sub 2} has improved. In particular, increasing the irradiation exposure from 10{sup 15} to 10{sup 18} fissions/cm{sup 3} can reduce the apparent diffusion rates for xenon in UO{sub 2} during subsequent anneals by a factor of 10{sup 3}. The gas is probably immobilized in minute traps, some existing in the original material and some generated by irradiation damage. Detailed analysis indicated slow escape from the traps, presumably from the finite solubility of the xenon in UO{sub 2

Spent fuel UO{sub 2} matrix corrosion behaviour studies through alpha-doped UO{sub 2} pellets leaching

Energy Technology Data Exchange (ETDEWEB)

Muzeau, B.; Jegou, C.; Broudic, V. [CEA-Valrho DEN/DTCD/SECM Laboratoire des Materiaux et Procedes Actifs BP 17171 F-30207 Bagnols-sur-Ceze cedex (France)

2005-07-01

Full text of publication follows: The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behaviour of the UO{sub 2} matrix in aqueous media subjected to {alpha}-{beta}-{gamma} radiations. The {beta}-{gamma} emitters account for the most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persist over much longer time periods and must therefore be taken into account over geological disposal scale. In the present investigation the UO{sub 2} matrix corrosion under alpha radiation is studied as a function of different parameters such as: the alpha activity, the carbonates and hydrogen concentrations,.. In order to study the effect of alpha radiolysis of water on the UO{sub 2} matrix, {sup 238/239}Pu doped UO{sub 2} pellets (0.22 %wt. Pu total) were fabricated with different {sup 238}Pu/{sup 239}Pu ratio to reproduce the alpha activity of a 47 GWd.t{sub HMi}{sup -1} UOX spent fuel at different milestones in time (15, 50, 1500, 10000 and 40000 years). Undoped UO{sub 2} pellets were also available as reference sample. Leaching experiments were conducted in deionized or carbonated water (NaHCO{sub 3} 1 mM), under Argon (O{sub 2} < 0.1 ppm), or Ar/H{sub 2} 30% gas mixture. Previous experiments conducted in deionized water under argon atmosphere, have shown a good correlation between alpha activity and uranium release for the 15-, 1500- and 40000-years alpha doped UO{sub 2} batches. Besides, uranium release in the leachate is controlled either by the kinetics, or by the thermodynamics. Provided the solubility limit of uranium is not achieved, uranium concentration increases and is only limited by the kinetics, unless precipitation occurs and the uranium concentration remains constant over time. These controls are highly dependant on the solution chemistry

The Surface Reactions of Ethanol over UO2(100) Thin Film

KAUST Repository

Senanayake, Sanjaya D.

2015-10-08

The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C1s, O1s and U4f to investigate the bonding mode, surface composition, electronic structure and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion sputtering of this UO2(100) did not result in noticeable reduction of U cations. The ethanol molecule has C-C, C-H, C-O and O-H bonds, and readily donates the hydroxyl H while interacting strongly with the UO2 surfaces. Upon ethanol adsorption (saturation occurred at 0.5 ML), only ethoxy (CH3CH2O-) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO-) on the Ar+-sputtered UO2(100) surface. All ethoxy and acetate species are removed from the surface between 600 and 700 K.

The Surface Reactions of Ethanol over UO2(100) Thin Film

KAUST Repository

Senanayake, Sanjaya D.; Mudiyanselage, Kumudu; Burrell, Anthony K; Sadowski, Jerzy T.; Idriss, Hicham

2015-01-01

The study of the reactions of oxygenates on well-defined oxide surfaces is important for the fundamental understanding of heterogeneous chemical pathways that are influenced by atomic geometry, electronic structure and chemical composition. In this work, an ordered uranium oxide thin film surface terminated in the (100) orientation is prepared on a LaAlO3 substrate and studied for its reactivity with a C-2 oxygenate, ethanol (CH3CH2OH). With the use of synchrotron X-ray photoelectron spectroscopy (XPS), we have probed the adsorption and desorption processes observed in the valence band, C1s, O1s and U4f to investigate the bonding mode, surface composition, electronic structure and probable chemical changes to the stoichiometric-UO2(100) [smooth-UO2(100)] and Ar+-sputtered UO2(100) [rough-UO2(100)] surfaces. Unlike UO2(111) single crystal and UO2 thin film, Ar-ion sputtering of this UO2(100) did not result in noticeable reduction of U cations. The ethanol molecule has C-C, C-H, C-O and O-H bonds, and readily donates the hydroxyl H while interacting strongly with the UO2 surfaces. Upon ethanol adsorption (saturation occurred at 0.5 ML), only ethoxy (CH3CH2O-) species is formed on smooth-UO2(100) whereas initially formed ethoxy species are partially oxidized to surface acetate (CH3COO-) on the Ar+-sputtered UO2(100) surface. All ethoxy and acetate species are removed from the surface between 600 and 700 K.

Oxidative dissolution of ADOPT compared to standard UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

Nilsson, Kristina [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Roth, Olivia [Studsvik Nuclear AB, SE-611 82 Nyköping (Sweden); Jonsson, Mats, E-mail: matsj@kth.se [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2017-05-15

In this work we have studied oxidative dissolution of pure UO{sub 2} and ADOPT (UO{sub 2} doped with Al and Cr) pellets using H{sub 2}O{sub 2} and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO{sub 2} and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO{sub 2} pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO{sub 2.} This could be attributed to differences in exposed surface area. However, fission products with low UO{sub 2} solubility display a higher relative release from ADOPT fuel compared to standard UO{sub 2}-fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO{sub 2} which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO{sub 2} fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

Removal of UO{sup 2+}{sub 2} from aqueous solution using halloysite nanotube-Fe{sub 3}O{sub 4} composite

Energy Technology Data Exchange (ETDEWEB)

He, Wenfang; Chen, Yuantao; Zhang, Wei; Hu, Chunlian; Wang, Jian; Wang, Pingping [Qinghai Normal University, Xining (China)

2016-01-15

Halloysite nanotubes (HNTs) were modified with Fe{sub 3}O{sub 4} to form novel magnetic HNTs-Fe{sub 3}O{sub 4} composites, and the composites were characterized by X-ray diffraction (XRD), transmission electron microscope (TEM), Fourier transform infrared spectroscopy (FT-IR) and vibrating sample magnetometer (VSM). The as-obtained results indicated that Fe{sub 3}O{sub 4} nanoparticles were successfully installed on the surface of HNTs. The adsorption of UO{sup 2+}{sub 2} on HNTs-Fe{sub 3}O{sub 4} was investigated as a function of solid content, contact time, pH, ionic strength and temperature by batch experiments. The consequences revealed that the adsorption of UO{sup 2+}{sub 2} onto HNTs-Fe{sub 3}O{sub 4} was strongly dependent on pH and ionic strength. Equilibrium data fitted well with the Langmuir isotherm. The experimental results demonstrated that the adsorbents with HNTs-Fe{sub 3}O{sub 4} had the largest adsorption capacity of 88.32mg/g for UO{sup 2+}{sub 2}.

Determination of the UO2-ZrO2-BaO equilibrium diagram

International Nuclear Information System (INIS)

Paschoal, J.O.A.; Kleykanp, H.; Thuemmler, F.

1984-01-01

It is determined the equilibrium diagram of UO 2 - ZrO 2 - BaO to interpret and predict changes in the chemical properties of ceramic (oxide) nuclear fuels during irradiation. The isothermal section of the system at 1700 0 C was determined experimentally, utilizing the techniques of ceramography, X-ray diffraction analysis, microprobe analysis and differential thermal analysis. The solid solubility limits at 1700 0 C between UO 2 and ZrO 2 , UO 2 and BaO, ZrO 2 and BaO, ZrO 2 and BaO and BaUO 3 and BaZrO 3 is presented. The influence of oxygen potential in relation to the different phases is discussed and the phase diagram of the system presented. (M.C.K.) [pt

Adsorptive features of poli(acrylic acid-co-hydroxyapatite) composite for UO22+

International Nuclear Information System (INIS)

Liu Tonghuan; Xu Zhen; Tan Yinping; Zhong Qiangqiang; Wu Wangsuo

2016-01-01

The copolymer of poli(acrylic acid-co-hydroxyapatite) (PAA-HAP) was prepared and characterized by means of FT-IR and SEM analysis. The adsorptive features of PAA-HAP for UO 2 2+ was studied as a function of pH, adsorbent dosage, initial metal ion concentration and temperature. The adsorption isotherm data fitted well to the Langmuir isotherm model. The adsorbed UO 2 2+ can be desorbed effectively by 0.1 M HNO 3 . The maximum adsorption capacities for UO 2 2+ of the dry PAA-HAP was 1.86 x 10 -4 mol/g. The high adsorption capacity and kinetics results indicate that PAA-HAP can be used as an alternative adsorbent to remove UO 2 2+ from aqueous solution. (author)

Ceramic UO2 powder production at Cameco Corporation

International Nuclear Information System (INIS)

Mulligan, J.J.

2005-01-01

This paper describes the various aspects of ceramic grade UO 2 powder production at Cameco Corporation's Port Hope conversion facility. It discusses the significant safety systems, production processes and plant monitoring and control systems. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development contribute to the consistent production of high quality UO 2 powder. The ability of Cameco to identify, measure and control the physical and chemical properties of ceramic grade UO 2 has resulted in the production of uniform quality powder that has consistently met customer requirements. (author)

Potentiometric and spectrophotometric characterization of the UO{sub 2}{sup 2+}-citrate complexes in aqueous solution, at different concentrations, ionic strengths and supporting electrolytes

Energy Technology Data Exchange (ETDEWEB)

Berto, S.; Daniele, P.G.; Prenesti, E. [Torino Univ. (Italy). Dipt. di Chimica Analitica; Crea, F.; De Stefano, C.; Sammartano, S. [Messina Univ. (Italy). Dipt. di Chimica Inorganica, Chimica Analitica e Chimica Fisica

2012-07-01

In this paper we report an investigation on the interactions between dioxouranium(VI) and citrate using potentiometry (H{sup +}-glass electrode) and UV-spectrophotometry. Potentiometric measurements were carried out in NaCl and KNO{sub 3} aqueous solutions at t = 25 C in a wide range of experimental conditions (concentrations, ligand/metal molar ratio, pH, titrants). Measurements in NaCl were carried out at different ionic strength values (0.1 {<=} I/mol L{sup -1} {<=} 1.0); different procedures were employed for the acquisition of experimental data and careful analysis of these data performed. In all cases the speciation model that best fits experimental data takes into account the formation of the following species: UO{sub 2}(Cit){sup -}, (UO{sub 2}){sub 2}(Cit){sub 2}{sup 2-}, (UO{sub 2}){sub 2}(Cit){sub 2}(OH){sub 2}{sup 4-}, (UO{sub 2}){sub 2}(Cit){sub 2}(OH){sup 3-}, (UO{sub 2}){sub 2}(Cit)(OH){sub 2}{sup -}, (UO{sub 2}){sub 2}(Cit)(OH){sup 0}, (UO{sub 2}){sub 3}(Cit){sub 2}(OH){sub 5}{sup 5-}. The dependence on ionic strength of formation constants was taken into account by using both a simple Debye-Hueckel type equation and the SIT (specific ion interaction theory) approach. Moreover, a visible absorption spectrum for each complex reaching a significant percentage of formation in solution (KNO{sub 3} medium) has been calculated to characterise the compounds found by pH-metric refinement. Recommended values for the uranyl-citrate species were proposed for each ionic strength values in NaCl aqueous solution. Comparison with literature stability constants is reported too. (orig.)

Overview of interatomic potentials

International Nuclear Information System (INIS)

Bonny, G.; Malerba, L.

2005-12-01

In this report an overview on interatomic potentials is given. This overview is by no means complete and it has merely the intention to give the reader an idea of where interatomic potentials come from, as well as to provide the basic ideas behind some commonly used methods for deriving interatomic potentials for molecular dynamics applications. We start by giving a short introduction about the concept of interatomic potential in the framework of quantum mechanics, followed by a short description of commonly used methods for deriving semi-empirical interatomic potentials. After some short theoretical notions on each method, some practical parameterizations of commonly used potentials are given, including very recent ones. An effort has been made to classify existing approaches within a rational and consequent scheme, which is believed to be of use for a thorough comprehension of the topic. Although these approaches can be used in a variety of different materials, we will only discuss the practical cases of metals. Following this, some widespread ad hoc modification of the general methods are discussed. The report is concluded by a generalization of the methods to multi-component materials, in particular metallic alloys. (author)

Fabrication of ThO2 and ThO2-UO2 pellets for proliferation resistant fuels

International Nuclear Information System (INIS)

Matthews, R.B.; Davis, N.C.

1979-10-01

To meet this objective, batches of ThO 2 powders were compared and milling parameters, pressing and sintering conditions were established. A method for blending ThO 2 and UO 2 into homogeneous powders that press and sinter into 95% TD pellets was determined. The effect of UO 2 additions on ThO 2 -UO 2 pellet properties was determined and a process for fabricating irradiation test quality ThO 2 -20 wt% UO 2 pellets containing CaO as a dissolution aid was established

Measurement of the friction coefficient between UO2 and cladding tube

International Nuclear Information System (INIS)

Tachibana, Toshimichi; Narita, Daisuke; Kaneko, Hiromitsu; Honda, Yutaka

1978-01-01

Most of fuel rods used for light water reactors or fast reactors consist of the cladding tubes filled with UO 2 -PuO 2 pellets. The measurement was made on the coefficient of static friction and the coefficient of dynamic friction in helium under high contact load on UO 2 /Zry-2 and UO 2 /SUS 316 combined samples at the temperature ranging from room temperature to 400 deg. C and from room temperature to 600 deg. C, respectively. The coefficient of static friction for Zry-2 tube and UO 2 pellets was 0.32 +- 0.08 at room temperature and 0.47 +- 0.07 at 400 deg. C, and increased with temperature rise in this temperature range. The coefficient of static friction between 316 stainless steel tube and UO 2 pellets was 0.29 +- 0.04 at room temperature and 1.2 +- 0.2 at 600 deg. C, and increased with temperature rise in this temperature range. The coefficient of dynamic friction for both UO 2 /Zry-2 and UO 2 /SUS 316 combinations seems to be equal to or about 10% excess of the coefficient of static friction. The coefficient of static friction for UO 2 /SUS 316 combination decreased with the increasing number of repetition, when repeating slip several times on the same contact surfaces. (Kobatake, H.)

Some aspects of UO{sub 2} powder production

Energy Technology Data Exchange (ETDEWEB)

Balakrishna, P; Asnani, C K; Prabhakar Rao, L; Kartha, R M; Pillai, P K.M. [Nuclear Fuel Complex, Hyderabad (India)

1994-06-01

UO{sub 2} powder is being produced in a chemical plant from enriched UF{sub 6} and supplied to the pelletizing plant. Small quantities of scrap UO{sub 2} received back from the pelletizing plant are also recycled in the chemical plant to produce UO{sub 2} powder. The powder should be of a consistently high quality so as to finally yield high density sintered pellets with minimum rejection. The final yield of acceptable finished pellets depends on the quality of the powder in the chemical plant as well as the quality of pressing in the pelletizing plant. In this paper, some examples of measures adopted for achieving good quality powder production are presented. (author). 9 refs., 2 figs.

Photochemical assessment of UO2+2 complexation in Triton X-100 micellar system

International Nuclear Information System (INIS)

Das, S.K.; Ganguly, B.N.

1994-01-01

This is a report on the spectral characteristics of UO 2 +2 in the excited state in the Triton X-100 micellar medium. The downward curving of the Stern-Volmer plot explains the two kinds of populations of UO 2 +2 upon micellization. A blue shift of the quenched emission is ascribed due to the collisional encounter of UO 2 +2 with the head groups of Triton X-100. (author). 5 refs., 2 figs

Thermal expansion of ThO2-2 wt% UO2 by HT-XRD

International Nuclear Information System (INIS)

Tyagi, A.K.; Mathews, M.D.

2000-01-01

The linear thermal expansion of polycrystalline ThO 2 -2 wt% UO 2 has been investigated from room temperature to 1473 K in flowing helium atmosphere using high temperature X-ray diffractometry. ThO 2 -2 wt% UO 2 shows a marginally higher linear thermal expansion as compared to pure ThO 2 . The average linear and volume thermal expansion coefficients of ThO 2 -2 wt% UO 2 are found to be α-bar a =9.74x10 -6 K -1 and α-bar v =29.52x10 -6 K -1 (298-1473 K). This study will be useful in designing the nuclear reactor fuel assembly based on ThO 2

Antiferromagnetic-ferromagnetic crossover in UO2-TiOx multi-phase systems

International Nuclear Information System (INIS)

Nakamura, Akio; Tsutsui, Satoshi; Yoshii, Kenji

2001-01-01

An antiferromagnetic (AF)-weakly ferromagnetic (WF) crossover has been found for UO 2 -TiO x multi-phase systems, (1-y)UO 2 +yTiO x (y=0.05-0.72, x=0, 1.0, 1.5 and 2.0), when these mixtures are heat treated at high temperature in vacuum. From the powder X-ray diffraction and electron-microprobe analyses, their phase assemblies were as follows: for x=0, 1.0 and 1.5, a heterogeneous two-phase mixture of UO 2 +TiO x ; for x=2.0, that of UO 2 +UTi 2 O 6 for y 0.67 that of UTi 2 O 6 +TiO 2 (plus residual minor UO 2 ). Magnetic susceptibility (χ) of the present UO 2 powder was confirmed to exhibit an antiferromagnetic sharp drop at T N (=30.5 K). In contrast, χ of these multi-phase systems was found to exhibit a sharp upturn at the respective T N , while their T N values remained almost constant with varying y. This χ upturn at T N is most pronounced for UO 2 +Ti-oxide (titania) systems (x=1.0, 1.5 and 2.0) over the wide mixture ratio above y∼0.10. These observations indicate that an AF-WF crossover is induced for these multi-phase systems, plausibly due to the interfacial magnetic modification of UO 2 in contact with the oxide partners

Development on UO3-K2O and MoO3-K2O binary systems and study of UO2MoO4-MoO3 domain within UO3-MoO3-K2O ternary system

International Nuclear Information System (INIS)

Dion, C.; Noel, A.

1983-01-01

This paper confirms the previous study on the MoO 3 -K 2 O system, and constitutes a clarity of the UO 3 -K 2 O system. Four distinct uranates VI with alkaline metal/uranium ratio's 2, 1, 0,5 and 0,285 exist. Preparation conditions and powder diffraction spectra of these compounds are given. Additional informations relative to K 2 MoO 4 allotropic transformations are provided. Study of UO 2 MoO 4 -K 2 MoO 4 diagram has brought three new phases into prominence: (B) K 6 UMo 4 O 18 incongruently melting point, (E) K 2 UMo 2 O 10 congruently melting and (F) K 2 U 3 Mo 4 O 22 incongruently melting point. Within MoO 3 -K 2 MoO 4 -UO 2 MoO 4 ternary system, no new phase is found. The general appearance of ternary liquidus and crystallization fields of several compounds are given. These three new compounds become identified with these of UO 2 MoO 4 -Na 2 MoO 4 binary system [fr

Fabrication and testing of ceramic UO{sub 2} fuel - I-III. Part II, Fabrication of sintered pressed samples UO{sub 2} (Final report); Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, II Deo - Dobijanje sinterovanih ispresaka UO{sub 2} (zavrsni izvestaj)

Energy Technology Data Exchange (ETDEWEB)

Novakovic, M; Ristic, M M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Procedure for fabrication of sintered ceramic UO{sub 2} pellets was developed in the Department of reactor materials. The tasks described in this report deal with design and construction of laboratory equipment for treatment of ceramic materials, and fabrication of UO{sub 2} pellets. The procedure was based on cold pressing of appropriately prepared powder and sintering of the of thus obtained pressed samples.

Particle size distribution of UO sub 2 aerosols

Energy Technology Data Exchange (ETDEWEB)

Raghunath, B. (Radiation Safety Systems Div., BARC, Bombay (India)); Ramachandran, R.; Majumdar, S. (Radiometallurgy Div., BARC, Bombay (India))

1991-12-01

The Anderson cascade impactor has been used to determine the activity mean aerodynamic diameter and the particle size distribution of UO{sub 2} powders dispersed in the form of stable aerosols in an air medium. The UO{sub 2} powders obtained by the calcination of ammonium uranyl carbonate (AUC) and ammonium diuranate (ADU) precipitates have been used. (orig./MM).

Porosity influence on UO2 pellet fracture

International Nuclear Information System (INIS)

Quadros, N.F. de; Abreu Aires, M. de; Gentile, E.F.

1976-01-01

Compression tests were made with UO 2 pellets with grain size of 0,01 mm, approximately the same for all pellets, and with different porosities. The strain rate was 5,5 X 10 -5 sec -1 at room temperature. From fractographic studies and observations made during the compression tests, it was suggested that the pores and flaws resulting from sintering at 1650 0 C, play a fundamental role on the fracture mechanism of the UO 2 pellets [pt

Recycling process of Mn-Al doped large grain UO2 pellets

International Nuclear Information System (INIS)

Nam, Ik Hui; Yang, Jae Ho; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Hun; Kim, Keon Sik; Song, Kun Woo

2010-01-01

To reduce the fuel cycle costs and the total mass of spent light water reactor (LWR) fuels, it is necessary to extend the fuel discharged burn-up. Research on fuel pellets focuses on increasing the pellet density and grain size to increase the uranium contents and the high burnup safety margins for LWRs. KAERI are developing the large grain UO 2 pellet for the same purpose. Small amount of additives doping technology are used to increase the grain size and the high temperature deformation of UO 2 pellets. Various promising additive candidates had been developed during the last 3 years and the MnO-Al 2 O 3 doped UO 2 fuel pellet is one of the most promising candidates. In a commercial UO 2 fuel pellet manufacturing process, defective UO 2 pellets or scraps are produced and those should be reused. A common recycling method for defective UO 2 pellets or scraps is that they are oxidized in air at about 450 .deg. C to make U 3 O 8 powder and then added to UO 2 powder. In the oxidation of a UO 2 pellet, the oxygen propagates along the grain boundary. The U 3 O 8 formation on the grain boundary causes a spallation of the grains. So, size and shape of U 3 O 8 powder deeply depend on the initial grain size of UO 2 pellets. In the case of Mn-Al doped large grain pellets, the average grain size is about 45μm and about 5 times larger than a typical un-doped UO 2 pellet which has grain size of about 8∼10μm. That big difference in grain size is expected to cause a big difference in recycled U 3 O 8 powder morphology. Addition of U 3 O 8 to UO 2 leads to a drop in the pellet density, impeding a grain growth and the formation of graph- like pore segregates. Such degradation of the UO 2 pellet properties by adding the recycled U 3 O 8 powder depend on the U 3 O 8 powder properties. So, it is necessary to understand the property and its effect on the pellet of the recycled U 3 O 8 . This paper shows a preliminary result about the recycled U 3 O 8 powder which was obtained by

Thermal reactions of uranium metal, UO 2, U 3O 8, UF 4, and UO 2F 2 with NF 3 to produce UF 6

Science.gov (United States)

McNamara, Bruce; Scheele, Randall; Kozelisky, Anne; Edwards, Matthew

2009-11-01

This paper demonstrates that NF 3 fluorinates uranium metal, UO 2, UF 4, UO 3, U 3O 8, and UO 2F 2·2H 2O to produce the volatile UF 6 at temperatures between 100 and 550 °C. Thermogravimetric and differential thermal analysis reaction profiles are described that reflect changes in the uranium fluorination/oxidation state, physiochemical effects, and instances of discrete chemical speciation. Large differences in the onset temperatures for each system investigated implicate changes in mode of the NF 3 gas-solid surface interaction. These studies also demonstrate that NF 3 is a potential replacement fluorinating agent in the existing nuclear fuel cycle and in actinide volatility reprocessing.

The production of sinterable UO2 from AUC

International Nuclear Information System (INIS)

Chang, I.S.; Do, J.B.; Choi, Y.D.; Park, M.H.; Yun, H.H.; Kim, E.H.; Kim, Y.W.

1982-01-01

Fluidization, feeding and discharging, and mixing of fine particles (-up to 40μ in diameter) in fluidized bed reactor has been examined. The degree of conversion has been estimated using the kinetic data differential scanning colorimetry(DSC) and thermogravimetic analysis (TGA) of ammonium uranyl carbonate (AUC) and residence time distribution data. Satisfactory operation is obtained with a sintered ceramic distributor and filters. The reactor equilvalent to approximately 1.1-1.3 stages. Thermal analysis of AUC in hydrogen atmosphere shows that the decomposition of AUC to UO 3 at 200degC is followed by reduction of UO 3 to UO 2 in two steps in the range between 400degC and 500degC and the complete conversion to UO 2 takes two minutes at 550degC. The overall conversion of above 99.5% in the fluidized bed reactor is estimated with 40 minutes of a mean particle residence time at 600degC. (Author)

Creep of UO2 at 25000C

International Nuclear Information System (INIS)

Slagle, O.D.

1977-01-01

Until an improved high temperature relationship is available, the previously derived low temperature relationship is a reasonable means for predicting the creep rates of UO 2 at 2500 0 C. The activation energy determined for creep at 2500 0 C is at least two times larger than that measured previously at the lower temperature. Stress induced grain growth under uniaxial compression at high temperatures in UO 2 results in preferential growth of grains having a cube axis closely aligned with the stress axis

Geometrical dimensioning of PWR UO2 pellets

International Nuclear Information System (INIS)

Silva, A.T.

1988-08-01

The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pelet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experiemtnal data. (author) [pt

TCA UO2/MOX core analyses

International Nuclear Information System (INIS)

Tahara, Yoshihisa; Noda, Hideyuki

2000-01-01

In order to examine the adequacy of nuclear data, the TCA UO 2 and MOX core experiments were analyzed with MVP using the libraries based on ENDF/B-VI Mod.3 and JENDL-3.2. The ENDF/B-VI data underpredict k eff values. The replacement of 238 U data with the JENDL-3.2 data and the adjustment of 235 ν-value raise the k eff values by 0.3% for UO 2 cores, but still underpredict k eff values. On the other hand, the nuclear data of JENDL-3.2 for H, O, Al, 238 U and 235 U of ENDF/B-VI whose 235 ν-value in thermal energy region is adjusted to the average value of JENDL-3.2 give a good prediction of k eff . (author)

A Feasibility Study on UO2/ZrO2 Mixture Melting using Induction Skull Melting Method

International Nuclear Information System (INIS)

Hong, S. W.; Kim, J. H.; Kim, H. D.

1998-01-01

Using ISM(Induction Skull Melting) method, which is usually used for the crystallization of refractory materials, a feasibility study on melting of the UO 2 /ZrO 2 mixture(w/o 8:2) is carried out. Frequency, one of main design parameters for ISM, is determined from electrical resistance of UO 2 /ZrO 2 mixture. Heat loss from the crucible for UO 2 /ZrO 2 20kg melting is predicted by comparison with the existing experimental data for UO , ZrO 2 , and ThO 2 . The analysis shows that melting and superheating of the UO 2 /ZrO 2 mixture using induction skull melting method is possible. To attain the superheat of 300K for 20 kg of UO 2 /ZrO 2 , 100kHz, 100 kW power input for induction coil, and 570L/min coolant flow rate are found to be required. The results of this feasibility study will be adopted for designing UO 2 /ZrO 2 furnace using actual corium material at KAERI

Irradiation of UO2+x fuels in the TANOX device

International Nuclear Information System (INIS)

Dehaudt, P.; Caillot, L.; Delette, G.; Eminet, G.; Mocellin, A.

1998-01-01

The TANOX analytical irradiation device is presented and the first results concerning stoichiometric and hyper stoichiometric uranium dioxide fuels with two different grain sizes are given. The TANOX device is designed to obtain rapidly significant burnups in fuels at relatively low temperatures. It is placed at the periphery of the SILOE reactor and translated to adjust the irradiation power. The continuous measure of the centre-line temperature allows to control the experiment and to evaluate the thermal behaviour of the rods. A TANOX fuel rod has a length of 100 mm with 20 fuel pellets in a stainless steel cladding and is inserted in a thick aluminium alloy overcladding which is cooled by the primary water circuit reactor. These conditions of small size pellets and improved thermal exchanges have been designed to dissipate the heat power due to fission densities three to five times higher than in a PWR. The first analytical irradiation was devoted to the study of UO 2.00 , UO 2.01 and UO 2.02 fuels with standard and large grain sizes obtained by annealing. A burnup of about 9000 MWd.t -1 U was reached in these fuels. The thermal analysis shows a degraded conductivity for the UO 2.02 fuel rod due to the hyper stoichiometry. The released fractions of 85 Kr during irradiation are negligible as expected (lower than 0,1%). Some of the pellets were heat treated at 1700 deg. C for 5 hours. The gas release was analysed after 30 minutes and at the end of the treatment. The main results are as follows: the fission gas release (FGR) of the standard UO 2 varies from one sample to another; the FGR of the hyper stoichiometric fuels is of the same order of magnitude than that of the stoichiometric UO 2 fuel of normal grain sizes; the grain size increase has no effect on FGR for UO 2.00 but considerably decreases the FGR for UO 2.01 and UO 2.02 fuels. These heat treated samples are also observed to characterize the inter- and intragranular fission gas bubbles. (author)

Defect trap model of gas behaviour in UO2 fuel during irradiation

International Nuclear Information System (INIS)

Szuta, A.

2003-01-01

Fission gas behaviour is one of the central concern in the fuel design, performance and hypothetical accident analysis. The report 'Defect trap model of gas behaviour in UO 2 fuel during irradiation' is the worldwide literature review of problems studied, experimental results and solutions proposed in related topics. Some of them were described in details in the report chapters. They are: anomalies in the experimental results; fission gas retention in the UO 2 fuel; microstructure of the UO 2 fuel after irradiation; fission gas release models; defect trap model of fission gas behaviour; fission gas release from UO 2 single crystal during low temperature irradiation in terms of a defect trap model; analysis of dynamic release of fission gases from single crystal UO 2 during low temperature irradiation in terms of defect trap model; behaviour of fission gas products in single crystal UO 2 during intermediate temperature irradiation in terms of a defect trap model; modification of re-crystallization temperature of UO 2 in function of burnup and its impact on fission gas release; apparent diffusion coefficient; formation of nanostructures in UO 2 fuel at high burnup; applications of the defect trap model to the gas leaking fuel elements number assessment in the nuclear power station (VVER-PWR)

Characterization of Compaction Process on UO2 Powder Pelletisation

International Nuclear Information System (INIS)

Rachmawati, M; Langenati, R; Saputra, T.T; Mahpudin, A; Histori; Sutarya, D; Zahedi

1998-01-01

Determination of compaction pressure of pelletization which is based on density characterization in conjunction with satisfactory green strength of the UO 2 pellet, is carried out in this experiment. Cameco UO 2 powder has been mixed up with Zn-stearate lubricant prior to compaction process. The compaction pressure is varied from the range of 2 Mp up to 6 Mp. The mechanical strength is determined using diametral compression strength with the speed of loading of 0.1 mm.min 1 . The density measurement and compression strength test are performed on each of the applied pressure. The result shows that compaction at 5 Mp gives the maximum green strength of UO 2 pellet, while the maximum density is achieved at 5.7 Mp. The maximum green strength and green density of UO 2 (+ TiO 2 ) pellets is achieved at the addition of 0.25% and 0.125% TiO 2 respectively. The compaction pressure which is showing the maximum pellet green strength but still having the required density, is chosen to be the determinant compaction pressure in condition of pelletization

A Knowledge- Based Computer System for UO2 Characterization According to ASTM Requirements

International Nuclear Information System (INIS)

Afifi, Y.K.; El-Hakim, E.

2000-01-01

The uranium dioxde (UO 2 ) powder properties and the pellets fabrication processes determine the characteristics of the sintered UO 2 pellets. The powder properties include chemical and physical characteristics. The physical and chemical properties of UO 2 powder are normally checked to ensure consistency and reproducibility of the sintered UO 2 pellets. Powder characteristics are known to influence the subsequent manufacturing performance or the fuel properties. The aim of this paper is to provide the nuclear industry with a program dealing with the processes and the related requirements to determine the specifications of UO 2 powder according to the American Standards for Testing and Materials (ASTM). This program covers the physical and chemical characteristics of UO 2 powder. A group of logic flow charts dealing with the data and information available in the ASTM for each step in the characterization of UO 2 powder process and the technical assistance are constructed. These logic flow charts are collected to form a module of the software to qualify the UO 2 powder. The program contains 8 modules, each one deals with one object. This program saves time, is also considered as a collective schema for all the required UO 2 powder characterization and the related processes, and could be used as a training tool for less skilled personnel involved in UO 2 powder characterization laboratories

PROCESS FOR THE PRODUCTION OF AN ACTIVATED FORM OF UO$sub 2$

Science.gov (United States)

Polissar, M.J.

1957-09-24

A process for producing a highly active form of UO/sub 2/ characterized both by rapid oxidation in air and by rapid chlorination with CCl/sub 4/ vapor at an elevated temperature is reported. In accordance with the process, commercial UO/sub 2/, is subjected to a series of oxidation-reduction operations to produce a form of UC/sub 2/ of enhanced reactivity. By treatimg commercial UO/sub 2/ at a temperature between 335 and 485 deg C with methane, then briefly with an oxygen containing gas and followimg this by a second treatment with a methane containing gas, the original relatively stable charge of UO/sub 2/ will be transformed into an active form of UO/sub 2/.

Study on the retention of enriched UO2F2 in the mouse and its radiogenotoxicological effects

International Nuclear Information System (INIS)

Hu Qiyue; Zhu Shoupeng

1991-06-01

The study on toxicological effects of enriched UO 2 F 2 was undertaken in purebred BALB/c male mice to examine: (a) the retention in body; (b) the testicular clearance; (c) the effect of sperm abnormality; (d) the effect of chromosomal aberrations in spermatogonia and primary spermatocytes; and (e) the effect of DNA damage in germ cells in various spermiogenic stages. Results show that enriched UO 2 F 2 mainly deposited in the kidneys, then the skeleton and liver. The amount of enriched UO 2 F 2 depositing in other tissues was small. Enriched UO 2 F 2 was similar to the natural uranium in transference and retention in the body. The testis had efficient clearance of enriched UO 2 F 2 . Enriched UO 2 F 2 could result in sperm abnormality. Even with the same treating does but at different treating time the rates of sperm abnormality were different. Enriched UO 2 F 2 could result in chromosomal aberrations in spermatogonia and primary spermatocytes. The important type of aberrations in spermatogonia was break. For primary spermatocytes the most significant aberration was multivalents. Enriched UO 2 F 2 could also result in DNA breakage in germ cells. The sensitivity of mouse germ cells at various stages to enriched UO 2 F 2 was different. There was a linear relationship between the amount of sperm DNA eluted and enriched UO 2 F 2 dose

A calorimetric and thermodynamic investigation of A2[(UO2)2(MoO4)O2] compounds with A = K and Rb and calculated phase relations in the system (K2MoO4 + UO3 + H2O)

International Nuclear Information System (INIS)

Lelet, Maxim I.; Suleimanov, Evgeny V.; Golubev, Aleksey V.; Geiger, Charles A.; Bosbach, Dirk; Alekseev, Evgeny V.

2015-01-01

Highlights: • We determined the low temperature heat capacity of A 2 [(UO 2 ) 2 (MoO 4 )O 2 ] compounds with A = K and Rb. • We determined enthalpy of formation of K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] by HF solution calorimetry. • We calculated Δ f G° (T = 298 K) of all phases from studied series. • Using obtained data we performed a thermodynamic modelling in the system (K 2 MoO 4 + UO 3 + H 2 O). - Abstract: A calorimetric and thermodynamic investigation of two alkali-metal uranyl molybdates with general composition A 2 [(UO 2 ) 2 (MoO 4 )O 2 ], where A = K and Rb, was performed. Both phases were synthesized by solid-state sintering of a mixture of potassium or rubidium nitrate, molybdenum (VI) oxide and gamma-uranium (VI) oxide at high temperatures. The synthetic products were characterised by X-ray powder diffraction and X-ray fluorescence methods. The enthalpy of formation of K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] was determined using HF-solution calorimetry giving Δ f H° (T = 298 K, K 2 [(UO 2 ) 2 (MoO 4 )O 2 ], cr) = −(4018 ± 8) kJ · mol −1 . The low-temperature heat capacity, C p °, was measured using adiabatic calorimetry from T = (7 to 335) K for K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] and from T = (7 to 326) K for Rb 2 [(UO 2 ) 2 (MoO 4 )O 2 ]. Using these C p ° values, the third law entropy at T = 298.15 K, S°, is calculated as (374 ± 1) J · K −1 · mol −1 for K 2 [(UO 2 ) 2 (MoO 4 )O 2 ] and (390 ± 1) J · K −1 · mol −1 for Rb 2 [(UO 2 ) 2 (MoO 4 )O 2 ]. These new experimental results, together with literature data, are used to calculate the Gibbs energy of formation, Δ f G°, for both phases giving: Δ f G° (T = 298 K, K 2 [(UO 2 ) 2 (MoO 4 )O 2 ], cr) = (−3747 ± 8) kJ · mol −1 and Δ f G° (T = 298 K, Rb 2 [(UO 2 ) 2 (MoO 4 )], cr) = −3736 ± 5 kJ · mol −1 . Smoothed C p °(T) values between 0 K and 320 K are presented, along with values for S° and the functions [H°(T) − H°(0)] and [G°(T) − H°(0)], for both phases. The

Experimental investigations into the spectral reflectivity and emissivity of liquid UO2, UC, ThO2, and Nd2O3

International Nuclear Information System (INIS)

Karow, H.U.; Bober, M.

1979-01-01

Fast reactor safety research requires knowledge of emissivity data of nuclear fuel materials up to temperatures of the liquid state. A special integrating sphere laser reflectometer has been used to measure the normal reflectivity and emissivity of UO 2 , UC, ThO 2 , and in addition of Nd 2 O 3 in the solid state (premolten, refrozen material) and in the liquid state up to temperatures of 4000 to 4800 K. The measuring wavelengths have been 0.63 μm and 10.6 μm. The emissivity curves of the oxidic specimens measured at 0.63 μm show the same characteristic course: little temperature dependence below the melting point, distinct increase in the liquid state. In the case of UO 2 the emissivity at the melting point (3120 K) is 0.84, at 4100 K it is 0.92. At 10.6 μm, a decrease has been measured for the liquid state of UO 2 and ThO 2 . UC shows in the solid and in the liquid state only a small temperature dependence with a marked drop, however, at the melting point (2780 K) from 0.54 to 0.45. The measuring results are presented by diagrams and by fit equations related to the true and the black temperature, respectively. (orig./HP) [de

Vapor deposition of large area NpO2 and UO2 deposits

International Nuclear Information System (INIS)

Adair, H.L.; Gibson, J.R.; Kobisk, E.H.; Dailey, J.M.

1976-01-01

Deposition of NpO 2 and UO 2 thin films over an area of 7.5 to 10 cm diam has become a routine operation in preparation of fission chamber plates. Vacuum evaporation or electroplating has been used for this purpose. The ''paint brush'' technique has been used as well; however, uniformity requirements normally eliminate this procedure. Vapor deposition in vacuum appears to be the most suitable technique for preparing NpO 2 and UO 2 deposits of >200 cm 2 . This paper describes the procedures used in preparing uniform large area deposits of NpO 2 (approximately 300 cm 2 ) and UO 2 (approximately 2000 cm 2 ) by vacuum evaporation using electron bombardment heating and several substrate motion and heating methods to achieve uniformity and adhesion

Sinterability of mixtures of UO2 of different morphological features

International Nuclear Information System (INIS)

Villegas de Maroto, Marina; Celora de Lavagnino, Julia; Marajofsky, Adolfo; Leyva, A.G.

1981-01-01

The reprocessing of scrap in the production of UO2 pellets, is important from an economical view-point of the fuel cycle. The recovery method by means of a humid process, tested for UO2 scrap, includes the dissolution of the pellets in a nitric media at boiling point, the precipitation of ammonium diuranates (ADU) and its conversion into UO2 at 600 deg C. The microestructural results and the sintering density of the pellets produced in these tests are compared. It is shown that, although the addition of said UO2 powders impaires the performance of the original mixture produced by the factory, the results thus obtained are, nevertheless, within specifications. This facts show that the mixture would then be able for production. (M.E.L.) [es

A new UO2 sintering technology for the recycling of defective fuel pellets

International Nuclear Information System (INIS)

Song, K. W.; Kim, K. S.; Jeong, Y. H.

1998-01-01

A new UO 2 sintering technology to recycle defective UO 2 pellets has been developed. The defective UO 2 pellets were oxidized in an air to produce U 3 O 8 powder, and the U 3 O 8 powder was mixed with fresh AUC-UO 2 powder in the range of 10 to 100 wt%. Nb 2 O 5 and TiO 2 are added to the mixed powder. The mixed powder was pressed and sintered at 1680 deg C for 4 hours in hydrogen. The density of UO 2 pellets without sintering agents decreased linearly with the U 3 O 8 content at the rate of 0.2 %TD per 1 wt% U 3 O 8 , and the density was below 93.5 %TD at the U 3 O 8 contents above 10 wt%. However, the mixed UO 2 and U 3 O 8 powder containing Nb 2 O 5 (≥0.3 wt%) and TiO 2 (≥0.1 wt%) yielded a sintered density above 94 %TD in all ranges of U 3 O 8 contents. It was found that higher mixing ratios of U 3 O 8 to UO 2 powder did not affect the grain size of UO 2 pellets under the addition of Nb 2 O 5 , but decreased the grain size of UO 2 pellets under the addition of TiO 2 . The doped UO 2 pellets have grain sizes larger than 20 μm, and have small density gain after re-sintering test, owing to large pores. Therefore, the sintering agents such as Nb 2 O 5 and TiO 2 can make highly densified UO 2 pellets from the powder comprising a large amount of U 3 O 8 powder

Boiling point measurements on liquid UO2

International Nuclear Information System (INIS)

Bober, M.; Singer, J.; Trapp, M.

1986-01-01

In analogy to the classic boiling point method, a quasi-stationary millisecond laser-heating technique was applied to measure the saturated-vapour pressure curve of liquid UO 2 in the temperature range of 3500 to 4500 K. The result is represented by log p(MPa) 5.049 -23042/T(K) according to an average heat of vaporization of 441 kJ/mol and a normal boiling point of 3808 K. Besides, spectral emissivities of liquid UO 2 were measured at the pyrometer wavelengths of 752 and 1064 nm. (author)

UO2: production based on two alternative lines

International Nuclear Information System (INIS)

Coppa, R.C.; Martin, H.R.

1987-01-01

The production of the uranium dioxide (UO 2 ) is carried out at the Cordoba factory, of the Argentine National Atomic Energy Commission, by the uranil carbonate method (AUC). The commercial uranium concentrates (yellow cake) is dissolved with HNO 3 and purificated with tributil phosphate (TBP). The pure uranium compound coming from the reextraction, is concentrated to 0.4 Kg U/l, then the precipitation with CO 2 and NH 3 gives the AUC crystalls. After conversion of AUC to UO 2 powder, the pellets are obtained by direct compacting. In the second experimental method used by CNEA, the yellow cake is dissolved with H 2 SO 4 , and then it is purified with a terciary amine and precipitated with (NH 4 ) 2 CO 3 . In this form the ammonium uranil tri-carbonate (AUT) crystals are obtained. The convertion to UO 2 is made under an atmosphere of dissociated NH 3 . (M.E.L.) [es

Molybdenum-UO2 cerment irradiation at 1145 K

Science.gov (United States)

Mcdonald, G.

1971-01-01

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Dissolution kinetics of UO2: Flow-through tests on UO2.00 pellets and polycrystalline schoepite samples in oxygenated, carbonate/bicarbonate buffer solutions at 25 degree C

International Nuclear Information System (INIS)

Nguyen, S.N.; Weed, H.C.; Leider, H.R.; Stout, R.B.

1991-10-01

The modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO 2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO 2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made. As part of a complex matrix to determine the dissolution kinetics of UO 2 as a function of time, pH, carbonate/bicarbonate concentration and oxygen activity, we have measured the dissolution rates at 25 degrees C of: (1) UO 2 pellets; (2) UO 2.00 powder and (3) synthetic dehydrated schoepite, UO 3 .H 2 O using a single-pass flow through system in an argon-atmosphere glove box. Carbonate, carbonate/bicarbonate, and bicarbonate buffers with concentrations ranging from 0.0002 M to 0.02 M and pH values form 8 to 11 have been used. Argon gas mixtures containing oxygen (from 0.002 to 0.2 atm) and carbon dioxide (from 0 to 0.011 atm) were bubbled through the buffers to stabilize their pH values. 12 refs., 2 tabs

Characterisation and compaction behaviour of UO2 powder prepared from ADU and AUC

International Nuclear Information System (INIS)

Rachmawati, M.

2000-01-01

UO 2 powder prepared from ADU and AUC route process are characterised primarily in terms of compaction and sintering behaviour. Scientific understanding of the phenomena will give useful information leading to processing and product improvement. The investigation has been done by characterising the particle size/shape distribution using SEM and compacting the powder at 4 and 5.4 tons/cm 2 . The behaviour of the powder under compaction is observed by characterizing the pellet length, green density, microstructure, and the compression strength using micrometer SEM, and Universal Testing Machine. The results of the experiment show that the UO 2 powder ex-AUC has particles of spherical type and separate individually which provide the flowable characteristic, important for the die filling aspect during compaction step. The UO 2 powder ex-ADU is more or less agglomerated and contains very fine particles causing the difficulty in pressing. Therefore the green density resulted from UO 2 ex-AUC (6.415 g/cm 3 ) is higher than UO 2 powder of UO 2 ex-ADU (6.117 g/cm 3 . UO 2 at lower pressure (4 tons/cm 3 ) the compression strength ex-AUC green pellet (47.144 kgf) is lower than UO 2 ex-ADU (63,364 kgf), and at higher temperature the compression strength of ex-AUC (92.86 kgf) is higher than UO 2 ex-ADU (82.664 kgf). It is suggested that UO 2 ex-ADU has to be precompacted and granulated in order to increase its flowability so that the pellet length can easily be controlled during pressing (improve reproducibility). (author)

Effect of alpha irradiation on UO2 surface reactivity in aqueous media

International Nuclear Information System (INIS)

Jegou, C.; Muzeau, B.; Broudic, V.; Poulesquen, A.; Roudil, D.; Jorion, F.; Corbel, C.

2005-01-01

The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO 2 matrix in aqueous media subjected to α-β-γ radiation. The β-γ emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO 2 pellets doped with alpha emitters ( 238 Pu and 239 Pu) to quantify the impact of alpha irradiation on UO 2 matrix alteration. Three batches of doped UO 2 pellets with different alpha flux levels (3.30 x 10 4 , 3.30 x 10 5 , and 3.2 x 10 6 α cm -2 s -1 ) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO 2 matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO 2 pellets doped with alpha emitters results in variations of the UO 2 surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO 2 pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO 2 to UO 2+x . However, leaching experiments performed in deaerated media after annealing the samples and preleaching the surface suggest that alpha radiolysis does indeed affect the dissolution, which varies with the

Methods of modification and investigations of properties of fuel UO2

International Nuclear Information System (INIS)

Kurina, I.; Popov, V.; Rogov, S.; Dvoryashin, A.; Serebrennikova, O.

2009-01-01

In the SSC RF-IPPE the researches are carried out directed towards the uranium dioxide fuel pellets modification with the purpose of improvement of their performance parameters (increase of thermal conductivity, growth of grain for decrease gas release, decrease of interaction with coolant). The following technological methods of manufacturing of modified pellets UO 2 were used: 1) The water method including precipitation of Ammonium Polyuranate (APU) with manufacturing of simultaneously coarse and super dispersed particles, and also coprecipitation APU with additives (Cr, Ti, etc.), with the after calcination of powders, reduction to UO 2 pressing and sintering of pellets; 2) A method including addition of chemical reagent containing ammonia to the powder UO 2 manufactured under the dry or water technology; mechanical mixture; pressing and sintering of pellets. Application of the specified up methods makes manufacturing the UO 2 fuel pellets having the properties differing from pellets manufactured by industrial technology. Conclusions: 1) Properties of powders and the pellets manufactured by different technologies considerably differ; 2) Precipitate manufactured by water industrial technology, consists of phase NH 3 ·3UO 3 ·5H 2 O whereas the precipitate manufactured by nanotechnology contains in addition phase NH 3 ·2UO 3 ·3H 2 O; 3) Powders of U 3 O 8 manufactured by water nanotechnology have particles size 300-500 nm and ultra dispersive particles size ∼70-75 nm; 4) Powder UO 2 obtained by water nanotechnology differs by greater activity because all phase changes under oxidation result at lower temperatures; 5) Basic differences of properties of modified UO 2 pellets was established: decreasing of defects inside and on grains boundaries, minor porosity (pore size 0,05-0,5 μm), presence of pore of spherical form, presence of additional chemical bond U-U (presence of metal clusters), polyvalence of U; 6) Methods including addition of Cr and Ti under

Development of an empirical interatomic potential for the Ag–Ti system

Energy Technology Data Exchange (ETDEWEB)

Zhou, Ying, E-mail: y.zhou3@lboro.ac.uk [Department of Mathematical Sciences, Loughborough University, Leicestershire LE11 3TU (United Kingdom); Smith, Roger [Department of Mathematical Sciences, Loughborough University, Leicestershire LE11 3TU (United Kingdom); Kenny, Steven D. [Department of Materials, Loughborough University, Leicestershire LE11 3TU (United Kingdom); Lloyd, Adam L. [Department of Mathematical Sciences, Loughborough University, Leicestershire LE11 3TU (United Kingdom)

2017-02-15

Highlights: • A new modified embedded-atom method interatomic potential for Ag and Ti was developed. • Binding energies for various configurations were calculated using SIESTA and were used as the fitting target. • Two mixing rules for the embedded-atom method based on the same elemental potentials were also investigated. • The results showed that the MEAM with the optimised parameters gives the best agreement to the DFT results. - Abstract: Two interatomic potential mixing rules for the Ti–Ag system were investigated based on the embedded-atom method (EAM) elemental potentials. First principles calculations were performed using SIESTA for various configurations of the Ti–Ag system to see which model best fitted the ab initio results. The results showed that the surface energies, especially that of Ti, were not well fitted by either model and the surface binding energies differed from the ab initio calculations. As a result, the modified embedded-atom method (MEAM) was investigated. In contrast to the other models, surface energies for pure Ti calculated by MEAM were in good agreement with the experimental data and the ab initio results. The MEAM mixing rule was used to investigate Ag ad-atoms on Ti and Ti ad-atoms on Ag. The results showed good agreement with SIESTA after parameter optimisation.

Synthesis and sintering of UN-UO{sub 2} fuel composites

Energy Technology Data Exchange (ETDEWEB)

Jaques, Brian J., E-mail: BrianJaques@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Tyburska-Püschel, Beata [Department of Engineering Physics, University of Wisconsin–Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Meyer, Mitch [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Xu, Peng; Lahoda, Edward J. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

2015-11-15

The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO{sub 2}, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO{sub 2} to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO{sub 2} in a planetary ball mill. UN and UN – UO{sub 2} composite pellets were sintered in Ar – (0–1 at%) N{sub 2} to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO{sub 2} composite pellets were also sintered in Ar – 100 ppm N{sub 2} to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.

Evaluation of the effective thermal conductivity of UO{sub 2} fuel by combining Potts model and finite difference method

Energy Technology Data Exchange (ETDEWEB)

Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, Daedeok-daero 1045, Yuseong, Daejeon 305-353 (Korea, Republic of); Koo, Yang-Hyun; Lee, Byung-Ho; Tahk, Young-Wook [Korea Atomic Energy Research Institute, Daedeok-daero 1045, Yuseong, Daejeon 305-353 (Korea, Republic of)

2011-07-15

This paper evaluated the effects of porosity on the effective thermal conductivity of UO{sub 2} fuel by combining the Potts model and the finite difference method (FDM). Two types of microstructures representing irradiated UO{sub 2} microstructures were simulated by the Potts model in the three dimensional cubic system. One represented very small intragranular bubbles and a few intergranular bubbles under a low temperature condition. The other represented large intergranular bubbles under a high temperature or annealing condition. For the simulated microstructures, the effective thermal conductivities were determined by FDM calculation of the temperature distributions under steady state condition. They were compared with an experimental equation and the effect of bubble morphology was investigated by fitting a porosity shape factor in the Maxwell-Eucken equation. The simulation results showed a good agreement with an experimental equation and demonstrated the capability of the Potts model to provide information on microstructure for calculating the effective thermal conductivity of UO{sub 2} fuel.

Oxidation of UO2 at 150 to 3500C

International Nuclear Information System (INIS)

White, G.D.; Knox, C.A.; Gilbert, E.R.; Johnson, A.B. Jr.

1983-07-01

Oxidation of UO 2 through breached LWR spent fuel rods during interim storage in air atmospheres is a potential mechanism for degradation of cladding integrity. The temperature-time range of published data are inadequate to establish long term behavior under dry storage conditions. Consequently, tests are being conducted in the temperature range of 150 to 350 0 C on unirradiated pellets to evaluate fuel oxidation behavior. The tests have revealed significant-to-minor oxidation at temperatures down to 200 0 C and no measurable oxidation at 150 0 C for times up to 3000 hours. Oxidation at 200 0 C for 2000 hours led to formation of low density particulate U 3 O 8 which destroys pellet integrity. Oxidation of UO 2 pellets at 215 and 250 0 C was signifcantly accelerated by the presence of 1 volume percent NO 2 in the air. NO 2 is a potential constituent of the air, forming by radiolysis in the gamma radiation field associated with spent fuel assemblies. NO 2 reaction with UO 2 pellets leads to accelerated formation of UO 3 and pellet disintegration. 11 references, 15 figures

Preparation of high density (8 to 9) uranium oxide UO2

International Nuclear Information System (INIS)

Eichner, C.; Ertaud, A.; Ortel, Y.; Stohr, J.; Vautrey, L.

1948-10-01

This report describes the process elaborated for the preparation of high density UO 2 . The thermal decomposition of uranium peroxide leads to UO 3 which is reduced by an hydrogen flow to obtain UO 2 . A UO 2 powder of good quality is obtained for temperatures below 650 deg. C. The powder is pulverized to obtain an homogeneous grain size and compressed inside a die to make pellets. Pellets are sintered up to 1600 deg. C in a reducing atmosphere and following a temperature rise law of 150 deg. C/hour. The equipment used (furnaces, gases purifier, control equipment, power supplies, thermoregulation systems) is described at the end. (J.S.)

Simulations and Experimental Measurements of UO2 Thermal Conductivity

International Nuclear Information System (INIS)

Stanek, Christopher Richard; Gofryk, Krzysztof; Tonks, Michael; Andersson, Anders David Ragnar; Liu, Xiang-Yang; Lashley, Jason Charles; Uberuaga, Blas P.; Mcclellan, Kenneth James

2015-01-01

Spin-phonon interactions lead to low @@ of UO 2 (and behave like a defect), and this has implications for nuclear fuel performace. The inability to capture spin-phonon scattering leads to inherent errors. The interplay between magnetism and structural asymmetry in UO 2 displays rich physics. Grain boundary structure plays a role which must be taken into account.

Production and release of the fission gas in (Th U)O2 fuel rods

International Nuclear Information System (INIS)

Dias, Marcio S.

1982-06-01

The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

Contribution of the study of a nuclear reactor accident: residual power aspects and thermodynamic of U-UO_2 and UO_2-ZrO_2 systems

International Nuclear Information System (INIS)

Baichi, Mehdi

2001-01-01

This work is a contribution to the study of early delocalization and fission product releases during the formation of corium coming from a nuclear reactor accident. The first part deals with an analysis of corium cooling. The contribution to the power of each corium element has been calculated with time. The main elements are represented but the addition of Pu, Mo and Nb has been proposed. The last release experimental data taken into account result in a loss of residual power of 25% exclusive of corium between the emergency stop and ten days. The second part deals with the early delocalization observed during Vercors experiments. A critical selection on the U-UO_2 and UO_2-ZrO_2 systems has been carried out. In order to complete the small and inconsistent data, thermodynamic activity measurements have been performed by mass spectrometry. The UO_2 activity on UO_2-ZrO_2 presents a positive deviation from ideality at 2200 K and approximates ideality at 2400 K. All the data have been used for optimizing the systems with Thermo-Calc. This work has allowed to calculate the ternary systems and to define the required approach to analyze the metallic phase and corium oxides densities. (author) [fr

Measurement of thermal conductivity of sintered UO{sub 2} in the reactor; Merenje toplotne provodljivosti sinterovanog UO{sub 2} u reaktoru

Energy Technology Data Exchange (ETDEWEB)

Katanic, J; Stevanovic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1965-10-15

Thermal conductivity is considered one of the fundamental properties of sintered UO{sub 2} fuel. Samples should be tested under real core conditions. This paper covers the methods and instruments for thermal conductivity measurement of UO{sub 2} samples in the reactor core, measurements outside the core under conditions similar to those in the core and outside the core after irradiation. Fuel samples are placed in capsules for irradiation in the reactor in-core loops.

Electron probe micro-analysis of irradiated Triso-coated UO2 particles, (1)

International Nuclear Information System (INIS)

Ogawa, Toru; Minato, Kazuo; Fukuda, Kosaku; Ikawa, Katsuichi

1983-11-01

The Triso-coated low-enriched UO 2 particles were subjected to the post-irradiation electron probe micro-analysis. Observations and analyses on the amoeba effect, inclusions and solutes in the UO 2 matrix were made. In the cooler side of the particle which suffered extensive kernel migration, two significant features were observed: (1) the wake of minute particles, presumably UO 2 , left by the moving kernel in the carbon phase and (2) carbon precipitation in the pores and along the grain boundaries of the UO 2 kernel. Both features could be hardly explained by the gas-phase mechanism of carbon transport and rather suggest the solid state mechanism. Two-types of 4d-transition metal inclusions were observed: the one which was predominantly Mo with a fraction of Tc and another which was enriched with Ru and containing significant amount of Si. The Mo and Si were also found in the UO 2 matrix; the observation led to the discussion of the oxygen potential in the irradiated Triso-coated UO 2 particle. (author)

Fabrication and testing of ceramic UO{sub 2} fuel - I-III. Part I; Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, I Deo

Energy Technology Data Exchange (ETDEWEB)

Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

The task described consists of the following: fabrication of UO{sub 2} with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO{sub 2}; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO{sub 2} powder. This volume includes reports on the first two tasks.

Radiation damage of UO{sub 2} fuel; Radijaciono ostecenje UO{sub 2} goriva

Energy Technology Data Exchange (ETDEWEB)

Stevanovic, M; Sigulinski, F [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

Radiation damage study of fuel and fuel elements covers: study of radiation damage methods in Sweden; analysis of testing the fuel and fuel elements at the RA reactor; feasibility study of irradiation in the Institute compared to irradiation abroad in respect to the reactor possibilities. Tasks included in this study are relater to testing of irradiated UO{sub 2} and ceramic fuel elements.

[UO{sub 2}Cl{sub 2}(phen){sub 2}], a simple uranium(VI) compound with a significantly bent uranyl unit (phen=1,10-phenanthroline)

Energy Technology Data Exchange (ETDEWEB)

Schoene, Sebastian; Radoske, Thomas; Maerz, Juliane; Stumpf, Thorsten; Patzschke, Michael; Ikeda-Ohno, Atsushi [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, Dresden (Germany)

2017-10-04

A simple synthesis based on UO{sub 2}Cl{sub 2}.n H{sub 2}O and 1,10-phenanthroline (phen) resulted in the formation of a new uranyl(VI) complex [UO{sub 2}Cl{sub 2}(phen){sub 2}] (1), revealing a unique dodecadeltahedron coordination geometry around the uranium center with significant bending of the robust linear arrangement of the uranyl (O-U-O) unit. Quantum chemical calculations on complex 1 indicated that the weak but distinct interactions between the uranyl oxygens and the adjacent hydrogens of phen molecules play an important role in forming the dodecadeltahedron geometry that fits to the crystal structure of 1, resulting in the bending the uranyl unit. The uranyl oxygens in 1 are anticipated to be activated as compared with those in other linear uranyl(VI) compounds. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

Structural effects in UO{sub 2} thin films irradiated with U ions

Energy Technology Data Exchange (ETDEWEB)

Popel, A.J., E-mail: apopel@cantab.net [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom); Adamska, A.M.; Martin, P.G.; Payton, O.D. [Interface Analysis Centre, School of Physics, University of Bristol, Bristol BS8 1TL (United Kingdom); Lampronti, G.I. [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom); Picco, L.; Payne, L.; Springell, R.; Scott, T.B. [Interface Analysis Centre, School of Physics, University of Bristol, Bristol BS8 1TL (United Kingdom); Monnet, I.; Grygiel, C. [CIMAP, CEA-CNRS-ENSICAEN-Université de Caen, BP 5133, 14070 Caen Cedex5 (France); Farnan, I. [Department of Earth Sciences, University of Cambridge, Downing Street, Cambridge CB2 3EQ (United Kingdom)

2016-11-01

Highlights: • Quantitative characterisation of radiation damage by kernel average misorientation. • UO{sub 2} (1 1 1) plane showed higher irradiation tolerance than (1 1 0) plane. • UO{sub 2} film-YSZ substrate interface is stable under low fluence irradiation. • (0 0 1), (1 1 0), (1 1 1) single crystal UO{sub 2} thin films on YSZ substrates are expected. - Abstract: This work presents the results of a detailed structural characterisation of irradiated and unirradiated single crystal thin films of UO{sub 2}. Thin films of UO{sub 2} were produced by reactive magnetron sputtering onto (0 0 1), (1 1 0) and (1 1 1) single crystal yttria-stabilised zirconia (YSZ) substrates. Half of the samples were irradiated with 110 MeV {sup 238}U{sup 31+} ions to fluences of 5 × 10{sup 10}, 5 × 10{sup 11} and 5 × 10{sup 12} ions/cm{sup 2} to induce radiation damage, with the remainder kept for reference measurements. It was observed that as-produced UO{sub 2} films adopted the crystallographic orientation of their YSZ substrates. The irradiation fluences used in this study however, were not sufficient to cause any permanent change in the crystalline nature of UO{sub 2}. It has been demonstrated that the effect of epitaxial re-crystallisation of the induced radiation damage can be quantified in terms of kernel average misorientation (KAM) and different crystallographic orientations of UO{sub 2} respond differently to ion irradiation.

INTERATOM experience of cleaning sodium-wetted components

International Nuclear Information System (INIS)

Haubold, W.

1978-01-01

INTERATOM has been concerned since 1967 with the development, testing, and application of methods to clean sodium wetted components by moist nitrogen, vacuum distillation or alcohol. The activities of INTERATOM in this area have been reported at the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' in Dounreay, April 9-12, 1973. The three cleaning methods mentioned above are practised at present, too - with minor modifications - by INTERATOM and in the facilities of the SNR project. This note summarizes the experiences of INTERATOM with methods of sodium removal since 1973

Determination of UO2 little quantity in UF4 by X-rays diffraction

International Nuclear Information System (INIS)

Costa, M.I.; Sato, I.M.; Imakuma, K.

1977-01-01

In the fluorination process, the final product UF 4 contain different levels of UO 2 as a contaminant. A routine method for quantitative analysis by x-ray diffraction has been developed. Standard curves have been plotted using mixtures of UO 2 /UF 4 with measures of intensity of (III) peak of UO 2 by the step scanning process. The integrated intensity versus UO 2 concentration curves present a linear behavior in the range from 0 to 4%. A good reprodutibility of measuring process has been observed through statistical analysis which permits to determine low fractions of UO 2 in UF 4 with +- 0,08% of accuracy [pt

Accumulation of enriched uranium UO2F2 in ultrastructure as studied by electron microscopic autoradiography

International Nuclear Information System (INIS)

Zhu Shoupeng; Wang Yuanchang

1992-01-01

A study was made on the retention of soluble enriched uranium UO 2 F 2 in ultrastructure by electron microscopic autoradiography. The early dynamic accumulation of radioactivity in the body showed that enriched uranium UO 2 F 2 was mainly localized in kidneys, especially accumulated in epithelial cells of proximal convoluted tubules leading to degeneration and necrosis of the tubules. In liver cells, enriched uranium UO 2 F 2 at first deposited in nuclei of the cells and in soluble proteins of the plasma, and later accumulated selectively in mitochondria and lysosomes. On electron microscopic autoradiographic study it was shown that the dynamic retention of radioactivity of enriched uranium UO 2 F 2 in skeleton increased steadily through the time period of exposure. Enriched uranium UO 2 F 2 chiefly deposited in nuclei and mitochondria of osteoblasts as well as of osteoclasts

Correlation between fuel structure and mechanical properties of UO2

International Nuclear Information System (INIS)

Blank, H.; Mandler, R.; Matzke, H.; Routbort, J.; Werner, P.

1982-10-01

The relation between the structure of a UO 2 fuel and its mechanical properties are discussed and illustrated for particular types of UO 2 by measurements of fracture surface energy, hardness, fracture stress and of compressive deformation at 1870 and 1970 0 K. This gives the background for treating the question whether it is possible to find a simple experimental method for correlating the mechanical properties of UO 2 before irradiation with those after various irradiation histories. Hardness measurements might be such a method if combined with a detailed structural analysis and sufficient knowledge about the irradiation history

Uranium migration in spark plasma sintered W/UO2 CERMETS

Science.gov (United States)

Tucker, Dennis S.; Wu, Yaqiao; Burns, Jatuporn

2018-03-01

W/UO2 CERMET samples were sintered in a Spark Plasma Sintering (SPS) furnace at various temperature under vacuum and pressure. High Resolution Transmission Electron Microscopy (HRTEM) with Energy Dispersive Spectroscopy (EDS) was performed on the samples to determine interface structures and uranium diffusion from the UO2 particles into the tungsten matrix. Local Electrode Atom Probe (LEAP) was also performed to determine stoichiometry of the UO2 particles. It was seen that uranium diffused approximately 10-15 nm into the tungsten matrix. This is explained in terms of production of oxygen vacancies and Fick's law of diffusion.

Electrochemical studies of the effect of H2 on UO2 dissolution

International Nuclear Information System (INIS)

King, F.; Shoesmith, D.W.

2004-09-01

This report summarises evidence for the effect of H 2 on the oxidation and dissolution of UO 2 derived from electrochemical studies. In the presence of γ-radiation or with SIMFUEL electrodes containing ε-particles, the corrosion potential (E CORR ) of UO 2 is observed to be suppressed in the presence of H 2 by up to several hundred milli volts. This effect has been observed at room temperature with 5 MPa H 2 (in the case of γ-irradiated solutions) and at 60 deg C with a H 2 partial pressure of only 0.002-0.014 MPa H 2 with the SIMFUEL electrode. The suppression of E CORR in the presence of H 2 indicates that the degree of surface oxidation and the rate of dissolution of UO 2 is lower in the presence of H 2 .The precise mechanism of the effect of H 2 is unclear at this time. The mechanism appears to involve a surface heterogeneous process, rather than a homogeneous solution process. Under some circumstances the value of E CORR approaches the equilibrium potential for the H 2 /H + couple, suggesting galvanic coupling between sites on which this electrochemical process is catalysed and the rest of the UO 2 surface. It is also possible that H* radical species, either produced radiolytically from H 2 O or by dissociation of H 2 on ε-particles or surface-active UO 2+x sites, reduce oxidised U(V)/U(VI) surface states to U(IV). The effect of H 2 on reducing the degree of surface oxidation is only partially reversible, since surfaces reduced in H 2 atmospheres (re-)oxidise more slowly and to a lesser degree than surfaces not previously exposed to H 2 . Homogeneous reactions between dissolved H 2 and either oxidants or dissolved U(VI) cannot explain the observed effects.Regardless of the precise mechanism, the suppression of the degree of surface oxidation results in lower UO 2 dissolution rates in the presence of H 2 . Application of an electro-chemical dissolution model to the observed E CORR values suggests that the fractional dissolution rate of used fuel in the

Geometric dimensioning of UO2 pellets for PWR

International Nuclear Information System (INIS)

Teixeira e Silva, A.

1988-01-01

The finite element structural program SAP-IV is used to calculate UO 2 pellet strains developed under thermal gradients in pressurized water reactors. The applied procedure allows to analyse the influence of various aspects of pellet geometry on cladding strains and can be utilized for the dimensioning of UO 2 pellets. Pellets purchased with flat ends, with dishes pressed into both ends, shouders, and a 45-deg edge chamfer are analysed. The analyse results are compared with experimental data.(autor) [pt

Complexing in (NH4)2SeO4-UO2SeO4 H2O system

International Nuclear Information System (INIS)

Serezhkina, L.B.

1994-01-01

Isotherm of solubility in the (NH 4 ) 2 SeO 4 -UO 2 SeO 4 -H 2 O system has been constructed at 25 deg C. (NH 4 ) 2 (UO 2 ) 2 (SeO 4 ) 3 x6H 2 O formation is established for the first time and certain its physicochemical properties are determined. Regularities of complexing in the R 2 Se) 4 -UO 2 SeO 4 -H 2 O systems, where R-univalent cation are under discussion. 6 refs.; 3 tabs

Preparation of high density (8 to 9) uranium oxide UO{sub 2}; Preparation de l'oxyde d'uranium UO{sub 2} de densite elevee (8 a 9)

Energy Technology Data Exchange (ETDEWEB)

Eichner, C; Ertaud, A; Ortel, Y; Stohr, J; Vautrey, L

1948-10-01

This report describes the process elaborated for the preparation of high density UO{sub 2}. The thermal decomposition of uranium peroxide leads to UO{sub 3} which is reduced by an hydrogen flow to obtain UO{sub 2}. A UO{sub 2} powder of good quality is obtained for temperatures below 650 deg. C. The powder is pulverized to obtain an homogeneous grain size and compressed inside a die to make pellets. Pellets are sintered up to 1600 deg. C in a reducing atmosphere and following a temperature rise law of 150 deg. C/hour. The equipment used (furnaces, gases purifier, control equipment, power supplies, thermoregulation systems) is described at the end. (J.S.)

Fuel elements based on mixed oxides UO{sub 2} - PuO{sub 2}; Gorivni elementi na bazi mesanih oksida UO{sub 2} - PuO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Katanic-Popovic, J; Stevanovic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1978-07-01

Questions concerning utilization of plutonium as a fissionable material in fuel elements for nuclear power plants have been discussed. Characteristics and application of fuel elements with mixed UO{sub 2} - PuO{sub 2} fuel for thermal and fast breeder reactors have also been dealt with. In the presentation of technological processes for production of fuel elements based on mixed oxides specific characteristics are given with respect to the work with plutonium and relatively high production costs as compared to classical fuel elements based on sintered UO{sub 2}. (author)

The treatment of large quantities of high fluorin contents UO2 by ammonium double uranate (ADU) techniques

International Nuclear Information System (INIS)

Wang Bangwu; Chen Ying

2010-01-01

The paper has discussed the sinter action of UO 2 in low temperature. The study indicates the over hot part of UO 2 by the deoxidization hot of oxidation uranate mostly results in the sinter in the process of trans form ADU into UO 2 . The UO 2 settling times in kiln little influences the sinter performance of UO 2 in the same condition of high fluorin contents UO 2 returning kiln, and high fluorin contents UO 2 returning kiln does not sinter UO 2 again. Experiment on large quantities of high fluorin contents UO 2 by Ammonium Double Uranate (ADU) techniques direct returning kiln, the result shows the sinter performance of UO 2 doesn't drop in the process of high fluor in contents UO 2 direct returning kiln, and the performance of UO 2 can meet specification. (authors)

Reinvestigation of the crystal structure of kasolite, Pb[(UO{sub 2})(SiO{sub 4})](H{sub 2}O), an important alteration product of uraninite, UO{sub 2+x}

Energy Technology Data Exchange (ETDEWEB)

Fejfarová, Karla; Dušek, Michal [Institute of Physics ASCR, v.v.i., Na Slovance 2, 18221 Praha (Czech Republic); Plášil, Jakub, E-mail: jakub_plasil@nm.cz [Department of Mineralogy and Petrology, National Museum, Václavské nám. 68, Prague 1, 115 79-CZ (Czech Republic); Institute of Geological Science, Faculty of Science, Masaryk University, Kotlářská 2, CZ-611 37, Brno (Czech Republic); Čejka, Jiří; Sejkora, Jiří [Department of Mineralogy and Petrology, National Museum, Václavské nám. 68, Prague 1, 115 79-CZ (Czech Republic); Škoda, Radek [Institute of Geological Science, Faculty of Science, Masaryk University, Kotlářská 2, CZ-611 37, Brno (Czech Republic)

2013-03-15

The crystal structure of kasolite, Pb[(UO{sub 2})(SiO{sub 4})](H{sub 2}O), Z = 4, monoclinic, with a = 6.7050(3), b = 6.9257(2), c = 13.2857(5) Å, β = 105.064(4)°, V = 595.74(3) Å{sup 3}, the space group P2{sub 1}/c, has been solved by charge-flipping method and refined by the full-matrix least-squares techniques to an agreement factor (R{sub obs}) of 2.2% and, a goodness-of-fit (GOF) of 1.26 using 1243 unique observed diffraction maxima (I{sub obs} > 3σ(I)) collected with MoKα X-radiation and a 4 K CCD area detector. The crystal structure of kasolite contains 1 unique U{sup 6+} position that is part of a nearly linear uranyl ion (UO{sub 2}){sup 2+}, coordinated in the equatorial plane by five O ligands, forming pentagonal bipyramid. The uranyl pentagonal bipyramids share edges to form chains parallel to [0 1 0]. The additional edge of uranyl polyhedra is shared by silicate tetrahedra to form sheets parallel to (1 0 0). There is one unique position of Pb{sup 2+} in the interlayer. O ligands and 1 (H{sub 2}O) non-transformer group coordinate Pb{sup 2+} exhibiting [2 + 6] coordination. A network of H-bonds provides an additional linkage of an interlayer to the sheets besides Pb–O bonds. Chemical composition of the studied crystals, obtained by the electron microprobe, is reported and is in agreement with the crystal structure refinement.

State of the art of UO2 fuel fabrication processes

International Nuclear Information System (INIS)

Henke, M.; Klemm, U.

1980-01-01

Starting from the need of UO 2 for thermal power reactors in the period from 1980 to 1990 and the role of UF 6 conversion into UO 2 within the fuel cycle, the state-of-the-art of the three established industrial processes - ADU process, AUC process, IDR process - is assessed. The number of process stages and requirements on process management are discussed. In particular, the properties of the fabricated UO 2 powders, their influence on the following pellet production and on operational behaviour of the fuel elements under reactor conditions are described. Hence, an evaluation of the three essential conversion processes is derived. (author)

Crystal structure of [UO2(NH35]NO3·NH3

Directory of Open Access Journals (Sweden)

Patrick Woidy

2016-12-01

Full Text Available Pentaammine dioxide uranium(V nitrate ammonia (1/1, [UO2(NH35]NO3·NH3, was obtained in the form of yellow crystals from the reaction of caesium uranyl nitrate, Cs[UO2(NO33], and uranium tetrafluoride, UF4, in dry liquid ammonia. The [UO2]+ cation is coordinated by five ammine ligands. The resulting [UO2(NH35] coordination polyhedron is best described as a pentagonal bipyramid with the O atoms forming the apices. In the crystal, numerous N—H...N and N—H...O hydrogen bonds are present between the cation, anion and solvent molecules, leading to a three-dimensional network.

Molecular dynamics simulation of local structure and vibrational spectrum of uranyl (UO2)2+ in vitreous B2O3

International Nuclear Information System (INIS)

Zhuang, Z.-H.; Liu, G. K.; Beitz, J. V.

2000-01-01

Laser spectroscopic and extended X-ray absorption fine structure (EXAFS) spectra have shown that uranium in B 2 O 3 glass matrix forms uranyl in the electronic configuration of (UO 2 ) 2+ ,but its surrounding structure is not well known. Understanding of uranyl local structure, ion-ligand interaction, and chemical stability on the nanometer scale in glasses is essential in management of long-term performance of high-level nuclear wastes after disposal in a geologic repository. In the present work, the structure, phonon density of states, and vibrational spectrum of vitreous B 2 O 3 and the surrounding environment that contains a uranyl ion have been studied using a molecular dynamics (MD) simulation method that utilizes the Born-Mayer-Huggins and Coulomb pair potentials and the Stillinger-Weber three-body potential. A system of 406 ions was considered in our calculation. Simulation of a thermal quenching from 3000 K to 300 K was performed to generate a uniform and equilibrium model glass matrix before structure configuration and vibrational frequencies were obtained from the system. The structure of the simulated glass is in agreement with that reported by Krogh-Moe and Mozzi et al. The characteristic network of planar boroxol (B 3 O 6 ) rings is evident in the simulated system. A configuration of a U 6+ cation in the vitreous B 2 O 3 matrix is shown in Fig. 1. It is shown that a nearly linear (UO 2 ) 2+ uranyl ion is coordinated by four equatorial oxygen anions in an approximately planar arrangement. The U-O bond length is approximately 0.178 nm for the axial oxygen and 0.254 nm for the equatorial oxygen, which is in good agreement with the U-O distances obtained from fitting EXAFS spectra. Based on the simulated model structure, the uranyl vibrational spectrum is simulated and compared with experimental results obtained using site-selective fluorescence line narrowing (FLN) techniques

Fabrication and testing of ceramic UO2 fuel - I-III. Part I

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

The task described consists of the following: fabrication of UO 2 with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO 2 ; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO 2 powder. This volume includes reports on the first two tasks

Homogenization in powder compacts of UO2-PuO2

International Nuclear Information System (INIS)

Verma, R.

1979-01-01

The homogenization kinetics in mixed UO 2 -PuO 2 compacts have been studied by adopting a concentric core-shell model of diffusion. An equation relating the extent of homogenization expressed in terms of the fraction of UO 2 remaining undissolved and the time of annealing has been derived. From the equation, the periods required at different annealing temperatures to attain a specified level of homogenization have been calculated. These calculated homogenization times have been found to be in fair agreement with the experimentally observed homogenization times. The derived relationship has also been shown to satisfactorily predict homogenization in Cu-Ni powder compacts. (Auth.)

UO2 dissolution rates: A review

International Nuclear Information System (INIS)

McKenzie, W.F.

1992-09-01

This report reviews literature data on UO 2 dissolution kinetics and provides a framework for guiding future experimental studies as well as theoretical modeling studies. Under oxidizing conditions, UO 2 dissolution involves formation of an oxidized surface layer which is then dissolved by formation of aqueous complexes. Higher oxygen pressures or other oxidants are required at higher temperatures to have dissolution rates independent of oxygen pressure. At high oxygen pressures (1-5 atm, 25-70 C), the dissolution rate has a one-half order dependence on oxygen pressure, whereas at oxygen pressures below 0.2 atm, Grandstaff (1976), but nobody else, observed a first-order dependence on dissolution rate. Most people found a first-order dependence on carbonate concentration; Posey-Dowty (1987) found independence of carbonate at pH 7 to 8.2. Dissolution rates increase with temperature except in experiments involving granitic groundwater. Dissolution rates were generally greater under acid or basic conditions than near neutral pH

The uranium recovery from UO{sub 2} kernel production effluent

Energy Technology Data Exchange (ETDEWEB)

Chen, Xiaotong, E-mail: chenxiaotong@tsinghua.edu.cn; He, Linfeng; Liu, Bing; Tang, Yaping; Tang, Chunhe

2016-12-15

Graphical abstract: In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent of HTR spherical fuel elements. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. Based on the above experimental results, the treating flow process in this study would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent of HTR spherical fuel elements. - Highlights: • A flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the UO{sub 2} kernel production effluent. • The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced. • The treating flow process would be feasible for laboratory- and engineering-scale treatment of UO{sub 2} kernel production effluent. - Abstract: For the fabrication of coated particle fuel elements of high temperature gas cooled reactors, the ceramic UO{sub 2} kernels are prepared through chemical gelation of uranyl nitrate solution droplets, which produces radioactive effluent with components of ammonia, uranium, organic compounds and ammonium nitrate. In this study, a flow sheet including evaporation, flocculation, filtration, adsorption, and reverse osmosis was established for the effluent treating. The uranium recovery could reach 99.9% after the treatment, with almost no secondary pollution produced.

Effect of alpha irradiation on UO{sub 2} surface reactivity in aqueous media

Energy Technology Data Exchange (ETDEWEB)

Jegou, C.; Muzeau, B.; Broudic, V.; Poulesquen, A.; Roudil, D. [Commissariat a l' Energie Atomique (CEA), Rhone Valley Research Center, DIEC/SESC/LMPA, Bagnols-sur-Ceze (France); Jorion, F. [Commissariat a l' Energie Atomique (CEA), Rhone Valley Research Center, DRCP/SE2A/LEMA, Bagnols-sur-Ceze (France); Corbel, C. [Commissariat a l' Energie Atomique (CEA), Saclay Research Center, DSM/DRECAM/SCM, Gif sur Yvette (France)

2005-07-01

The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO{sub 2} matrix in aqueous media subjected to {alpha}-{beta}-{gamma} radiation. The {beta}-{gamma} emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO{sub 2} pellets doped with alpha emitters ({sup 238}Pu and {sup 239}Pu) to quantify the impact of alpha irradiation on UO{sub 2} matrix alteration. Three batches of doped UO{sub 2} pellets with different alpha flux levels (3.30 x 10{sup 4}, 3.30 x 10{sup 5}, and 3.2 x 10{sup 6} {alpha} cm{sup -2} s{sup -1}) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO{sub 2} matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO{sub 2} pellets doped with alpha emitters results in variations of the UO{sub 2} surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO{sub 2} pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO{sub 2} to UO{sub 2+x}. However, leaching experiments performed in deaerated media after annealing the samples and

Analysis of flux standards in a fluized bed for AUC - UO2 convertion

International Nuclear Information System (INIS)

Juanico, L.E.; Clausse, A.; Guido Lavalle, G.

1990-01-01

One of the fuel cycle stages is the convertion (reduction) of ammonium uranyl carbonate (AUC) in UO 2 which, after being directly compacted, allows pellet obtainment acquire the correct density to be used as nuclear fuel during sintering. AUC's reduction in UO 2 is made on a fluidized bed in which AUC powder going into the upper part at a countercurrent to the gas flux (superheated steam), is converted into UO 2 ; after the reaction, UO 2 is collected at the lower part of the reactor. (Author) [es

Study on factors affecting sintering density of Gd2O3-UO2 pellets

International Nuclear Information System (INIS)

Zhu Shuming; Zou Congpei; Yang Jing; Yang Youqing; Mei Xiaohui

1996-02-01

The sintered density of Gd 2 O 3 -UO 2 burnable poison fuel pellets is an important quality index and is one of main QC items. Therefore, the efforts were made to investigate the factors affecting the sintered density of Gd 2 O 3 -UO 2 , that is, the influences of pre-treatment of Gd 2 O 3 powder, additives, mixing methods and time, sintering atmosphere, sintering temperature and time on the final density of Gd 2 O 3 UO 2 pellets contained 0, 3%, 7% and 10% (mass percentage) Gd 2 O 3 . The results show: the pre-treatment is useful for improving the distribution of Gd 2 O 3 ; the additive of ammonium oxalate will effectively adjust the density of pellets; 1750 degree C is the suitable sintering temperature. The proper process parameters have been obtained, and the Gd 2 O 3 -UO 2 pellets prepared for in-pile irradiation test meet the design requirements for the density (93.5%∼96.5% of T.D.), homogeneity, microstructure, etc. (8 refs., 3 figs., 8 tabs.)

Variable dimensionality and new uranium oxide topologies in the alkaline-earth metal uranyl selenites AE[UO2)(SeO3)2] (AE=Ca, Ba) and Sr[UO2)(SeO3)2] · 2H2O

International Nuclear Information System (INIS)

Almond, Philip M.; Peper, Shane M.; Bakker, Eric; Albrecht-Schmitt, Thomas E.

2002-01-01

Three new alkaline-earth metal uranyl selenites, Ca[UO 2 )(SeO 3 ) 2 ] (1), Sr[UO 2 )(SeO 3 ) 2 ] · 2H 2 O (2), and Ba[UO 2 )(SeO 3 ) 2 ] (3), have been prepared from the reactions of CaCO 3 and Ca(OH) 2 , SrCl 2 and Sr(OH) 2 , or BaCl 2 and Ba(OH) 2 with UO 3 and SeO 2 under mild hydrothermal conditions. Single-crystal X-ray diffraction experiments reveal that the structures of 1-3 differ in both connectivity and dimensionality even though all contain the same fundamental building unit, namely [UO 2 (SeO 3 ) 4 ]. This polyhedron consists of a linear uranyl unit that is bound by one chelating and three bridging selenite anions creating a pentagonal bipyramidal environment around the U(VI) center. The crystal structure of 1 contains one-dimensional ribbons where the edges are terminated by monodentate selenite anions. The interior of the ribbons are constructed from edge-sharing pentagonal bipyramidal UO 7 units. The structure of 2 is also one-dimensional; however, here there are chains of edge-sharing pentagonal bipyramidal UO 7 dimers that are connected by bridging selenite anions. Ba[(UO 2 )(SeO 3 ) 2 ] (3) is two-dimensional, and the highly ruffled anionic sheets present in this structure are formed from both bridging and chelating/bridging selenite anions bound to uranyl moieties. The anionic substructures in 1-3 are separated by Ca 2+ , Sr 2+ , or Ba 2+ cations. Crystallographic data (193 K, MoKα, λ=0.71073): 1, triclinic, space group P1-bar, a=5.5502(6) A, b=6.6415(7) A, c=11.013(1) A, α=104.055(2) deg., β=93.342(2) deg., γ=110.589(2) deg. , Z=2, R(F)=4.56% for 100 parameters with 1530 reflections with I>2σ(I); 2, triclinic, space group P1-bar, a=7.0545(5) A, b=7.4656(5) A, c=10.0484(6) A, α=106.995(1) deg., β=108.028(1) deg., γ=98.875(1) deg., Z=2, R(F)= 2.43% for 128 parameters with 2187 reflections with I>2σ(I); 3, monoclinic, space group P2 1 /c, a=7.3067(6) A, b=8.1239(7) A, c=13.651(1) A, β=100.375(2) deg., Z=4, R(F)=4.31% for 105 parameters

Construction of the vibrator for UO{sub 2} powder compacting; Izrada vibratora za kompaktiranje praha UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Vrgora, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report contains the description and the scheme of the device for vibration compacting of sintered UO{sub 2} powder. This device was designed and constructed in the Department for reactor materials.

UO{sub 2} Kernel Preparation by M-EG Process and Its Irradiation Test

Energy Technology Data Exchange (ETDEWEB)

Jeong, K. C.; Eom, S. H.; Kim, Y. K.; Yeo, S. H.; Kim, Y. M.; Kim, B. G.; Cho, M. S. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Kernels of KAERI TRISO fuels are prepared in the following steps: (1) preparation of a raw material solution(UN solution) by UO{sub 3} (or U{sub 3}O{sub 8}) powder dissolution in the concentrated HNO{sub 3}; (2) broth preparation and physical property control by mixing UN, THFA, PVA, and H{sub 2}O; (3) preparation of spherical liquid gel droplets and dried-ADU gels in sequence through a reaction between uranyl ions and ammonia ions in a gelation column; (4) ageing, washing, and drying processes of ADU gel using AWD equipment; (5) UO{sub 3} calcination by thermal decomposition of driedADU gel in the air; (6) fabrication of UO{sub 2} kernel by reducing the UO{sub 3} and sintering in the H{sub 2}. In this study, improved KAERI processes for UO{sub 2} kernel preparation were presented. ADU gel washing procedure in AWD processes and the heating mode in sintering process were modified and the internal structures of UO{sub 2} kernels are presented as a result.

SEM hot stage sintering of UO2

International Nuclear Information System (INIS)

Miller, D.J.

1976-06-01

The sintering of hyperstoichiometric uranium dioxide powder compacts, in the hot stage of a scanning electron microscope, was continuously monitored using 16 mm time lapse movies. From alumina microspheres placed on the surface of the compacts, shrinkage measurements were obtained. Converting shrinkage measurements into densification profiles indicates that a maximum densification rate is reached at a critical density, independent of the constant heating rates. At temperatures above 1350 0 C, the movement of the reference microspheres made shrinkage measurements impossible. It is believed the evolution of UO 3 gas from hyperstoichiometric UO 2 is the cause of this limitation

Experimental Determination of the Neutron Characteristics of UO{sub 2}-PuO{sub 2}-H{sub 2}O Lattices; Determination Experimentale Des Caracteristiques Neutroniques De Reseaux UO{sub 2}-PuO{sub 2}-H{sub 2}O

Energy Technology Data Exchange (ETDEWEB)

Debrue, J.; Fabry, A.; Leenders, L.; Motte, F.; Van Den Broeck, H. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

1967-09-15

As part of the investigation, in the VENUS test facility, of the variably moderated core of the BR3/VULCAIN reactor, a fuel assembly consisting of 37 UO{sub 2}-PuO{sub 2} pins (94% natural UO{sub 2}, 6% PuO{sub 2} ) was substituted for one of the enriched (to 7% {sup 235}U) UO{sub 2} fuel assemblies constituting the reactor core. Experiments were carried out with the object of refining the mathematical models for calculating the performance of this special assembly; inter alia, the fission density distribution and the changing ratio of the effective cross-sections for fission in the {sup 233}Pu and {sup 235}U were measured. Using the same critical facility, the authors are carrying out a critical experiment related directly to the problems of plutonium recycling in pressurized light-water thermal reactors. Three types of fuel are being used: UO{sub 2}-PuO{sub 2} with 3% {sup 235}U and 1% fissile plutonium, UO{sub 2}-PuO{sub 2} with 2% {sup 235}U and 2% fissile plutonium, and UO{sub 2} with 4% {sup 235}U. The two UO{sub 2}-PuO{sub 2} mixtures have completely different isotopic contents of {sup 240}Pu: 7% and 17%. In the first part of the experimental programme, a study is being made of regular lattices in cores having two co-axial cylindrical zones: a UO{sub 2}-PuO{sub 2} zone and a UO{sub 2} zone. Particular attention is being paid to investigating the region on either side of the interface separating the two zones, where the neutron spectrum reflects the characteristic energy distributions in each of the two lattices. The experimental results are to be used in calibrating the computational methods. In the second part of the experimental programme, parts of the core of the SENA power reactor will be simulated with a view to studying the problems of reloading one third of the core with mixed UO{sub 2}-PuO{sub 2} fuel. Among the experimental techniques employed in these various experiments emphasis is given to those most specifically related to the presence of

Factors Affecting the Sintering of UO2 Pellets

International Nuclear Information System (INIS)

El-Hakim, E.; Afifi, Y.K.

1999-01-01

Sintering of UO 2 pellets is affected by many parameters such as; UO 2 powder parameters, the conditions followed for preparing the green UO 2 pellets and the sintering scheme(heating and cooling rate, soaking time and temperature). The aim of this work is to study the effect of some these parameters on the characteristics of the sintered UO 2 pellets were qualified according to the technical specifications of Candu fuel. Pressed green pellets at different pressing force (15 to 50 k N) were sintered at 1650 ±20 degree for two hours to study the effect of pressing force on the sintered pellets characteristics; visual inspection, pellet dimensions, density and shrinkage ratio. Compacted green pellets at a pressing force of 48 k N were sintered at different sintering temperature (1600± 20 degree, 1650 ±20 degree, 1700± 20 degree) for two hours to study the effect of sintering temperature on the sintered pellets characteristics. The effect of the heating rate (200,300 and 400 degree per hour) on the sintered pellets characteristics was also investigated. It was found that the pressing force used to compact the green pellets had an effect on the density of the sintered pellets. Pellets pressed at 15 k N have a density of 10.3 g/cm 3 while, those pressed at 50 k N have a density of 10.6 g/cm 3. It was observed that increasing the heating rate to 400 degree /h lead to cracked pellets

The influence of moisture on air oxidation of UO2: Calculations and observations

International Nuclear Information System (INIS)

Taylor, P.; Lemire, R.J.; Wood, D.D.

1993-01-01

Phase relationships among solids in the UO 2 -O 2 -H 2 O system at 25, 100, and 200C and pressures to 2 MPa have been calculated from critically evaluated thermodynamic data. Stability limits of the solids are expressed in terms of oxygen and water partial pressures at each temperature. The results are then discussed in terms of known UO 2 oxidation reactions and uranium mineralogy. Particular attention is paid to UO 3 hydrates, some of which are shown to be stable phases in air at very low relative humidities (down to ∼0.1% at 25C). This is relevant to fuel storage because of the very high molar volumes of these phases, relative to UO 2 , and consequent potential for damage to defected fuel assemblies. Comparison of the calculated phase relationships with observed UO 2 oxidation behavior helps to identify those phase interconversions that are kinetically constrained

Hypoeutectic melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system

Energy Technology Data Exchange (ETDEWEB)

Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Fouquart, Pascal [CEA, DEN, SMTA, LPMA, Cadarache, F13108 St Paul lez Durance (France); Domenger, Renaud; Allegri, Patrick [CEA, DEN, SGCS, LMAC, Marcoule, F30207 Bagnols sur Cèze (France)

2017-05-15

Gadolinium is one of the best neutron absorber materials and its use can be considered as a sacrificial material in a Sodium Fast Reactor core catcher in view of preventing recriticallity. A series of experiments have been conducted in the VITI induction-heated facility to study the melting in the UO{sub 2-x}-Gd{sub 2}O{sub 3} system with 60–87 mol% gadolinia. These experiments have indicated that the eutectic composition is around 92 mol% Gd{sub 2}O{sub 3} – 8 mol% UO{sub 2-x} and that the liquidus line is close to that of Popov et al. [Atom. Energ. 110 (2011) pp. 221–229] phase diagram. - Highlights: •Melting/Solidification experiments with UO{sub 2-x} and Gd{sub 2}O{sub 3} in reducing environment. •Eutectic composition around 92 mol% Gd{sub 2}O{sub 3}-8 mol% UO{sub 2-x}. •UO{sub 2-x} - Gd{sub 2}O{sub 3} liquidus line seems close to that of the pseudobinary phase diagram proposed by Popov et al. •Results will support the assessment of Gd{sub 2}O{sub 3} as a sacrificial material to mitigate criticality risk in SFR core catchers.

Electrochemical Reduction of solid UO2 in Molten Fluoride Salts

International Nuclear Information System (INIS)

Gibilaro, Mathieu; Cassayre, Laurent; Massot, Laurent; Chamelot, Pierre; Malmbeck, Rikard; Dugne, Olivier; Allegri, Patrick

2010-01-01

The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF-CaF 2 (+ 2wt % Li 2 O) at 850 deg. C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, reduction of oxide ions yields O 2 gas evolution on the anode. Electrochemical characterisations of UO 2 pellets have been performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to its direct reduction have been observed at a potential 150 mV more positive in comparison with the solvent reduction. Then, galvano-static electrolyses runs have been realised and products were characterised by SEM-EDX, EPMA/WDS and XRD. In one of the runs, uranium oxide was partially reduced and three phases were observed: non reduced UO 2 in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduction taking place at the grain boundaries. In another run, the UO 2 sample was fully reduced. Due to oxygen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides. (authors)

Near Surface Stoichiometry in UO2: A Density Functional Theory Study

Directory of Open Access Journals (Sweden)

Jianguo Yu

2015-01-01

Full Text Available The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110 surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT calculations. On the basis of the point-defect model (PDM, a general expression for the near surface stoichiometric variation is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300 K with a depth around 3 nm to near-stoichiometric at 1000 K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.

Homogeneity Study of UO2 Pellet Density for Quality Control

International Nuclear Information System (INIS)

Moon, Je Seon; Park, Chang Je; Kang, Kwon Ho; Moon, Heung Soo; Song, Kee Chan

2005-01-01

A homogeneity study has been performed with various densities of UO 2 pellets as the work of a quality control. The densities of the UO 2 pellets are distributed randomly due to several factors such as the milling conditions and sintering environments, etc. After sintering, total fourteen bottles were chosen for UO 2 density and each bottle had three samples. With these bottles, the between-bottle and within-bottle homogeneity were investigated via the analysis of the variance (ANOVA). From the results of ANOVA, the calculated F-value is used to determine whether the distribution is accepted or rejected from the view of a homogeneity under a certain confidence level. All the homogeneity checks followed the International Standard Guide 35

Thermodynamic properties and behaviour of A2[(UO2)(MoO4)2] compounds with A = Li, Na, K, Rb, and Cs

International Nuclear Information System (INIS)

Lelet, Maxim I.; Suleimanov, Evgeny V.; Golubev, Aleksey V.; Geiger, Charles A.; Depmeier, Wulf; Bosbach, Dirk; Alekseev, Evgeny V.

2014-01-01

Highlights: • Low temperature heat capacity of A 2 [(UO 2 )(MoO 4 ) 2 ] (A = Li, Na, K, Rb, and Cs) series was determined. • Enthalpy of formation of Li 2 [(UO 2 )(MoO 4 ) 2 ] was determined by HF solution calorimetry. • Δ f G° (T = 298 K) of all phases from studied series were calculated. - Abstract: A thermodynamic investigation of five alkali-metal uranyl molybdates of the general formula A 2 [(UO 2 )(MoO 4 ) 2 ], where A = Li, Na, K, Rb, and Cs, was undertaken. The various phases were synthesized by solid-state reaction of ANO 3, with A = Li, Na, K, Rb, or Cs, MoO 3 and γ-UO 3 . The synthetic products were characterized by X-ray powder diffraction and X-ray fluorescence methods. The low-temperature heat capacity, S r °, was measured using adiabatic calorimetry from T = (6 to 335) K. Based on these data, the third law entropy at T = 298.15 K, S°, is (345 ± 1) J · K −1 · mol −1 for Li 2 [(UO 2 )(MoO 4 ) 2 ], (373 ± 1) J · K −1 · mol −1 for Na 2 [(UO 2 )(MoO 4 ) 2 ], (390 ± 1) J · K −1 · mol −1 for K 2 [(UO 2 )(MoO 4 ) 2 ], (377 ± 1) J · K −1 · mol −1 for Rb 2 [(UO 2 )(MoO 4 ) 2 ] and (394 ± 1) J · K −1 · mol −1 for Cs 2 [(UO 2 )(MoO 4 ) 2 ]. The enthalpy of formation of Li 2 [(UO 2 )(MoO 4 ) 2 ] was determined using HF solution calorimetry giving Δ f H°(T = 298 K, Li 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3456 ± 9) kJ · mol −1 . Using these new experimental results, together with literature data, the Gibbs free energy of formation of each compound was calculated, giving: Δ f G°(T = 298 K, Li 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3204 ± 9) kJ · mol −1 , Δ f G°(T = 298 K, Na 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3243 ± 2) kJ · mol −1 , Δ f G°(T = 298 K, K 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3269 ± 3) kJ · mol −1 , Δ f G°(T = 298 K, Rb 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3262 ± 3) kJ · mol −1 , and Δ f G°(T = 298 K, Cs 2 [(UO 2 )(MoO 4 ) 2 ], cr) = −(3259 ± 3) kJ · mol −1 . Smoothed S r °(T) values

Effect of water α radiolysis on the spent nuclear fuel UO2 matrix alteration

International Nuclear Information System (INIS)

Lucchini, J.F.

2001-01-01

In the option of long term storage or direct disposal of nuclear spent fuel, it is essential to study the long-term behaviour of the spent fuel matrix (UO 2 ) in water, in presence of ionizing radiations. This work gives some knowledge elements about the impact of aerated water alpha radiolysis on UO 2 alteration. An original experiment method was used in this study. UO 2 /water interfaces were irradiated by an external He 2+ ions beam. The sequential batch dissolution tests on UO 2 samples were performed in aerated deionized water, before, during and after a-irradiation under high fluxes. A corrosion product, identified as hydrated uranium peroxide, was formed on the UO 2 surface. The uranium release was 3 to 4 orders of magnitude higher under irradiation than out of irradiation. The concentrations of the radiolysis products H 2 O 2 and H 3 O + were affected by the uranium oxide surface. They could not only explain the whole uranium release reached during irradiation in water. Leaching experiments on UO X spent fuel samples (with or without the Zircaloy clad) were also performed, in hot cells. The uranium release was relatively small, and H 2 O 2 was not detected in solution. The rates of uranium release in aerated water during one hour were calculated. They were about mg -1 .m -2 .d -1 for spent fuel and for UO 2 , and about g -1 .m -2 .d -1 for UO 2 irradiated by He 2+ ions. The comparison of the results between the two kinds of experiment shows a difference of the behaviour in water between UO 2 irradiated by He 2+ ions and spent fuel. Some hypothesis are given to explain this difference. (author)

Determination of U{sub 3}O{sub 8} in UO{sub 2} by infrared spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Silva, Liliane Aparecida; Lameiras, Fernando Soares; Santos, Ana Maria Matildes dos; Ferraz, Wilmar Barbosa; Barbosa, Joao Batista Santos, E-mail: lasfisica@gmail.com, E-mail: sl@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: jbsb@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), Belo Horizonte, MG (Brazil)

2017-01-15

The oxygen-uranium (O-U) system has various oxides, such as UO{sub 2}, U{sub 4}O{sub 9}, U{sub 3}O{sub 8}, and UO{sub 3}. Uranium dioxide is the most important one because it is used as nuclear fuel in nuclear power plants. UO{sub 2} can have a wide stoichiometric variation due to excess or deficiency of oxygen in its crystal lattice, which can cause significant modifications of its proprieties. O/U relation determination by gravimetry cannot differentiate a stoichiometric deviation from contents of other uranium oxides in UO{sub 2}. The presence of other oxides in the manufacturing of UO{sub 2} powder or sintered pellets is a critical factor. Fourier Transform Infrared Spectroscopy (FTIR) was used to identify U{sub 3}O{sub 8} in samples of UO{sub 2} powder. UO{sub 2} can be identified by bands at 340 cm{sup -1} and 470 cm{sup -1}, and U{sub 3}O{sub 8} and UO{sub 3} by bands at 735 cm{sup -1}, 910 cm{sup -1}, respectively. The methodology for sample preparation for FTIR spectra acquisition is presented, as well as the calibration for quantitative measurement of U{sub 3}O{sub 8} in UO{sub 2}. The content of U{sub 3}O{sub 8} in partially calcined samples of UO{sub 2} powder was measured by FTIR with good agreement with X-rays diffractometry (XRD). (author)

Recovery of UO[sub 2]/PuO[sub 2] in IFR electrorefining process

Science.gov (United States)

Tomczuk, Z.; Miller, W.E.

1994-10-18

A process is described for converting PuO[sub 2] and UO[sub 2] present in an electrorefiner to the chlorides, by contacting the PuO[sub 2] and UO[sub 2] with Li metal in the presence of an alkali metal chloride salt substantially free of rare earth and actinide chlorides for a time and at a temperature sufficient to convert the UO[sub 2] and PuO[sub 2] to metals while converting Li metal to Li[sub 2]O. Li[sub 2]O is removed either by reducing with rare earth metals or by providing an oxygen electrode for transporting O[sub 2] out of the electrorefiner and a cathode, and thereafter applying an emf to the electrorefiner electrodes sufficient to cause the Li[sub 2]O to disassociate to O[sub 2] and Li metal but insufficient to decompose the alkali metal chloride salt. The U and Pu and excess lithium are then converted to chlorides by reaction with CdCl[sub 2].

Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores

International Nuclear Information System (INIS)

Zhao, X.; Driscoll, M.J.; Kazimi, S.

2001-01-01

A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

2012-01-01

CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

Science.gov (United States)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Irradiation behaviour of UO2/Mo porous cermets for thermionic converters

International Nuclear Information System (INIS)

Stora, J.P.; Kauffmann, Y.

1975-01-01

Two types of UO 2 Mo porous cernets have been fabricated and irradiated in a Cythere irradiation device. The first cermet is constituted by little bits of dense fuel in which the two constituants are finely dispersed. The whole open porosity is located between the granules. This type of cermet is called breche (33.4vol%UO 2 , 51vol%Mo, 14.8vol%porosity). At the end of the irradiation the burn up was 19000MWd/t(U) and neither swelling of the cermet nor deformation of the can were noted. On the contrary, a shrinkage of the emitter was observed attributed to a fuel densification under irradiation. The second type of cermet is called macrogranule (36vol%UO 2 , 49vol%Mo 15vol%porosity). UO 2 granules of 0.07cm mean diameter are dispersed in the molybdenum matrix. The porosity is regularly distributed all around the UO 2 kernels. The post irradiation metrology shows that the emitter is fairly stable. Only a slight ovalisation of about 0.5% was noted, but the granules of UO 2 were redistributed inside the molybdenum matrix, overlapping the metallic cavity by a condensation-evaporation process. The matrix has crept into the central void and consequently the volume has grown and the whole porosity has increased from about 15% to about 23%. This creeping is due to the fission gas pressure in the molybdenum cavities after 3000 hours of irradiation. In conclusion two types of cermets have shown good behaviour under irradiation and should allow lifetimes of several thousand hours of operation for thermionic fuel elements [fr

Physical characteristics of Gd2O3-UO2 fuel in LWR

International Nuclear Information System (INIS)

Matsuura, Shojiro; Kobayashi, Iwao; Furuta, Toshiro; Toba, Masao; Tsuda, Katsuhiro.

1981-12-01

A series of critical experiments in light water lattice were carried out on five kinds of Gadolinia-Uranium dioxide (Gd 2 O 3 -UO 2 ) test fuel rods containing 0.0, 0.05, 0.25, 1.50, 3.00 weight % of Gd 2 O 3 in Gd 2 O 3 -UO 2 . Reactivity effect, power distribution, neutron flux distribution, and temperature coefficient were measured for three types of lattices which were in shapes of annular, rectangular parallele-piped, and JPDR mockup core. The theoretical values corresponding to the measured ones were obtained by means of the design method for the FTA which is the test fuel assembly with Gd 2 O 3 -UO 2 rods for JPDR, and the accuracy was checked. In general, the calculated values were in good agreement with the measured ones. Besides, the following characteristics of Gd 2 O 3 -UO 2 rods are recognized both in measurement and calculation, i.e. (1) the effect due to gadolinia on reactivity, power distribution, and thermal neutron flux distribution are steeply saturating; the gadolinia content of only 1.50 weight % is enough to reach the almost saturated condition, (2) the relative power becomes 20% to that of normal fuel under the saturated condition, (3) the relation between the negative reactivity and the power depression effect due to gadolinia is almost linear, and (4) the interference on power depression between the adjacent gadolinia loaded rods is almost negligible, and that on reactivity effect is 15% at most. (author)

Perovskite phases in the systems Asup(II)O-UO/sub 3/. 1. Tetragonal perovskite Ba/sub 2/Basub(7/8)vacantsub(1/8)UO/sub 5/sub(7/8)

Energy Technology Data Exchange (ETDEWEB)

Griffiths, A J; Kemmler-Sack, S [Tuebingen Univ. (Germany, F.R.)

1979-10-01

The new tetragonal compound Ba/sub 2/Basub(7/8)vacantsub(1/8)UO/sub 5/sub(7/8) (a = 2 x 6.31/sub 2/ A; c = 2 x 8.76/sub 7/ A) has been found besides Ba/sub 3/UO/sub 6/ (triclinic) in the BaO-UO/sub 3/ system. It crystallizes with a superstructure of perovskite type. The differences in properties between Ba/sub 3/UO/sub 6/ and Ba/sub 2/Basub(7/8)vacantsub(1/8) UO/sub 5/sub(7/8) are discussed.

Preliminary study of determination of UO2 grain size using X-ray diffraction method

International Nuclear Information System (INIS)

Mulyana, T.; Sambodo, G. D.; Juanda, D.; Fatchatul, B.

1998-01-01

The determination of UO 2 grain size has accomplished using x-ray diffraction method. The UO 2 powder is obtained from sol-gel process. A copper target as radiation source in the x-ray diffractometer was used in this experiment with CμKα characteristic wavelength 1.54433 Angstrom. The result indicate that the UO 2 mean grain size on presintered (temperature 800 o C) has the value 456.8500 Angstrom and the UO 2 mean grain size on sintered (temperature 1700 o C) has value 651.4934 Angstrom

Hydrothermal synthesis, structure, and catalytic properties of UO2Sb2O4

International Nuclear Information System (INIS)

Sykora, Richard E.; King, Joseph E.; Illies, Andreas J.; Albrecht-Schmitt, Thomas E.

2004-01-01

A new uranyl antimonite, UO 2 Sb 2 O 4 (1), has been prepared from the hydrothermal reaction of UO 3 with Sb 2 O 3 and KCl. The structure of 1 consists of neutral two-dimensional ∞ 2 [UO 2 Sb 2 O 4 ] layers. The U(VI) centers are ligated by two trans oxo ligands and four square pyramidal antimonite anions. In addition, the U(VI) also forms long contacts with two additional oxygen atoms that are distorted by 12.7(2) degree sign out of the equatorial plane perpendicular to the uranyl unit. These long interactions are significant owing to evidence supplied by bond valence sum calculations. The two-dimensional layers found in 1 are built from one-dimensional chains formed from edge-sharing UO 6 octahedra that run along the b-axis, and are linked together by [Sb 2 O 4 ] 2- chains. A flow microreactor system has been used to study the catalytic activity of 1, and these results show that it can be used as a catalyst in the conversion of propene and O 2 to acrolein. Crystallographic data: 1, monoclinic, space group C2/m, a=13.490(2) A, b=4.0034(6) A, c=5.1419(8) A, β=104.165(3) deg., Z=2, MoKα, λ=0.71073, R(F)=1.74% for 30 parameters with 365 reflections with I>2σ(I)

Development of UO{sub 2}-Stainless Steel Fuel Plates Containing 30-50 Vol. % Oxide; Fabrication de plaques de combustible en acier inoxydable-UO{sub 2} contenant 30 a 40% d'oxyde (en volume); Razrabotka toplivnykh ehlementov iz nerzhaveyushchej stali i UO{sub 2}, soderzhashchikh 30 - 50 OB.% okisi; Elaboracion de placas de combustible de acero inoxidable UO{sub 2} conteniendo 30 a 40% de oxido (en volumen)

Energy Technology Data Exchange (ETDEWEB)

Lloyd, H. [Atomic Energy Research Establishment, Harwell (United Kingdom)

1963-11-15

This paper describes developments associated with the fabrication of UO{sub 2}-stainless steel plate type fuel elements containing up to 50 vol.% UO{sub 2}. The preparation of high-density spherical UO{sub 2} sintered particles in the 100- to 500-{mu}m size range and the compacting and sintering of cermet plate cores with the particles uniformly distributed in the stainless steel matrix are described together with procedures for hot roll-bonding the fuel plates. Rolling at temperatures up to 1300{sup o}C using total deformations in the 40% to 90% range were studied to establish optimum conditions for the production of high-density cores and to achieve good bonding between the plate components with minimum fragmentation and stringering of the UO{sub 2} particles. The manufacture of large fuel plates utilizing multi-core plates which are bonded together during hot rolling is also described. Data are presented on the mechanical properties of 30, 40 and 50 vol.% UO{sub 2}-stainless steel cermets, prepared as described above, and tested in the as ''rolled'' and annealed condition at various temperatures up to 700{sup o}C, using specimens taken laterally and longitudinally to the direction of rolling. The influence of size and uniformity of distribution of the UO{sub 2} spheres on consistency of mechanical properties are discussed. The strength of bonding between core and cladding for similar cermets in the same temperature range was also assessed. Results are also included of thermal cycling tests between 50 and 800{sup o}C, which was done to study the effects on bond stability and cermet structure after 100, 500 and 1000 cycles. (author) [French] L'auteur expose le processus de fabrication d'elements de combustible UO{sub 2}-Inox en plaques contenant jusqu'a 50% en volume d'UO{sub 2}; il decrit la preparation de particules spheriques de UO{sub 2} frittees de densite elevee (taille dans la gamme de 100 a 500), le pressage et le frittage des plaques de cermet dans

Multi-scale modeling of inter-granular fracture in UO2

Energy Technology Data Exchange (ETDEWEB)

Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonks, Michael R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, S. Bulent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

A hierarchical multi-scale approach is pursued in this work to investigate the influence of porosity, pore and grain size on the intergranular brittle fracture in UO2. In this approach, molecular dynamics simulations are performed to obtain the fracture properties for different grain boundary types. A phase-field model is then utilized to perform intergranular fracture simulations of representative microstructures with different porosities, pore and grain sizes. In these simulations the grain boundary fracture properties obtained from molecular dynamics simulations are used. The responses from the phase-field fracture simulations are then fitted with a stress-based brittle fracture model usable at the engineering scale. This approach encapsulates three different length and time scales, and allows the development of microstructurally informed engineering scale model from properties evaluated at the atomistic scale.

Analysis of UO{sub 2}-BeO fuel under transient using fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), São Paulo, SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-11-01

Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO{sub 2}), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO{sub 2}. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10 - 20Cr, 3 - 5Al, and 0 - 0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO{sub 2}-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO{sub 2}-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO{sub 2}-BeO/Zr, UO{sub 2}-BeO/FeCrAl also were compared with UO{sub 2}/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO{sub 2}/Zr alloys and compared with UO{sub 2}-BeO/FeCrAl. (author)

Measurements of density and of thermal expansion coefficient of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.

1980-12-01

Measurements have been performed of the density and volumetric thermal expansion coefficient of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. These data are required for the design of core-catchers based on sodium borates. The measurements have been performed with the buoyancy method in the temperature range from 850 0 C to 1325 0 C. The data for the pure borax and for the sodium metaborate agree reasonably well with the data from the literature, giving confidence that the measurements are correct and the new data for the salts with UO 2 are reliable. (orig.) [de

Thermal stability test of UO{sub 2}-doped pellet manufactured at INB

Energy Technology Data Exchange (ETDEWEB)

Costa, Diogo R., E-mail: diogoribeiro@inb.gov.br [Indústrias Nucleares do Brasil S.A. (FCN/INB), Resende, RJ (Brazil). Fábrica de Combustível Nuclear; Freitas, Artur C., E-mail: artur.freitas@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

The thermal stability test of UO{sub 2}-doped pellet manufactured at INB was carried out in order to analyze the resintering behavior. This analysis is fundamental for predicting dimensional behavior during irradiation. INB commonly performs resintering test to qualify its production lots, and the same methodology was applied to UO{sub 2}-doped pellets. In this preliminary study, three sets of experiments have been made: 1) without any chemical additive (Z test, the standard UO{sub 2} pellets - undoped); 2) UO{sub 2} pellets doped with 0.1, 0.2 and 0.3 wt% of Al{sub 2}O{sub 3}; and 3) 0.1, 0.2 and 0.3 wt% of Nb{sub 2}O{sub 5}. The preliminary results showed an increase in sintered density in all resintering experiments. So as to obtain the percentage increase, the theoretical densities (g/cm{sup 3} and %TD) were calculated based on the undoped UO{sub 2} pellets. All samples increased in a range of 0.27 to 0.32 %TD the out-pile densification during the resintering process. However, the Z(Nb)3 test showed the lowest value of 0.08 %TD, which is not in agreement with the INB specification limits. The sintered density of this test (0.3 wt% niobia) was 96.15% TD. This fact might be related to the competitive mechanism between Kirkendall effect, forming porosity owing to niobium solubilization on UO{sub 2} matrix, and densification process as a result of uranium diffusivity. Thus, the densification was only 0.08 %TD in Z(Nb)3 sample. All the other samples were in agreement with INB specification. (author)

High temperature interaction between UO2 and Zircaloy-4/silver mixture

International Nuclear Information System (INIS)

Uetsuka, Hiroshi; Nagase, Fumihisa; Otomo, Takashi

1995-12-01

The reaction between UO 2 and Zircaloy is a main material interaction in the reactor core during a severe accident of LWR. With a view of examining the influence of the core materials having low melting temperatures on the reaction, the effect of silver that is main component of PWR control rod alloy was investigated in the temperature range from 1373 to 1703K. Zircaloy was completely liquefied by the same weight of liquid silver at tested temperatures. The reaction between UO 2 and (Zircaloy+silver) mixture roughly obeyed a parabolic rate law. The determined reaction rate below about 1600K was much lower than that obtained by Hofmann et al. for the reaction between UO 2 and Zircaloy. However, it sharply increased with temperature and became comparable with the rate of UO 2 /Zircaloy reaction at about 1700K. Metallurgical examination including EPMA analysis revealed that Zr(O) layer formed at the reaction interface only for the tests below about 1600K correlated with the discontinuity of the temperature dependence of reaction rate. (author)

Bayesian ensemble approach to error estimation of interatomic potentials

DEFF Research Database (Denmark)

Frederiksen, Søren Lund; Jacobsen, Karsten Wedel; Brown, K.S.

2004-01-01

Using a Bayesian approach a general method is developed to assess error bars on predictions made by models fitted to data. The error bars are estimated from fluctuations in ensembles of models sampling the model-parameter space with a probability density set by the minimum cost. The method...... is applied to the development of interatomic potentials for molybdenum using various potential forms and databases based on atomic forces. The calculated error bars on elastic constants, gamma-surface energies, structural energies, and dislocation properties are shown to provide realistic estimates...

Effect of the microstructural morphology on UO{sub 2} powders

Energy Technology Data Exchange (ETDEWEB)

Ziouane, Y.; Lalleman, S.; Leturcq, G. [CEA, Centre de Marcoule, Nuclear Energy Division, RadioChemistry and Processes Department, SERA, LED, F-30207 Bagnols sur Ceze (France); Arab-Chapelet, B. [CEA, Centre de Marcoule, Nuclear Energy Division, RadioChemistry and Processes Department, SERA, LCAR, F-30207 Bagnols sur Ceze (France)

2016-07-01

Several UO{sub 2} powders with different morphologies were synthesized and characterized. Three different morphologies were synthesized thanks to sol-gel process (big heap of about 200 μm wide consisting of sintered crystallites) on the one hand, and to oxalic precipitations (one square platelet morphology and one hexagonal stick morphology) on the other hand. Significant differences in dissolution kinetics were observed. Therefore, the morphology of the powders was found to be a key parameter that has to be considered in the studies of UO{sub 2} dissolution kinetics. The second part of the study consists in dissolving in nitric acid in in the same operating conditions three UO{sub 2} powders having different crystallites sizes. It was shown that dissolution kinetics is dependent on the morphology at the micrometer scale but also on the powder oxygen stoichiometry. (authors)

Corrosion behaviour of the UO2 pellet in corrosive solutions using electrochemical Technique

International Nuclear Information System (INIS)

Taftanzani, A.; Sucipto; Lahagu, F.; Irianto, B.

1996-01-01

The UO 2 electrodes has been made from the local product of UO 2 pellets. The corrosion behaviour of the UO 2 pellets is affected by solution, by pH value and by concentration of salt solution. Investigation into corrosion behaviour of UO 2 electrodes have been carried out in saturated salt solutions using electrochemical technique. The saturated solutions have been made from salts NaCl, Na 2 CO 3 , Na 2 SO 4 and Na 3 PO 4 . The pH value have been done over range 1 pH 10 and the salt concentration (C) over range 0,001 mol/l C 1,0 mol/l, Na 2 CO 3 solution produced the lowest corrosion rates of UO 2 pellets. Those rates were relative constant in the range of pH = 4 - 8. The results indicate an influence of the Na 2 CO 3 concentrations on the corrosions on the corrosion rate, and the lowest rates occur in 0,10 mol/l Na 2 CO 3 . The lowest corrosion rate was 0.3388 mil/year in 0.10 mol/l Na 2 CO 3 by pH = 4. (author)

Correlations between different methods of UO2 pellet density measurement

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1977-07-01

Density of UO 2 pellets was measured by three different methods, i.e., geometrical, water-immersed and meta-xylene immersed and treated statistically, to find out the correlations between UO 2 pellets are of six kinds but with same specifications. The correlations are linear 1 : 1 for pellets of 95% theoretical densities and above, but such do not exist below the level and variated statistically due to interaction between open and close pores. (auth.)

Study of physical properties of UO2 quality improvement result

International Nuclear Information System (INIS)

Rachmat-Pratomo; Hidayati; Didiek Herhady, R; Busron-Masduki

1996-01-01

Activation of uranium dioxide (UO 2 ) by reoxidation to U 3 O 8 and reduction to uranium dioxide (UO 2 ) by temperature reduction variation of 850 o C and 900 o C for 3 hours has been studied. The physical properties before and after treatment are compared. It proved that the oxidation-reduction cycle increased the physical properties. It can be concluded that the reoxidation of UO 2 to U 3 O 8 on fourth cycle and reduction at 900 o C for 3 hours result in a density of 1.32 gram/ml a tap density of 1.60 gram/ml, true density of 9.08 gram/ml and O/U ratio : 2.04. Reduction at 850 o C, for 3 hours result in the bulk density of 1.30 gram/ml, tap density of 1.58 gram/ml, true density of 9.04 gram/ml and O/U ratio 2.09

Fission gas release from the sintered UO{sub 2} fuel; Oslobadjanje fisionih gasova iz goriva od sinterovanog UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Sigulinski, F; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1966-11-15

This paper shoes the phenomena which control fission gases release from the sintered UO{sub 2} dependent of the burnup rate: ejection, release, diffusion, increased fission gas accumulation causing structural changes in the fuel. release of fission gases from the fuel for power reactors was studied as well. The influence of factors as temperature, characteristics of fuel, burnup rate and burnup level was analyzed. Prikazani su mehanizmi koji kontrolisu izdvajanje fisionih gasova iz sinterovanog UO{sub 2} pri razlicitim brzinama izgaranja: izletanje, izbijanje, difuzija, povecano izdvajanje fisionih gasova koje prati strukturne promene u gorivu. Razmatrano je proucavanje izdvajanja fisionih gasova iz goriva za reaktore snage. Analiziran je uticaj faktora kao sto su temperatura, karakteristike goriva, brzina i stepen izgaranja (author)

Cs2SeO4-UO2SeO4-H2O system at 25 deg C

International Nuclear Information System (INIS)

Serezhkina, L.B.; Serezhkin, V.N.

1987-01-01

Using the method of isothermal solubility at 25 deg C the interaction of cesium and uranyl selenates in aqueous solution is studied. Formation of congruently soluble Cs 2 UO 2 (SeO 4 ) 2 x2H 2 O and Cs 2 (UO 2 ) 2 x(SeO 4 ) 3 is ascertained, their crystallographic characteristics being determined

Process for uranium separation and preparation of UO4.2NH3.2HF

International Nuclear Information System (INIS)

Dokuzoguz, H.Z.

1976-01-01

A process for treating the aqueous effluents that are produced in converting gaseous UF 6 (uranium hexafluoride) into solid UO 2 (uranium dioxide) by way of an intermediate (NH 4 ) 4 UO 2 (CO 3 ) 3 (''AUC'' Compound) is disclosed. These effluents, which contain large amounts of NH 4 + , CO 3 2- , F - , and a small amount of U are mixed with H 2 SO 4 (sulfuric acid) in order to expel CO 2 (carbon dioxide) and thereby reduce the carbonate concentration. The uranium is precipitated through treatment with H 2 O 2 (hydrogen peroxide) and the fluoride is easily recovered in the form of CaF 2 (calcium fluoride) by contacting the process liquid with CaO (calcium oxide). The presence of SO 4 2- (sulfate) in the process liquid during CaO contacting seems to prevent the development of a difficult-to-filter colloid. The process also provides for NH 3 recovery and recycling. Liquids discharged from the process, moreover, are essentially free of environmental pollutants. The waste treatment products, i.e., CO 2 , NH 3 , and U are economically recovered and recycled back into the UF 6 → UO 2 conversion process. The process, moreover, recovers the uranium as a precipitate in the second stage. This precipitate is a new inorganic chemical compound UO 4 .2NH 3 .2HF [uranyl peroxide-2-ammonia-2-(hydrogen fluoride)

Irradiation effects in UO2 and CeO2

International Nuclear Information System (INIS)

Ye, Bei; Oaks, Aaron; Kirk, Mark; Yun, Di; Chen, Wei-Ying; Holtzman, Benjamin; Stubbins, James F.

2013-01-01

Single crystal CeO 2 , as a surrogate material to UO 2 , was irradiated with 500 keV xenon ions at 800 °C while being observed using in situ transmission electron microscopy (TEM). Experimental results show the formation and growth of defect clusters including dislocation loops and cavities as a function of increasing atomic displacement dose. At high dose, the dislocation loop structure evolves into an extended dislocation line structure, which appears to remain stable to the high dose levels examined in this study. A high concentration of cavities was also present in the microstructure. Despite high atomic displacement doses, the specimen remained crystalline to a cumulated dose of 5 × 10 15 ions/cm 2 , which is consistent with the known stability of the fluorite structure under high dose irradiation. Kinetic Monte Carlo calculations show that oxygen mobility is substantially higher in hypo-stoichiometric UO 2 /CeO 2 than hyper-stoichiometric systems. This result is consistent with the ability of irradiation damage to recover even at intermediate irradiation temperatures

Mechanism of UO2 selfdisintegration by oxidation

International Nuclear Information System (INIS)

Ohai, D.; Furtuna, I.; Dumitrescu, I.

2008-01-01

Full text: The paper present the results of the study of UO 2 sintered pellets oxidation, part of FIPRED (Fission Product Release from Debris Bed) Project. The FIPRED Project is dedicated to the study the fission products release from irradiated pellets existing in debris bed. The product release is produced by oxidative self disintegration of sintered pellets at air ingress and it depends on temperature. The experimental program covered experiments of 300-1000 deg. C in air diluted with nitrogen at different oxygen concentrations. The experiments were performed using the SETARAM thermo gravimetric equipment and the FIPRED EQ equipment designed and manufactured especially for this type of experiment. The powders (fragments), resulted from UO 2 pellets self disintegration, were characterized by sieving and SEM. The self disintegration mechanism was demonstrated using the experimental results obtained and thermodynamical data of uranium oxides. (authors)

Consistency in thermophysical properties: enthalpy, heat capacity, thermal conductivity and thermal diffusivity of solid UO2

International Nuclear Information System (INIS)

Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

Equations have been derived for the enthalpy, heat capacity, thermal conductivity, and thermal diffusivity of UO 2 . In selection of these equations, we considered the traditional criterion of lowest relative standard deviation between experimental data and the function chosen to fit these data as well as consistency between the thermophysical properties. In the latter case, we considered consistency in (1) thermodynamic relations among properties, (2) the choice of physical phenomena on which to base the theoretical formulation of the equations, and (3) the existence and temperature of phase transitions

Determination of UO2F2, UO2 and UF4 in tetrafluoride of uranium samples

International Nuclear Information System (INIS)

Contreras Guzman, Ariel; Arlegui Hormazabal, Oscar

2003-01-01

The combustible elements for investigation reactors that at the present are manufacturing by the Chilean Nuclear Energy Commission (CCHEN) they are based on aluminum and silicide uranium powdered which is obtained from metallic uranium. At the present the Conversion Units, is developing the technology of transformation UF 6 in metallic Uranium, reason for which is necessary that the Chemical Analysis Laboratory have a methodology that allows to quantify the presence of UO 2 F 2 , UO 2 and UF 4 in the samples obtained in this transformation process. For this reason we are implements the methodology of sequential analysis that had been developed previously, for the Institute of Energy and Nuclear Investigations, IPEN Brasil, and to adapt it to the present conditions in the Laboratory of Chemical Analysis of the CCHEN. This method is based on the different solubilities that present those sample in front of solvents as ethanol and solutions of ammonium oxalate, what allows the separation of these compounds for a later analysis by means of the method of Davies and Gray. This method is based on the reduction of the uranium (VI) to uranium (IV) with ferrous ion amid phosphoric acid, quantifying the present uranium in the samples by means of titration with potassium dicromate. With the purpose of checking the efficiency of the method, the sum of all values of uranium coming from each compound and compares it with the total uranium of the sample (author)

Milestone report: The simulation of radiation driven gas diffusion in UO₂ at low temperature

Energy Technology Data Exchange (ETDEWEB)

Cooper, Michael William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kuganathan, Navaratnarajah [Imperial College, London (United Kingdom); Burr, Patrick A [Univ. of New South Wales (Australia); Rushton, Michael J. [Imperial College, London (United Kingdom); Grimes, Robin W [Imperial College, London (United Kingdom); Turbull, James Anthony [Independent Consultant (United Kingdom); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-10-24

Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. This is an important process for nuclear reactor performance as it affects fission gas release, particularly from the periphery of the pellet where such temperatures are normal. Here we present a molecular dynamics study of Xe and Kr diffusion due to irradiation. Thermal spikes and cascades have been used to study the electronic stopping and ballistic phases of damage, respectively. Our results predict that O and Kr exhibit the greatest diffusivity and U the least, while Xe lies in between. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Preliminary thermal spike calculations indicate that the electronic stopping phase generates greater fission gas displacement than the ballistic phase, although further calculation must be carried out to confirm this. A good description of the system by the empirical potentials is important over the very wide temperatures induced during thermal spike and damage cascade simulations. This has motivated the development of a parameter set for gas-actinide and gas-oxygen interactions that is complementary for use with a recent many-body potential set. A comprehensive set of density functional theory (DFT) calculations were used to study Xe and Kr incorporation at a number of sites in CeO₂, ThO₂, UO₂ and PuO₂. These structures were used to fit a potential, which was used to generate molecular dynamics (MD) configurations incorporating Xe and Kr at 300 K, 1500 K, 3000 K and 5000 K. Subsequent matching to the forces predicted by DFT for these MD configurations was used to refine the potential set. This fitting approach ensured weighted fitting to configurations that are thermodynamically significant over a broad temperature range, while avoiding computationally expensive DFT-MD calculations

Fabrication of metallic channel-containing UO2 fuels

International Nuclear Information System (INIS)

Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

2004-01-01

The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

Technological aspects concerning the production procedures of UO2-Gd2O3 nuclear fuel

International Nuclear Information System (INIS)

Durazzo, Michelangelo; Riella, Humberto Gracher

2007-01-01

The direct incorporation of Gd 2 O 3 powder into UO 2 powder by dry mechanical blending is the most attractive process for producing UO 2 -Gd 2 O 3 nuclear fuel. However, previous experimental results by our group indicated that pore formation due to the Kirkendall effect delays densification and, consequently, diminishes the final density of this type of nuclear fuel. Considering this mechanism as responsible for the poor sintering behavior of UO 2 -Gd 2 O 3 fuel prepared by the mechanical blending method, it was possible to propose, discuss and, in certain cases, preliminarily test feasible adjustments in fabrication procedures that would minimize, or even totally compensate, the negative effects of pore formation due to the Kirkendall effect. This work presents these considerations. (author)

Adsorption of UO2+2 by polyethylene adsorbents with amidoxime, carboxyl, and amidoxime/carboxyl group

International Nuclear Information System (INIS)

Choi, Seong-Ho; Nho, Young Chang

2000-01-01

The polyethylene (PE) adsorbents were prepared by a radiation-induced grafting of acrylonitrile (AN), acrylic acid (AA), and the mixture of AN/AA onto PE film, and by subsequent amidoximation of cyano groups of poly-AN graft chains. With an increase of AA composition in AN/AA monomer mixture, the water uptake of the grafted polyethylene film increased. In AN/AA mixture, the maximum adsorption of UO 2+ 2 was observed in the adsorbent with a ratio of AN/AA (50/50, mol%) in copolymer. The amidoxime, carboxyl, and amidoxime/carboxyl groups onto PE acted as a chelating site for the selected UO 2+ 2 . The complex structure of polyethylene with three functional groups and UO 2+ 2 was confirmed by Fourier Transform Infrared (FTIR) spectroscopy. (author)

Transmission electron microscopic study of reduced Ca2UO5

International Nuclear Information System (INIS)

Krasevec, V.; Prodan, A.; Holc, J.; Kolar, D.

1983-01-01

Structural changes of Ca 2 UO 5 during reduction in hydrogen were studied by transmission electron microscopy. It was shown that monoclinic Ca 2 UO 5 changes into triclinic Ca 4 U 2 O 9 . They are related, respectively, to the fluorite and the bixbyite (C-M 2 O 3 ) structures, so that the product is a superstructure of the latter. Reduction occurs along the (100)/sub t/ planes originating from the (006)/sub m/ planes of the parent structure by diminishing the coordination number of the Ca cation from 7 to 6. 5 figures

Thermal decomposition of (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}·2H{sub 2}O: Influence on structure, microstructure and hydrofluorination

Energy Technology Data Exchange (ETDEWEB)

Thomas, R. [Univ. Lille, CNRS, Centrale Lille, ENSCL, Univ. Artois, UMR 8181 - UCCS - Unité de Catalyse et Chimie du Solide, F-59000 Lille (France); Hall de Recherche de Pierrelatte, AREVA NC, BP 16, 26701 Pierrelatte (France); Rivenet, M., E-mail: murielle.rivenet@ensc-lille.fr [Univ. Lille, CNRS, Centrale Lille, ENSCL, Univ. Artois, UMR 8181 - UCCS - Unité de Catalyse et Chimie du Solide, F-59000 Lille (France); Berrier, E. [Univ. Lille, CNRS, Centrale Lille, ENSCL, Univ. Artois, UMR 8181 - UCCS - Unité de Catalyse et Chimie du Solide, F-59000 Lille (France); Waele, I. de [Université de Lille, CNRS, UMR 8516 – LASIR - Laboratoire de Spectrochimie Infrarouge et Raman, F-59000 Lille (France); Arab, M.; Amaraggi, D.; Morel, B. [Hall de Recherche de Pierrelatte, AREVA NC, BP 16, 26701 Pierrelatte (France); Abraham, F. [Univ. Lille, CNRS, Centrale Lille, ENSCL, Univ. Artois, UMR 8181 - UCCS - Unité de Catalyse et Chimie du Solide, F-59000 Lille (France)

2017-01-15

The thermal decomposition of uranyl peroxide tetrahydrate, (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}.2H{sub 2}O, was studied by combining high temperature powder X-ray diffraction, scanning electron microscopy, thermal analyses and spectroscopic techniques (Raman, IR and {sup 1}H NMR). In situ analyses reveal that intermediates and final uranium oxides obtained upon heating are different from that obtained after cooling at room temperature and that the uranyl precursor used to synthesize (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}·2H{sub 2}O, sulfate or nitrate, has a strong influence on the peroxide thermal behavior and morphology. The decomposition of (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}·2H{sub 2}O ex sulfate is pseudomorphic and leads to needle-like shaped particles of metastudtite, (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}, and UO{sub 3-x}(OH){sub 2x}·zH{sub 2}O, an amorphous phase found in air in the following of (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2} dehydration. (UO{sub 2})O{sub 2}(H{sub 2}O){sub 2}·2H{sub 2}O and the compounds resulting from its thermal decomposition are very reactive towards hydrofluorination as long as their needle-like morphology is kept.

Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

International Nuclear Information System (INIS)

Hofmann, P.

1983-01-01

Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

Effect of additives in sintering UO2-7wt%Gd2O3 fuel pellets

International Nuclear Information System (INIS)

Santos, L.R.; Riella, H.G.

2009-01-01

Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd 2 O 3 powder directly into the UO 2 powder enables longer fuel cycles and optimized fuel utilization. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The process for manufacturing UO 2 - Gd 2 O 3 generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U 3 O 8 to return to the fabrication process. Also, the incorporation of Gd 2 O 3 in UO 2 requires the use of an additive to improve the sintering process, in order to achieve the physical properties specified for the mixed fuel, mainly density and microstructure. This paper describes the effect of the addition of fabrication scraps on the properties of the UO 2 -Gd 2 O 3 fuel. Aluminum hydroxide Al(OH) 3 was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH) 3 as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

Tellurites of hexavalent uranium: first observation of polymerized (UO{sub 4}){sup 2-} tetraoxido cores

Energy Technology Data Exchange (ETDEWEB)

Zadoya, Anastasiya I.; Siidra, Oleg I.; Nazarchuk, Evgeny V.; Bocharov, Sergey N. [Department of Crystallography, St. Petersburg State University (Russian Federation); Bubnova, Rimma S. [Department of Crystallography, St. Petersburg State University (Russian Federation); Institute of Silicate Chemistry, Russian Academy of Sciences, St. Petersburg (Russian Federation)

2016-09-15

Two novel Ca{sub 2}(UO{sub 3})(TeO{sub 3}){sub 2} (1) and K{sub 2}(UO{sub 2}){sub 2}O{sub 2}(TeO{sub 3}) (2) uranyl tellurites were obtained from telluric acid, used as a starting reagent for both compounds. In 1, the tetraoxido core is coordinated by TeO{sub 3} groups and UO{sub 4} squares polymerize into [UO{sub 3}] chains. The tetraoxido core coordination modes in compound 1 are unique. New layered {sub ∞}{sup 2}[(UO{sub 2}){sub 2}(TeO{sub 3})O{sub 2}]{sup 2-} topology is observed for 2. Both of the compounds were studied by the means of high-temperature X-ray diffraction. The thermal decomposition of 1 and 2 is different and leads to formation of uranate compounds. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Orientational anharmonicity of interatomic interaction in cubic monocrystals

International Nuclear Information System (INIS)

Belomestnykh, Vladimir N.; Tesleva, Elena P.

2010-01-01

Anharmonicity of interatomic interaction from a position of physical acoustics under the standard conditions is investigated. It is shown that the measure of anharmonicity of interatomic interaction (Grilneisen parameter) is explicitly expressed through velocities of sound. Calculation results of orientation anharmonicity are shown on the example of 116 cubic monocrystals with different lattice structural type and type of chemical bond. Two types of anharmonicity interatomic interaction anisotropy are determined. Keywords: acoustics, orientational anharmonicity, Gruneisen parameter, velocity of sound

BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry

International Nuclear Information System (INIS)

Rosa, I.; Zara, G.; Guidotti, R.

1974-01-01

1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO 2 and PUO 2 -UO 2 fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed

Complexation in the system K2SeO4-UO2SeO4-H2O

International Nuclear Information System (INIS)

Serezhkina, L.B.; Kuchumova, N.V.; Serezhkin, V.N.

1994-01-01

Complexation in the system K 2 SeO 4 -UO 2 SeO 4 -H 2 O at 25 degrees C is studied by isothermal solubility. Congruently soluble K 2 UO 2 (SeO 4 ) 2 ·4H 2 O (I) and incongruently soluble K 2 (UO 2 ) 2 (SeO 4 ) 3 ·6H 2 O (II) are observed. The unit-cell constants of I and II are determined from an X-ray diffraction investigation. For I, a = 12,969, b = 11.588, c = 8.533 angstrom, Z = 4, space group Pmmb. For II, a = 23.36, b = 6.784, c = 13.699 angstrom, β = 104.42 degrees, Z = 4, space group P2/m, P2, or Pm. Complexes I and II are representatives of the crystal-chemical groups AB 2 2 M 1 and A 2 T 3 3 M 1 , respectively, of uranyl complexes

Ultrasonic analysis of UO{sub 2} pellets

Energy Technology Data Exchange (ETDEWEB)

Bittencourt, Marcelo de S.Q.; Baroni, Douglas B.; Martorelli, Daniel S., E-mail: bittenc@ien.gov.br, E-mail: douglasbaroni@ien.gov.br, E-mail: daniel@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Ultrassom; Dias, Fabio C.; Silva, Jose W.S. da, E-mail: fabio@ird.gov.br, E-mail: wanderley@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Salvaguardas

2013-07-01

Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. In the nuclear area, ceramics are of great importance due to the process of fabrication of fuel pellets for nuclear reactors. Generally, high accuracy destructive techniques are used to characterize nuclear materials for fuel fabrication. These techniques usually require costly equipment and facilities, as well as experienced personnel. This paper aims at presenting an analysis methodology for UO2 pellets using a non-destructive ultrasonic technique for porosity measurement. This technique differs from traditional ultrasonic techniques in the sense it uses ultrasonic pulses in frequency domain instead of time domain. Therefore, specific characteristics of the analyzed material are associated with the obtained frequency spectrum. In the present work, four fuel grade UO2 pellets were analyzed and the corresponding results evaluated. (author)

Preliminary Results on a Contact between 4 kg of Molten UO2 and Liquid Sodium

International Nuclear Information System (INIS)

Amblard, M.

1976-01-01

The CORECT II Experiment consists in simulating the penetration of sodium into an assembly when the fuel is molten. In other words, it is a shock-tube type of experiment with dimensions representative of a full-scale assembly. the experiment consists in dropping a 100 litre column of sodium onto partially molten UO 2 . The following measurements are carried out in transient regime: - sodium velocity in the column; - pressure in the interaction chamber; - pressures at the bottom and at the top of a 5 m tube; - pressure in the argon blanket. The experimental parameters are: - the mass of UO 2 involved (about 4 or 7 kg of 80% molten UO 2 ); - the initial temperature of the sodium (up to 700 deg. C); - the pressure of the residual gas in the interaction chamber during the fall of the sodium; - the dimensions of the interaction chamber and the sodium supply tube; - the form of contact between the UO 2 and the sodium (the sodium may fall on partially liquid and settled UO 2 or on UO 2 pre-dispersed by forced trapping of sodium). To date, 6 tests have been performed. These tests have always resulted in fine fragmentation without any violent interaction. Since no knowledge is available on the change of grain size distribution with time, on the temperature of grain formation, and on the grain movement in the sodium, it is very difficult to interpret these UO 2 -Na tests. We intend to carry out more severe interaction tests on this experimental set-up, by eliminating as much as possible the non-condensable gas which cushions the mechanical impact of the sodium on the UO 2 (tests have shown that by strongly de-pressurizing the liquid UO 2 the fuel could be dispersed by boiling, and this effect should also improve the possibilities of a liquid/liquid contact). - by injecting a little sodium into the UO 2 to facilitate its dispersion in the coolant

Interaction and penetration of heated UO2 with limestone concrete

International Nuclear Information System (INIS)

Farhadieh, R.; Pedersen, D.R.; Purviance, R.; Carlson, N.

1982-01-01

To safeguard the environment against radiological releases, the major question of concern in PAHR safety assessment, following an HCDA, involves confinement and dilution of the molten core-debris. Significant to the study is the directional growth of the core-debris in the concrete foundation of the reactor building or the concrete below the reactor cavity. The real material experiments were carried out in the test apparatus shown. Casts of CRBRP limestone concrete were prepared in graphite cylinders, each having an internal diameter of 8.9 cm and a depth of 30.5 cm. The 17.8-cm-deep concrete samples were allowed to cure for at least 28 days. Experiments were conducted within two months of curing time. The cavity above concrete was packed with 3 kg of pure UO 2 particles (1 to 3 mm). A uranothermic mixture was placed on the top of UO 2 powder. Heating and possible melting of UO 2 was achieved resistively after the ignition of the thermite. Total experimental time was about 60 minutes, during which time a maximum electrical power input of 1.8 watts/gr was applied to the UO 2 . Three gas samples were taken at temperatures of 100, 600, and 950 0 C, measured in the plane of the No. 2 thermocouple. Selection of three temperatures were to study the amount and the type of gases released from different phases of concrete

Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

International Nuclear Information System (INIS)

Linde, A. van der.

1989-04-01

Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

Topologically identical, but geometrically isomeric layers in hydrous α-, β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O and anhydrous Rb[UO2(AsO3OH)(AsO2(OH)2)

International Nuclear Information System (INIS)

Yu, Na; Klepov, Vladislav V.; Villa, Eric M.; Bosbach, Dirk; Suleimanov, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.; Alekseev, Evgeny V.

2014-01-01

The hydrothermal reaction of uranyl nitrate with rubidium nitrate and arsenic (III) oxide results in the formation of polymorphic α- and β-Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )]·H 2 O (α-, β-RbUAs) and the anhydrous phase Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )] (RbUAs). These phases were structurally, chemically and spectroscopically characterized. The structures of all three compounds are based upon topologically identical, but geometrically isomeric layers. The layers are linked with each other by means of the Rb cations and hydrogen bonding. Dehydration experiments demonstrate that water deintercalation from hydrous α- and β-RbUAs yields anhydrous RbUAs via topotactic reactions. - Graphical abstract: Three different layer geometries observed in the structures of Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )] and α- and β- Rb[UO 2 (AsO 3 OH)(AsO 2 (OH) 2 )]·H 2 O. Two different coordination environments of uranium polyhedra (types I and II) are shown schematically on the top of the figure. - Highlights: • Three new uranyl arsenates were synthesized from the hydrothermal reactions. • The phases consist of the topologically identical but geometrically different layers. • Topotactic transitions were observed in the processes of mono-hyrates dehydration

Densification behaviour of UO2 in six different atmospheres

International Nuclear Information System (INIS)

Kutty, T.R.G.; Hegde, P.V.; Khan, K.B.; Basak, U.; Pillai, S.N.; Sengupta, A.K.; Jain, G.C.; Majumdar, S.; Kamath, H.S.; Purushotham, D.S.C.

2002-01-01

The shrinkage behaviour of UO 2 has been studied using a dilatometer in various atmospheres of Ar, Ar-8%H 2 , vacuum, CO 2 , commercial N 2 and N 2 +1000 ppm of O 2 . The onset of shrinkage occurs at around 300-400 deg. C lower in oxidizing atmospheres such as CO 2 , commercial N 2 and N 2 +1000 ppm O 2 compared to that in reducing or inert atmospheres. Shrinkage behaviour of UO 2 is almost identical in Ar, Ar-8%H 2 and vacuum. The shrinkage in N 2 +1000 ppm O 2 begins at a lower temperature than that in the commercial N 2 . The mechanism of sintering in the reducing, inert and vacuum atmospheres is explained by diffusion of uranium vacancies and that in the oxidizing atmospheres by cluster formation

Synthesis and investigation of uranyl molybdate UO2MoO4

International Nuclear Information System (INIS)

Nagai, Takayuki; Sato, Nobuaki; Kitawaki, Shin-ichi; Uehara, Akihiro; Fujii, Toshiyuki; Yamana, Hajimu; Myochin, Munetaka

2013-01-01

In order to examine easily synthetic conditions of uranyl molybdate, UO 2 MoO 4 , used for the reprocessing process study of spent nuclear oxide fuels in alkaline molybdate melts, the uranium molybdate compounds were produced from U 3 O 8 powder and anhydrous MoO 3 reagent. The results of having investigated them in solid state by using X-ray diffractometry and Raman spectrometry, it was confirmed that UO 2 MoO 4 could be synthesized by heating mixed powder of U 3 O 8 and MoO 3 with stoichiometric mole ratio at 770 °C for 4 h under air atmosphere. Moreover, adding this UO 2 MoO 4 into Li 2 MoO 4 -Na 2 MoO 4 eutectic melt, most of the dissolved uranium species in the melt were observed as hexa–valent uranyl ions by absorption spectrophotometry

Fission gas release from ThO2 and ThO2--UO2 fuels (LWBR development program)

International Nuclear Information System (INIS)

Goldberg, I.; Spahr, G.L.; White, L.S.; Waldman, L.A.; Giovengo, J.F.; Pfennigwerth, P.L.; Sherman, J.

1978-08-01

Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO 2 or ThO 2 -UO 2 fuel pellets, with UO 2 compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71 inches and fuel active lengths ranged from 3 to 84 inches. Peak linear power outputs ranged from 2 to 22 kw/ft for peak fuel burnups up to 56,000 MWD/MTM. Measured fission gas release was quite low, ranging from 0.1 to 5.2 percent. Fission gas release was higher at higher temperature and burnup and was lower at higher initial fuel density. No sensitivity to UO 2 composition was evidenced

Role of nitrous acid during the dissolution of UO2 in nitric acid

International Nuclear Information System (INIS)

Deigan, N.; Pandey, N.K.; Kamachi Mudali, U.; Joshi, J.B.

2016-01-01

Understanding the dissolution behaviour of sintered UO 2 pellet in nitric acid is very important in designing an industrial scale dissolution system for the plutonium rich fast reactor MOX fuel. In the current article we have established the role of nitrous acid on the dissolution kinetics of UO 2 pellets in nitric acid. Under the chemical conditions that prevail in a typical Purex process, NO and NO 2 gases gets generated in the process streams. These gases produce nitrous acid in nitric acid medium. In addition, during the dissolution of UO 2 in nitric acid medium, nitrous acid is further produced in-situ at the pellet solution interface. As uranium dissolves oxidatively in nitric acid medium wherein it goes from U(IV) in solid to U(VI) in liquid, presence of nitrous acid (a good oxidizing agent) accelerates the reaction rate. Hence for determining the reaction mechanism of UO 2 dissolution in nitric acid medium, knowing the nitrous acid concentration profile during the course of dissolution is important. The current work involves the measurement of nitrous acid concentration during the course of dissolution of sintered UO 2 pellets in 8M starting nitric acid concentration as a function of mixing intensity from unstirred condition to 1500 RPM

In-pile vapor pressure measurements on UO2 and (U,Pu)O2

International Nuclear Information System (INIS)

Breitung, W.; Reil, K.O.

1985-08-01

The Effective-Equation-of-State (EEOS) experiments investigated the saturation vapor pressures of ultra pure UO 2 , reactor grade UO 2 , and reactor grade (Usub(.77)Pusub(.23))O2 using newly developed in-pile heating techniques. For enthalpies between 2150 and 3700 kJ/kg (about 4700 to 8500 K) vapor pressures from 1.3 to 54 MPa were measured. The p-h curves of all three fuel types were identical within the experimental uncertainties. An assessment of all published p-h measurements showed that the p-h saturation curve of UO 2 appears now well established by the EEOS and the CEA in-pile data. Using an estimate for the heat capacity of liquid UO 2 , the in-pile results were also compared to earlier p-T measurements. The assessments lead to proposal of two equations. Equation I, which includes a factor-of-2 uncertainty band, covers all p-T equilibrium evaporation measurements. Equation I yields 3817 K for the normal boiling point, 415.4 kJ/mol for the corresponding heat of vaporization, and 1.90 MPa for the vapor pressure at 5000 K. Equations I and II, which represent a parametric form of the p-h curve (T=parameter), also give a good description of the EEOS and CEA in-pile data. Thus the proposed equations allow a consistent representation of both p-T and p-h measurements, they are sufficiently precise for CDA analyses and cover the whole range of interest (3120-8500 K, 1400-3700 kJ/kg). (orig./HP) [de

Oxidation of UO2 at 150 to 3500C

International Nuclear Information System (INIS)

Gilbert, E.R.; White, G.D.; Knox, C.A.

1985-02-01

Tests were performed on nonirradiated UO 2 pellets from 150 to 350 0 C in atmospheric air and controlled environments and on spent light-water reactor (LWR) fuel fragments at 200 and 230 0 C in atmospheric air to determine the variables that affect oxidation behavior under dry storage conditions. The weight of spent fragments increased 50 to 100 times faster than the weight of nonirradiated UO 2 pellets at 230 0 C. Non-irradiated pellet fragments gained weight 5 to 7 times faster than nonirradiated pellets. The fragments simulated fuel fragmented by thermal gradients during reactor power changes. Low-density powder (U 3 O 8 ) formed at 0.05 and 0.3% weight gain for nonirradiated pellets and fragments, respectively, but had not formed at 3% weight gain for spent fuel fragments with a burnup of 29,000 MWd/MTU. Canadian investigators had found that powder formed at intermediate levels of weight gain in CANDU spent fuel fragments with an approximate burnup of 8000 MWd/MTU. The combined effects of the high rate of weight gain in spent fuel and the burnup dependence of weight gain at powder formation resulted in a minimum in a plot of the time for the onset of powder formation versus burnup. The minimum in powder induction time occurs at or below burnup levels typical of CANDU spent fuel and spent fuel at the ends of some LWR rods. The results are described in terms of thermal and neutron irradiation-induced changes in UO 2 pellet structure and chemical composition. Other tests were performed at up to 275 0 C with spent fuel fragments and nonirradiated UO 2 pellets in moist nitrogen to determine the suitability of nitrogen as a cover gas. No measurable weight gain or visible physical changes occurred during the first 2 months of testing. 22 figures, 7 tables

Synthesis of nc-UO{sub 2} by controlled precipitation in aqueous phase

Energy Technology Data Exchange (ETDEWEB)

Jovani-Abril, R., E-mail: raqueljovaniabril@gmail.com [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O.Box 2340, D-76125 Karlsruhe (Germany); Gibilaro, M. [Laboratoire de Génie Chimique (LGC), Université de Toulouse, UMR CNRS 5503, 31062 Toulouse cedex 9 (France); Janßen, A.; Eloirdi, R.; Somers, J.; Spino, J.; Malmbeck, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O.Box 2340, D-76125 Karlsruhe (Germany)

2016-08-15

Nanocrystalline UO{sub 2} has been produced through controlled precipitation from an electrolytically reduced U(IV) solution. The reduction process of U(VI) to U(IV) was investigated by cyclic voltammetry in combination with absorption spectrophotometry. Precipitation was achieved by controlled alkalinisation following closely the solubility line of U(IV) in aqueous media. The highest starting concentration used was 0.5 M uranylnitrate which yielded, with the equipment used, around 10 g material pro batch. The produced nc-UO{sub 2} was characterised by transmission electron microscopy (TEM) and x-ray diffraction (XRD) and exhibited the typical UO{sub 2+x} fcc fluorite structure with an average crystallite size of 3.9 nm.

High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

Energy Technology Data Exchange (ETDEWEB)

Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

1965-06-28

Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

International Nuclear Information System (INIS)

Bain, A.S.

1961-08-01

Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

Interactions with Small and Large Sodium to UO2 Mass Ratios

International Nuclear Information System (INIS)

Clerici, G.; Holtbecker, H.; Schins, H.; Schlittenbardt, P.

1976-01-01

This paper is divided into the following three parts: - Presentation of final results of the Ispra dropping experiments; - Discussion of preliminary Na entrapment tests; - Presentation of the Press I and II codes. The experiments for which the Ispra UO 2 dropping facility was originally designed were completed in 1975. The experimental facility which initially had had difficulties in reaching the predefined working conditions gave in the last year a series of results. For this reason Ispra decided to built a similar plant for dropping experiments into water which started working in 1975. Concerning the entrapment tests it was originally foreseen to built in collaboration with GfK Karlsruhe a test section having subassembly geometry and in which the UO 2 would have been violently dispersed into the surrounding Na by the expansion of a small quantity of superheated sodium. Preliminary tests and the design work for the facility could be completed. The Press I + II codes were developed to support the above mentioned experiment - al activity. A 1-D analysis is made to investigate phenomena like UO 2 crust formation and calculate delay times between the time of the Na injection into UO 2 and the violent expansion of superheated Na. An estimate was also made of the available mechanical work in such a process which should allow to get an idea of possible energy release in a reactor core. First conclusions can be drawn from this estimate concerning the mechanical energy release in a WCA due to SPI. The result is that considerably lower energies are calculated from Na entrapment in a reactor core due to the limited amount of molten UO 2 present in the core

In-Situ Observation of Sintering Shrinkage of UO2 Compacts Derived from Different Powder Routes

International Nuclear Information System (INIS)

Rhee, Young Woo; Oh, Jang Soo; Kim, Dong Joo; Kim, Keon Sik; Kim, Jong Hun; Yang, Jae Ho; Koo, Yang Hyun

2015-01-01

In-situ observations on the shrinkage of green pellets with precisely controlled dimensions were carefully conducted by using TOM during H2 atmosphere sintering. The shrinkage retardation in IDR-UO 2 might be attributed to the larger primary particle size of IDRUO 2 than those of ADU- and AUC- UO 2 powders. It would be important to understand the different sintering characteristics of UO 2 powders according to the powder routes, when it comes to designing a new sintering process or choosing a sintering additive for new fuel pellet like PCI (Pellet Cladding Interaction) remedy pellet. In this paper, we have investigated the initial and intermediate sintering shrinkage of UO 2 from different powder routes by in-situ observation of green samples during H2 atmosphere sintering. Effect of powder characteristics of three different UO 2 powders on the initial and intermediate sintering were closely reviewed including crystal structure, powder size, specific surface area, primary crystal size, and O/U ratio

Growth of Gd{sub 2}O{sub 3} coherent layers on UO{sub 2} microsphere surface via sol-gel process

Energy Technology Data Exchange (ETDEWEB)

Ribeiro, Luciana S.; Silva, Edilaine F.; Oliveira, Felipe W.F.; Pereira, Yara S.; Brandão, Alisson F.C.; Santos, Ana Maria M.; Lameiras, Fernando S.; Reis, Sergio C.; Pedrosa, Tércio A.; Santos, Armindo, E-mail: santosa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

In this work, we synthesized and characterized UO{sub 2}-Gd{sub 2}O{sub 3} nuclear fuel via three routes, aiming to solve the problems arising from the addition of Gd{sub 2}O{sub 3} in UO{sub 2} matrix. By the industrial route, the mixture of powders (UO{sub 2}, <90 μ and 6 wt% Gd{sub 2}O{sub 3} <10 μm) results in pellets with 91% TD at 1677 °C/H{sub 2}/4 h. By the mixed route, the formation of Gd{sub 2}O{sub 3} coherent layers on UO{sub 2} powder (particles <90 μ) and microsphere (225 μm) surface produced UO{sub 2} - 6 wt% Gd{sub 2}O{sub 3} pellets with 95% (powder; 1625 °C/H{sub 2}/4 hr) and 83% (microsphere; 1677°C/H{sub 2}/4 hr) TD. By the sol-gel route, we obtained UO{sub 2} - 6 wt% Gd{sub 2}O{sub 3} in a deagglomerated (powder; <70 μm) or agglomerated microsphere 232 μm) form whose pellets reached > 97% (powder) and >98% (microsphere) TI) at 1677 °C/H{sub 2}/4h. According to XRD, OM, and SEM/EDS analysis, the referred three routes do not form a complete solid solution of UO{sub 2}-Gd{sub 2}O{sub 3} at the temperatures and time of sintering used; Gd{sub 2}O{sub 3} granule islands are present in the pellets originating from these routes. The obtained results suggest that the topological arrangement and the deficient nanostructuring of UO{sub 2} and Gd{sub 2}O{sub 3} phases, either in the raw material (powder and microsphere) as in their compacts, are the cause of low densification and irregular distribution of Gd{sub 2}O{sub 3} in UO{sub 2} matrix; mixing of U and Gd at the molecular level does not form a solid solution; and the mixed route is a good alternative to the industrial route. (author)

An improved UO2 thermal conductivity model in the ELESTRES computer code

International Nuclear Information System (INIS)

Chassie, G.G.; Tochaie, M.; Xu, Z.

2010-01-01

This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)

Effects of MnO-Al2O3 on the grain growth and high-temperature deformation strain of UO2 fuel pellets

International Nuclear Information System (INIS)

Kang, Ki Won; Yang, Jae Ho; Kim, Jong Hun; Rhee, Young Woo; Kim, Dong Joo; Kim, Keon Sik; Song, Kun Woo

2010-01-01

The fabrication and high-temperature deformation strain of MnO-Al 2 O 3 -doped UO 2 pellets were studied. The effects of additive composition and amount on the microstructure evolution of a UO 2 pellet were investigated. The compressive creep behaviors of MnO-Al 2 O 3 -doped UO 2 pellets were examined. The results indicated that a MnO-Al 2 O 3 binary additive can effectively promote the grain growth of UO 2 pellets. In addition, the high-temperature deformation strain of the UO 2 pellet can be improved significantly with 1,000 ppm 95MnO-5Al 2 O 3 (mol%). The developed MnO-Al 2 O 3 -additive-containing UO 2 pellets can be a potential candidate for a high-burn-up fuel and a pellet-cladding interaction (PCI) remedy. (author)

Modeling conversion of ammonium diuranate (ADU) into uranium dioxide (UO{sub 2}) powder

Energy Technology Data Exchange (ETDEWEB)

Hung, Nguyen Trong; Thuan, Le Ba [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Khoai, Do Van [Institute for Technology of Radioactive and Rare Elements (ITRRE), 48 Lang Ha, Dong Da, Ha Noi (Viet Nam); Current Postdoctoral Fellow at Tokai Reprocessing Technology Development Center, Japan Atomic Energy Agency (JAEA), 4-33 Tokaimura, Nakagun, Ibaraki, 319-1194 (Japan); Lee, Jin-Young, E-mail: jylee@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of); Jyothi, Rajesh Kumar, E-mail: rkumarphd@kigam.re.kr [Convergence Research Center for Development of Mineral Resources (DMR), Korea Institute of Geoscience and Mineral Resources (KIGAM), Daejeon, 34132 (Korea, Republic of)

2016-10-15

In the paper, Brandon mathematical model that describes the relationship between the essential fabrication parameters [reduction temperature (T{sub R}), calcination temperature (T{sub C}), calcination time (t{sub C}) and reduction time (t{sub R})] and specific surface area of ammonium diuranate (ADU)-derived UO{sub 2} powder products was established. The proposed models can be used to predict and control the specific surface area of UO{sub 2} powders prepared through ADU route. Suitable temperatures for conversion of ADU and ammonium uranyl carbonate (AUC) was examined with the proposed model through assessment of the sinterability of UO{sub 2} powders.

On the Role of the Electrical Field in Spark Plasma Sintering of UO2+x

Science.gov (United States)

Tyrpekl, Vaclav; Naji, Mohamed; Holzhäuser, Michael; Freis, Daniel; Prieur, Damien; Martin, Philippe; Cremer, Bert; Murray-Farthing, Mairead; Cologna, Marco

2017-01-01

The electric field has a large effect on the stoichiometry and grain growth of UO2+x during Spark Plasma Sintering. UO2+x is gradually reduced to UO2.00 as a function of sintering temperature and time. A gradient in the oxidation state within the pellets is observed in intermediate conditions. The shape of the gradient depends unequivocally on the direction of the electrical field. The positive surface of the pellet shows a higher oxidation state compared to the negative one. An area with larger grain size is found close to the positive electrode, but not in contact with it. We interpret these findings with the redistribution of defects under an electric field, which affect the stoichiometry of UO2+x and thus the cation diffusivity. The results bear implications for understanding the electric field assisted sintering of UO2 and non-stoichiometric oxides in general. PMID:28422164

Reduction behavior of UO{sub 2}{sup 2+} in molten LiCl–RbCl and LiCl–KCl eutectics by using tungsten

Energy Technology Data Exchange (ETDEWEB)

Nagai, Takayuki, E-mail: nagai.takayuki00@jaea.go.jp [Nuclear Fuel Cycle Engineering Lab., Japan Atomic Energy Agency, Muramatsu, Tokai, Naka, Ibaraki 319-1194 (Japan); Uehara, Akihiro; Fujii, Toshiyuki; Yamana, Hajimu [Research Reactor Institute, Kyoto University, Asashironishi, Kumatori, Sen-nan, Osaka 590-0494 (Japan)

2013-08-15

The reduction of uranium from UO{sub 2}{sup 2+} to UO{sub 2}{sup +} or U{sup 4+} in molten LiCl–RbCl and LiCl–KCl eutectics was examined by using tungsten and chlorine gas. Spectrophotometric technique was adopted to determine the concentration of uranium species. When tungsten was immersed into the LiCl–RbCl eutectic melt at 400 °C without supplying chlorine gas, 36% of the total weight of the hexavalent of UO{sub 2}{sup 2+} was reduced to the pentavalent of UO{sub 2}{sup +}. Under purging chlorine gas into the melt, 96% of UO{sub 2}{sup 2+} was reduced to the tetravalent of U{sup 4+}. Tungsten oxy-chloride of WOCl{sub 4} was produced via the reductions of UO{sub 2}{sup 2+}, which was volatized from the melt and adsorbed on the upper part of experimental cell. On the other hand, 84% of UO{sub 2}{sup 2+} in the LiCl–KCl eutectic melt at 500 °C was reduced to U{sup 4+} by using tungsten and chlorine gas.

Hydrothermal synthesis and crystal structures of new uranyl oxalate hydroxides: α- and β-[(UO2)2(C2O4)(OH)2(H2O)2] and [(UO2)2(C2O4)(OH)2(H2O)2].H2O

International Nuclear Information System (INIS)

Duvieubourg, Laurence; Nowogrocki, Guy; Abraham, Francis; Grandjean, Stephane

2005-01-01

Two modifications of the new uranyl oxalate hydroxide dihydrate [UO 2 ) 2 (C 2 O 4 )(OH) 2 (H 2 O) 2 ] (1 and 2) and one form of the new uranyl oxalate hydroxide trihydrate [(UO 2 ) 2 (C 2 O 4 )(OH) 2 (H 2 O) 2 ].H 2 O (3) were synthesized by hydrothermal methods and their structures determined from single-crystal X-ray diffraction data. The crystal structures were refined by full-matrix least-squares methods to agreement indices R(wR)=0.0372(0.0842) and 0.0267(0.0671) calculated for 1096 and 1167 unique observed reflections (I>2σ(I)), for α (1) and β (2) forms, respectively and to R(wR)=0.0301(0.0737) calculated for 2471 unique observed reflections (I>2σ(I)), for 3. The α-form of the dihydrate is triclinic, space group P1-bar , Z=1, a=6.097(2), b=5.548(2), c=7.806(3)A, α=89.353(5), β=94.387(5), γ=97.646(5) o , V=260.88(15)A 3 , β-form is monoclinic, space group C2/c, Z=4, a=12.180(3), b=8.223(2), c=10.777(3)A, β=95.817(4), V=1073.8(5)A 3 . The trihydrate is monoclinic, space group P2 1 /c, Z=4, a=5.5095(12), b=15.195(3), c=13.398(3)A, β=93.927(3), V=1119.0(4)A 3 . In the three structures, the coordination of uranium atom is a pentagonal bipyramid composed of dioxo UO 2 2+ cation perpendicular to five equatorial oxygen atoms belonging to one bidentate oxalate ion, one water molecule and two hydroxyl ions in trans configuration in 2 and in cis configuration in 1 and 3. The UO 7 polyhedra are linked through hydroxyl oxygen atoms to form different structural building units, dimers [U 2 O 10 ] obtained by edge-sharing in 1, chains [UO 6 ] ∼ and tetramers [U 4 O 26 ] built by corner-sharing in 2 and 3, respectively. These units are further connected by oxalate entities that act as bis-bidentate to form one-dimensional chains in 1 and bi-dimensional network in 2 and 3. These chains or layers are connected in frameworks by hydrogen-bond arrays

Thermal conductivity of sintered UO2 under in-pile conditions

International Nuclear Information System (INIS)

Stora, J.P.; Bernardy De Sigoyer, B.; Delmas, R.; Deschamps, P.; Lavaud, B.; Ringot, C.

1964-01-01

The temperature distribution in a stack of sintered UO 2 cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO 2 cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k a (T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k a (T) ≅ k(T). In the domain of phonon conduction (T ≤ 1350 deg C), the expression kw.cm -1 .C -1 =1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm 2 is insufficient to restore completely the thermal conductivity of the fuel. (authors) [fr

Crystal field levels of tetravalent actinide ions in actinide dioxides UO2, NpO2 and PuO2

International Nuclear Information System (INIS)

Krupa, J.C.; Gajek, Z.

1991-01-01

Crystal-field parameters resulting from analysis of optical spectroscopy and neutron diffraction data recorded on UO 2 and NpO 2 as well as ab-initio calculated parameters were used to calculate the crystal-field eigenfunctions and eigenvalues for the J ground-state manifold of U 4+ , Np 4+ and Pu 4+ in UO 2 , NpO 2 and PuO 2

An interatomic potential for studying CuZr bulk metallic glasses

International Nuclear Information System (INIS)

Paduraru, A.; Kenoufi, A.; Bailey, N.P.; Schioetz, J.

2007-01-01

Glass forming ability has been found in only a small number of binary alloys, one being CuZr. In order to simulate this glass, we fitted an interatomic potential within Effective Medium Theory (EMT). For this purpose we use basic properties of the B2 crystal structure as calculated from Density Functional Theory (DFT) or obtained from experiments. We then performed Molecular Dynamics (MD) simulations of the cooling process and studied the thermodynamics and structure of CuZr glass. We find that the potential gives a good description of the CuZr glass, with a glass transition temperature and elastic constants close to the experimental values. The local atomic order, as witnessed by the radial distribution function, is also consistent with similar experimental data. (Abstract Copyright [2007], Wiley Periodicals, Inc.)

Densities and apparent molar volumes for aqueous solutions of HNO3-UO2(NO3)2 at 298.15 K

International Nuclear Information System (INIS)

Yang-Xin Yu; Tie-Zhu Bao; Guang-Hua Gao; Yi-Gui Li

1999-01-01

In order to obtain the exact information of atomic number density in the ternary system of HNO 3 -UO 2 (NO 3 ) 2 -H 2 O, the densities were measured with an Anton-Paar DMA60/602 digital density meter thermostated at 298.15±0.01 K. The apparent molal volumes for the systems were calculated from the experimental data. The present measured apparent molar volumes have been fitted to the Pitzer ion-interaction model, which provides an adequate representation of the experimental data for mixed aqueous electrolyte solutions up to 6.2 mol/kg ionic strength. This fit yields θ V , and Ψ V , which are the first derivatives with respect to pressure of the mixing interaction parameters for the excess free energy. With the mixing parameters θ V , and ψ V , the densities and apparent molar volumes of the ternary system studied in this work can be calculated with good accuracy, as shown by the standard deviations. (author)

Acoustic emission from thermal-gradient cracks in UO2

International Nuclear Information System (INIS)

Kennedy, C.R.; Kupperman, D.S.; Wrona, B.J.

1975-01-01

A feasibility study has been conducted to evaluate the potential use of acoustic emission to monitor thermal-shock damage in direct electrical heating of UO 2 pellets. In the apparatus used for the present tests, two acoustic-emission sensors were placed on extensions of the upper and lower electrical feedthroughs. Commercially available equipment was used to accumulate acoustic-emission data. The accumulation of events displayed on a cathode-ray-tube screen indicates the total number of acoustic-emission events at a particular location within the pellet stack. These tests have indicated that acoustic emission can be used to monitor thermal-shock damage in UO 2 pellets subjected to direct-electrical heating. 8 references

Automation system for production of UO2 granules

International Nuclear Information System (INIS)

Swaminathan, N.; Setty, C.R.P.; Banerjee, P.K.; Husnain, G.; Rao, K.C.M.; Satyanarayana, A.

1990-01-01

Precompaction of UO 2 powder into slugs and granulation of the slugs were used to be carried out in two different work centres involving manual loading/handling of powder and compacts which resulted in a very high level of air-borne activity. This has been simplified by integrating both the operations into one work centre on both the precompaction presses. In the present system, UO 2 powder is transferred to feed hopper through the use of high vac. feeder. The powder in metered quantities is fed into the shoe by deploying screw feeder driven by a compact hydraulic motor. The die cavity is filled with just the right quantity of powder to prevent spillage. The compacts are pushed on to the granulator through a set of guides mounted on the die platform. The granulated powder is made to pass through Vibro screen for separating the fines before collecting in a replaceable S.S. Container. This container is mounted on the final compacting press by using job crane installed on the press. The replaceable container handling facility drastically cuts down the manual handling of UO 2 granules and also eliminates spillage, air borne activity. The development and fabrication of hydraulically operated screw feeder, feed shoe, replaceable container and the job crane structure etc., were completely carried out at Nuclear Fuel Complex, Hyderabad. Paper deals in detail the design of the system developed, present operational experiences and further improvements planned. (author). 6 figs

Interatomic inelastic current

DEFF Research Database (Denmark)

Hansen, Tim; Solomon, Gemma C.; Hansen, Thorsten

2017-01-01

In order to identify the location of an inelastic event and to distinguish between situations that are before or after this event, we derive equations for the interatomic inelastic transmission as a perturbation series in the electron-phonon interaction. This series contains both even and odd...... to second order and the 1st order correction represents the lowest order term of this new family of terms. We apply this to three model systems and are able to distinguish between situations before and after the inelastic event as steps in the 2nd order transmission. We also see that when the transmission...

Temperature dependence in interatomic potentials and an improved potential for Ti

International Nuclear Information System (INIS)

Ackland, G J

2012-01-01

The process of deriving an interatomic potentials represents an attempt to integrate out the electronic degrees of freedom from the full quantum description of a condensed matter system. In practice it is the derivatives of the interatomic potentials which are used in molecular dynamics, as a model for the forces on a system. These forces should be the derivative of the free energy of the electronic system, which includes contributions from the entropy of the electronic states. This free energy is weakly temperature dependent, and although this can be safely neglected in many cases there are some systems where the electronic entropy plays a significant role. Here a method is proposed to incorporate electronic entropy in the Sommerfeld approximation into empirical potentials. The method is applied as a correction to an existing potential for titanium. Thermal properties of the new model are calculated, and a simple method for fixing the melting point and solid-solid phase transition temperature for existing models fitted to zero temperature data is presented.

Retrieval of interatomic separations of molecules from laser-induced high-order harmonic spectra

International Nuclear Information System (INIS)

Le, Van-Hoang; Nguyen, Ngoc-Ty; Jin, C; Le, Anh-Thu; Lin, C D

2008-01-01

We illustrate an iterative method for retrieving the internuclear separations of N 2 , O 2 and CO 2 molecules using the high-order harmonics generated from these molecules by intense infrared laser pulses. We show that accurate results can be retrieved with a small set of harmonics and with one or few alignment angles of the molecules. For linear molecules the internuclear separations can also be retrieved from harmonics generated using isotropically distributed molecules. By extracting the transition dipole moment from the high-order harmonic spectra, we further demonstrated that it is preferable to retrieve the interatomic separation iteratively by fitting the extracted dipole moment. Our results show that time-resolved chemical imaging of molecules using infrared laser pulses with femtosecond temporal resolutions is possible

Behaviour of high purity UO2/H2O interfaces under helium beam irradiation in deaerated conditions

International Nuclear Information System (INIS)

Mendes, E.

2005-11-01

A question put within the framework of the nuclear fuel storage worn in geological site is what become to them in the presence of water. The aim of a fundamental program, of PRECCI project (ECA), is to highlight the behaviour of interfaces which can be used as models for the interfaces nuclear spent fuel/water if the fuel is uranium UO 2 dioxide. This doctorate is interested in the effect of the alpha activity which is the only one that exist in the spent fuel after long periods. The aim is to identify the mechanisms of alteration and of leaching of surfaces under alpha irradiation. A method is developed to irradiate UO 2 /H 2 O interfaces in deaerated conditions with the beam of He 2+ produced by a cyclotron. The He 2+ ions cross an UO 2 disc and emerge in water with an energy of 5 MeV. Leachings under irradiation are carried with a large range of particles flux. The post-irradiation characterization of the surface of the discs realised by micro-Raman spectroscopy allowed to identify the alteration layer. It is made up of studtite UO 2 (O 2 ),4H 2 O, and of schoepite UO 3 ,xH 2 O. The analysis of the solutions shows that the uranium release strongly increases. The electrochemical properties of the interfaces under irradiation strongly differ from those before irradiation. This work allows to propose that the radiolytic species seen by the interface are it during the heterogeneous phase of evolution of the traces and are species of short lives. Modeling show that the radiolytic radicals species can migrate toward the interface and react with the UO 2 surface. (author)

UO2 production process with methanol washing

International Nuclear Information System (INIS)

Sondermann, T.

1978-01-01

The invention refers to a process for the recovery of methanol used for washing the ammonium uranyl carbonate obtained during UO 2 production. The methanol contains about 50% H 2 O, about 10% (NH 4 ) 2 CO 3 , and is radioactive. According to the invention the methanol is purified at reduced pressure in a distillation unit and then led back to the washing unit. (UWI) 891 HP/UWI 892 MBE [de

Development of Innovative Accident Tolerant High Thermal Conductivity UO₂-Diamond Composite Fuel Pellets

Energy Technology Data Exchange (ETDEWEB)

Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

2016-01-01

The University of Florida (UF) evaluated a composite fuel consisting of UO₂ powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO₂-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO₂ – SiC and UO₂ – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO₂ may greatly enhanced the thermal conductivity of the UO₂ pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

Determination of the cationic self-diffusion coefficient in ThO2-5%UO2 nuclear fuel

International Nuclear Information System (INIS)

Sabioni, A.C.S.

1984-01-01

The cation self-diffusion coefficient for the ThO 2 -5%UO 2 by means of the densification model developed by Assmann and Stehle was determined. The experimental data of the fuel densification, used in the calculations, were obtained from thermal resinter tests. Our result is comparable to previously published values for U and Th diffusion in polycrystalline ThO 2 and (Th, U)O 2 . (Author) [pt

Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells

International Nuclear Information System (INIS)

Foad, Basma; Takeda, Toshikazu

2015-01-01

Highlights: • A method for calculating sensitivity coefficients has been improved. • The IR approximation was used in order to get accurate results. • Sensitivities and uncertainties are calculated using the improved method. • The method is applied for UO 2 and MOX fueled PWR cells. • The verification was performed by comparing our results with MCNP6 and TSUNAMI-1D. - Abstract: This paper discusses the improvement of a method for calculating sensitivity coefficients of neutronics parameters relative to infinite dilution cross-sections because the conventional method neglects resonance self-shielding effect. In this study, the self-shielding effect is taken into account by using the intermediate resonance approximation in order to get accurate results in both high and low energy groups. The improved method is applied to calculate sensitivity coefficients and uncertainties of eigenvalue responses for UO 2 and MOX (ThO 2 –UO 2 and PuO 2 –UO 2 ) fueled pressurized water reactor cells. The verification of the improved method was performed by comparing the sensitivities with MCNP6 and TSUNAMI-1D. For uncertainty, calculation comparisons were done with TSUNAMI-1D, and we demonstrate that the differences are caused by the use of different covariance matrices

Complexing in the system Rb2SeO4-UO2SeO4-H2O

International Nuclear Information System (INIS)

Kuchumova, N.V.; Shtokova, I.P.; Serezhkina, L.B.; Serezhkin, V.N.

1989-01-01

Method of isothermal solubility at 25 deg C is used to study interaction of rubidium and uranyl selenates in aqueous solution. Formation of congruently soluble Rb 2 UO 2 (SeO 4 ) 2 x2H 2 O and Rb 2 (UO 2 ) 2 x(SeO 4 ) 3 x6H 2 O is stated. For the last compound crystallographic characteristics (a=10.668; b=14.935(9); c=13.891(7) A; β=104.94(1); Z=4, sp.gr. P2 1 /c) are determined. Thermal decomposition of a compound results in formation of Rb 2 U 2 O 7

Behaviour of high purity UO{sub 2}/H{sub 2}O interfaces under helium beam irradiation in deaerated conditions; Comportement des interfaces UO{sub 2}/H{sub 2}O de haute purete sous faisceau d'ions He{sup 2+} en milieu desaere

Energy Technology Data Exchange (ETDEWEB)

Mendes, E

2005-11-15

A question put within the framework of the nuclear fuel storage worn in geological site is what become to them in the presence of water. The aim of a fundamental program, of PRECCI project (ECA), is to highlight the behaviour of interfaces which can be used as models for the interfaces nuclear spent fuel/water if the fuel is uranium UO{sub 2} dioxide. This doctorate is interested in the effect of the alpha activity which is the only one that exist in the spent fuel after long periods. The aim is to identify the mechanisms of alteration and of leaching of surfaces under alpha irradiation. A method is developed to irradiate UO{sub 2}/H{sub 2}O interfaces in deaerated conditions with the beam of He{sup 2+} produced by a cyclotron. The He{sup 2+} ions cross an UO{sub 2} disc and emerge in water with an energy of 5 MeV. Leachings under irradiation are carried with a large range of particles flux. The post-irradiation characterization of the surface of the discs realised by micro-Raman spectroscopy allowed to identify the alteration layer. It is made up of studtite UO{sub 2}(O{sub 2}),4H{sub 2}O, and of schoepite UO{sub 3},xH{sub 2}O. The analysis of the solutions shows that the uranium release strongly increases. The electrochemical properties of the interfaces under irradiation strongly differ from those before irradiation. This work allows to propose that the radiolytic species seen by the interface are it during the heterogeneous phase of evolution of the traces and are species of short lives. Modeling show that the radiolytic radicals species can migrate toward the interface and react with the UO{sub 2} surface. (author)

Review of the effects of burnup on the thermal conductivity of UO2

International Nuclear Information System (INIS)

Lokken, R.O.; Courtright, E.L.

1976-01-01

The general trends which relate changes in thermal conductivity of UO 2 fuel as a function of temperature and burnup can be summarized as follows: (1) At temperatures below 500 0 C, reductions in UO 2 thermal conductivity relative to the unirradiated values can be expected up to a saturation level of approximately 10 19 fissions/cc. (2) At temperatures above 500 0 C, the thermal conductivity will undergo little change at low burnups, (less than 10 19 fissions/cc) but at higher exposures some decrease can be expected which should, in turn, diminish with increasing temperature. (3) A review of the data reported by Berman on the ThO 2 --UO 2 fuel indicates that the basic behavior is the same as for UO 2 in the temperature range of major interest. The applicability of this data to LWR UO 2 fuel is somewhat questionable because of basic physical property differences, and limited data on irradiation effects, and would not seem to support concerns that the effects of burnup on thermal conductivity for LWR fuel may be of more significance than currently believed. (4) A mathematical expression of the type proposed by Daniel and Cohen seems to provide a reasonable approximation for the behavioral trends reported in the literature which relate changes in thermal conductivity to increasing burnup in certain temperature regimes. Calculations indicate that only small incremental increases in the fuel centerline temperature might be expected if burnup effects are taken into account

Effect of the UO{sub 2} powder type and mixing method on microstructure of Mn-Al doped pellet

Energy Technology Data Exchange (ETDEWEB)

Na, Yeon Soo; Lim, Kwang Young; Choi, Min young; Jung, Tae Sik; Lee, Seung Jae; Yoo, Jong Sung [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

Recently, the commercial LWRs are focused on the extending the burn-up and fuel cycle length in order to increase nuclear power plant economy as a maintenance and fuel cycle cost. Increasing the burn-up may lead to a faster and higher power variation such as a peak local linear power and normal operating transient (Load following operation). In such operating conditions, the risk of a fuel failure is considerably related to a pellet clad-interaction (PCI). So, recent development of advanced UO{sub 2} pellet for the LWRs is mainly focused on the large grain and soft pellet as they can reduce corrosive fission gas release and pellet-clad-interaction. In terms of the UO{sub 2} pellet, the prevention of PCI induced fuel failure can be achieved by enlarging the UO{sub 2} pellet grain size and enhancing the pellets deformation at an elevated temperature. In Korea, in order to increase the grain size and deformation of UO{sub 2} pellet on the high temperature, Mn-Al doped pellet with ADU (Ammonium Diuranate)-UO{sub 2} powder are developed in lab scale. But, the UO{sub 2} pellets for the commercial nuclear power plants in Korea are fabricated using the DC (Dry Conversion)-UO{sub 2} powder. So, it is necessary to understand the effect of microstructure on UO{sub 2} powder type for Mn-Al doped pellets. In this work, to investigate the effect of UO{sub 2} powder type and mixing method on the microstructure of the Mn-Al doped UO{sub 2} pellets, we fabricated the Mn-Al doped pellets using the DC-UO{sub 2} powder. The measurement of sintered density and mean grain size for fabricated pellets was performed, and then the results of test was evaluated in comparison with a Reference 2.

On the correlation between fuel structure and mechanical properties of UO2

International Nuclear Information System (INIS)

Blank, H.; Mandler, R.; Matzke, H.; Routbort, J.; Werner, P.

1983-01-01

The relation between the structure of a UO 2 fuel and its mechanical properties are discussed and illustrated for particular types of UO 2 by measurements of fracture surface energy, hardness, fracture stress and compressive deformation at 1870 and 1970 K. This gives the background for treating the question whether it is possible to find a simple experimental method for correlating the mechanical properties of UO 2 before irradiation with those after various irradiation histories. Hardness measurements might be such a method if combined with a detailed structural analysis and sufficient knowledge about the irradiation history. However, for a meaningful interpretation of the data the presently available 'classical' methods of fracture mechanics are inadequate and, furthermore, sufficient additional (not yet available) information on the relations between mechanical properties and structural details are required. (author)

Reducing the stoichiometric excess of HF in the hydrofluorination of UO2

International Nuclear Information System (INIS)

Zhao Jun; Qiu Lufu; Zhong Xing; Xu Heqing

1989-11-01

In a fluidized bed, UO 2 obtained from the decomposition-reduction of AUC (Ammonium Uranyl Carbonate) was fed to absorb HF remaining in the exhaust gas of UF 4 production process. In the case of 60% conversion of UO 2 and the reaction temperature in the region of 300 deg C, HF remaining in the exhaust gas in absorbing fluidized bed was less than 7 ∼ 8% (w/w), i.e. apparent stoichiometric excess of HF had reduced to 0% more or less. Hence, with the high hydrofluorination reactivity of UO 2 obtained from the decomposition-reduction of AUC, it is possible to reduce evidently the stoichiometric excess of HF in the hydrofluorination process by two fluidized beds in series in which solids move against the gas flow

Density, thermal expansion coefficient and viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 solutions at high temperatures

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Roth, A.

1983-01-01

Measurements have been performed of the density, of the volumetric thermal expansion coefficient and of the viscosity of liquid sodium tetraborate (borax) and of sodium metaborate both pure and with two different amounts of UO 2 dissolved in each. The viscosity measurements have been performed for the solution of sodium tetraborate with UO 2 and CeO 2 , and with CeO 2 only as well. These data are required for the design of core-catchers based on sodium borates. The density measurements have been performed with the buoyancy method in the temperature range from 825 0 C to 1300 0 C, the viscosity measurements in the temperature range 700-1250 0 C with a modified Haake viscosity balance. The balance was previously calibrated at ambient temperature with a standard calibration liquid and at high temperatures, with data for pure borax available from the literature. (orig.)

Thermal Shock Tests on UO{sub 2} Small Spheres; Essais de choc thermique sur des elements spheriques de UO{sub 2}; Ispytaniya nebol'shikh sharikov iz UO{sub 2} teplovykh udarom; Ensayo de pequenas esferas de UO{sub 2} por choque.termico

Energy Technology Data Exchange (ETDEWEB)

Perona, G.; Brutto, E.; Galbusera, U.; Palladino, G.; Sesini, R. [Centro Informazioni Studi Esperienze, Milan (Italy)

1963-11-15

If UO{sub 2} small spheres are used as fuel in a reactor in contact with the cooler, it is necessary to know the maximum value of the thermal stress, due to the work conditions in the reactor, which the small spheres are able to withstand without breaking. These conditions can be calculated if the physical properties of the material are known. Owing to the considerable number of properties involved, and in consideration of the uncertainty which always exists in each of them, it is preferable to test directly the spheres, submitting them to the same kind of stresses that they undergo in thereactor. In this work a thermal shock method for the small spheres has been studied, while conditions are indicated in which this method can reproduce stress conditions directly comparable with those existing in the reactor. As for small spheres, the difficulty consists in producing coolings with very high values of the coefficient of surface heat transfer. The experimental methods are described and the results obtained are indicated. The application of this method seems to be very interesting particularly in the field of the technological research for improving the characteristics of the UO{sub 2} small spheres by means of additives. In fact it allows the control of the total interesting effect with a single measurement. (author) [French] Si l'on veut utiliser comme combustible dans un reacteur des elements spheriques de UO{sub 2} en contact avec le refroidisseur, il faut au prealable determiner la valeur maximum de la contrainte thermique - due aux conditions regnant dans le reacteura laquelle les elements sont capables de resister sans se fissurer. Il est possible de calculer ces conditions si l'on connait les proprietes physiques du materiau utilise. En raison du nombre important des proprietes a prendre en consideration, et compte tenu de l'incertitude qui existe toujours pour chacune d'elles, il est preferable de faire des essais thermiques en soumettant directement les

In-Situ Observation of Sintering Shrinkage of UO{sub 2} Compacts Derived from Different Powder Routes

Energy Technology Data Exchange (ETDEWEB)

Rhee, Young Woo; Oh, Jang Soo; Kim, Dong Joo; Kim, Keon Sik; Kim, Jong Hun; Yang, Jae Ho; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In-situ observations on the shrinkage of green pellets with precisely controlled dimensions were carefully conducted by using TOM during H2 atmosphere sintering. The shrinkage retardation in IDR-UO{sub 2} might be attributed to the larger primary particle size of IDRUO{sub 2} than those of ADU- and AUC- UO{sub 2} powders. It would be important to understand the different sintering characteristics of UO{sub 2} powders according to the powder routes, when it comes to designing a new sintering process or choosing a sintering additive for new fuel pellet like PCI (Pellet Cladding Interaction) remedy pellet. In this paper, we have investigated the initial and intermediate sintering shrinkage of UO{sub 2} from different powder routes by in-situ observation of green samples during H2 atmosphere sintering. Effect of powder characteristics of three different UO{sub 2} powders on the initial and intermediate sintering were closely reviewed including crystal structure, powder size, specific surface area, primary crystal size, and O/U ratio.

Complexation of Cu2+, Ni2+ and UO22+ by radiolytic degradation products of bitumen

International Nuclear Information System (INIS)

Loon, L.R. Van; Kopajtic, Z.

1990-05-01

The radiolytic degradation of bitumen was studied under conditions which reflect those which will exist in the near field of a cementitious radioactive waste repository. The potential complexation capacity of the degradation products was studied and complexation experiments with Cu 2+ , Ni 2+ and UO 2 2+ were performed. In general 1:1 complexes with Cu 2+ , Ni 2+ and UO 2 2+ , with log K values of between 5.7 and 6.0 for Cu 2+ , 4.2 for Ni 2+ and 6.1 for UO 2 2+ , were produced at an ionic strength of 0.1 M. The composition of the bitumen water was analysed by GC-MS and IC. The major proportion of the bitumen degradation products in solution were monocarboxylic acids (acetic acid, formic acid, myric acid, stearic acid ...), dicarboxylic acids (oxalic acid, phthalic acid) and carbonates. The experimentally derived log K data are in good agreement with the literature and suggest that oxalate determines the speciation of Cu 2+ , Ni 2+ and UO 2 2+ in the bitumen water below pH=7. However, under the high pH conditions typical of the near field of a cementitious repository, competition with OH-ligands will be large and oxalate, therefore, will not play a significant role in the speciation of radionuclides. The main conclusion of the study is that the radiolytic degradation products of bitumen will have no influence on radionuclide speciation in a cementitious near field and, as such, need not to be considered in the appropriate safety assessment models. (author) 12 figs., 11 tabs., 31 refs

Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

International Nuclear Information System (INIS)

Holub, F.

1987-08-01

The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

Numerical characterization of micro-cell UO{sub 2}−Mo pellet for enhanced thermal performance

Energy Technology Data Exchange (ETDEWEB)

Lee, Heung Soo [School of Mechanical Engineering, Hanyang University, Seoul, 133-791 (Korea, Republic of); Kim, Dong-Joo [LWR Fuel Technology Division, Korea Atomic Energy Research Institute, Daejeon, 305-353 (Korea, Republic of); Kim, Sun Woo [School of Mechanical Engineering, Hanyang University, Seoul, 133-791 (Korea, Republic of); Yang, Jae Ho; Koo, Yang-Hyun [LWR Fuel Technology Division, Korea Atomic Energy Research Institute, Daejeon, 305-353 (Korea, Republic of); Kim, Dong Rip, E-mail: dongrip@hanyang.ac.kr [School of Mechanical Engineering, Hanyang University, Seoul, 133-791 (Korea, Republic of)

2016-08-15

Metallic micro-cell UO{sub 2} pellet with high thermal conductivity has received attention as a promising accident-tolerant fuel. Although experimental demonstrations have been successful, studies on the potency of current metallic micro-cell UO{sub 2} fuels for further enhancement of thermal performance are lacking. Here, we numerically investigated the thermal conductivities of micro-cell UO{sub 2}−Mo pellets in terms of the amount of Mo content, the unit cell size, and the aspect ratio of the micro-cells. The results showed good agreement with experimental measurements, and more importantly, indicated the importance of optimizing the unit cell geometries of the micro-cell pellets for greater increases in thermal conductivity. Consequently, the micro-cell UO{sub 2}−Mo pellets (5 vol% Mo) with modified geometries increased the thermal conductivity of the current UO{sub 2} pellets by about 2.5 times, and lowered the temperature gradient within the pellets by 62.9% under a linear heat generation rate of 200 W/cm. - Highlights: • Thermal conductivities of micro-cell UO{sub 2}−Mo pellets were numerically studied in terms of their unit cell geometries. • Numerical calculations qualitatively well agreed with experimental measurements. • Optimizing the unit cell geometries of the micro-cell pellets could greatly enhance their thermal conductivities.

Anodic dissolution of UO2 in slightly alkaline sodium perchlorate solutions

International Nuclear Information System (INIS)

Sunder, S.; Strandlund, L.K.; Shoesmith, D.W.

1996-04-01

The anodic dissolution of UO 2 has been studied in aqueous sodium perchlorate solutions at pH ∼ 9.5. Under potentiostatic conditions two distinct regions of oxidation/dissolution behaviour were observed. In the potential (E) range 0.100 V A , Q C respectively) obtained by integration of the anodic current-time plots (Q A ) and cathodic potential scans to reduce accumulated oxidized surface films (Q C ), it was shown that > ∼ 90% of the anodic oxidation current went to produce these films. For E > ∼ 0.350 V, steady-state currents were obtained and measurements of Q A and Q C showed the majority of the current went to produce soluble species. The film blocking anodic dissolution appeared to be either UO 2.27 or, more probably, UO 3 .2H 2 O located primarily at grain boundaries. It is proposed that, at the higher potentials, rapid oxidation and dissolution followed by the hydrolysis of dissolved uranyl species leads to the development of acidic conditions in the grain boundaries. At these lower pH values the UO 3 .2H 2 O is soluble and therefore does not accumulate. Alternatively, if this oxide has been formed by prior oxidation at a lower potential, the formation of protons on oxidizing at E > ∼ 0.350V causes its redissolution, allowing the current to rise to a steady-state value. On the basis of Tafel slopes, an attempt was made to demonstrate that the observed behaviour was consistent with dissolution under acidic conditions. This analysis was only partially successful. (author) 34 refs. 11 figs

Effect of titania addition on the thermal conductivity of UO2 fuel [Paper IIIB-C

International Nuclear Information System (INIS)

Sengupta, A.K.; Kumar, A.; Arora, K.B.S.; Pandey, V.D.; Nair, M.R.; Kamath, H.S.

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO 2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO 2 and UO 2 containing 0.05wt per cent and 0.1wt per cent TiO 2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO 2 on the thermal conductivity of UO 2 fuel. (author)

Effect of titania addition on the thermal conductivity of UO2 fuel (Paper IIIB-C)

Energy Technology Data Exchange (ETDEWEB)

Sengupta, A K; Kumar, A; Arora, K B.S.; Pandey, V D; Nair, M R; Kamath, H S

1986-01-01

Pellet clad interaction in nuclear reactor fuel elements can be reduced by the use of higher grain size UO2 fuel. This is achieved by the addition of dopant like titania, niobia etc. However, these dopants are considered as impurities which may affect the thermophysical and thermomechanical properties of the fuel. Thermal Conductivity which is one of the important properties controlling the inpile performance of the fuel has been measured for pure UO2 and UO2 containing 0.05wt per cent and 0.1wt per cent TiO2 in the temperature range 900K to 1900K in vacuum. Thermal conductivity was obtained from thermal diffusivity data measured by laser flash method. The paper highlights the experimental results and discusses the effect of TiO2 on the thermal conductivity of UO2 fuel. 5 figures.

A prediction of the inert gas solubilities in stoichiometric molten UO2

International Nuclear Information System (INIS)

Gunnerson, F.S.; Cronenberg, A.W.

1975-01-01

To analyze the effect of fission gas behaviour on fast reactor fuels during a hypothetical overpower transient, the solubility characteristics of the noble gases in molten UO 2 have been assessed. To accomplish this, a theoretical estimation of such solubilities is made by determining the reversible work required to introduce a hard sphere, the size of the gas atom, into the liquid solvent. Results indicate that the solubility of the noble gases in molten UO 2 is quite low, the molar fraction of gas-to-liquid being approximately 10 -6 . Such a low solubility of fission gases suggests that for preirradiated fuels, added swelling or formation may occur upon melting. In addition, such low solubility potential indicates that the fission gases do not play an appreciable role in the fragmentation of molten UO 2 upon quenching in sodium coolant. (Auth.)

Identification of secondary phases formed during unsaturated reaction of UO2 with EJ-13 water

International Nuclear Information System (INIS)

Bates, J.K.; Tani, B.S.; Veleckis, E.

1989-01-01

A set of experiments, wherein UO 2 has been contacted by dripping water, has been conducted over a period of 182.5 weeks. The experiments are being conducted to develop procedures to study spent fuel reaction under unsaturated conditions that are expected to exist over the lifetime of the proposed Yucca Mountain repository site. One half of the experiments have been terminated, while one half are ongoing. Analyses of solutions that have dripped from the reacted UO 2 have been performed for all experiments, while the reacted UO 2 surfaces have been examined for the terminated experiments. A pulse of uranium release from the UO 2 solid, combined with the formation of schoepite on the surface of the UO 2 , was observed between 39 and 96 weeks of reaction. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporated cations from the EJ-13 water and included boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are continuing to monitor whether additional changes in solution chemistry or secondary phase formation occurs. 6 refs., 2 figs., 2 tabs

X-ray photoelectron investigation of UO2(ClO4)2 interaction with diabase

International Nuclear Information System (INIS)

Teterin, Yu.A.; Nefedov, V.I.; Ivanov, K.E.; Baev, A.S.; Gajpel', G.; Rajkh, T.; Niche, Kh.

1996-01-01

X-ray diffraction study was made on interaction of soluble uranyl perchlorate with diabase, composing rocks in regions of Chelyabinsk and Chernobyl accidents and in places of radioactive waste burials. Absence of ClO 4 - anion in analyzed products of UO 2 (ClO 4 ) 2 interaction with diabase testifies to the absence of physio- or chemosorbed UO 2 (ClO 4 ) 2 layer on its surface. Formation of uranyl compounds (uranyl hydroxides) on diabase surface was confirmed, and bond lengths for these compounds were determined. Reaction of substitution of uranium ions for calcium ions proceeds more actively in the surface layers of diabase grains. 4 refs.; 3 tabs

Interatomic spacing distribution in multicomponent alloys

International Nuclear Information System (INIS)

Toda-Caraballo, I.; Wróbel, J.S.; Dudarev, S.L.; Nguyen-Manh, D.; Rivera-Díaz-del-Castillo, P.E.J.

2015-01-01

A methodology to compute the distribution of interatomic distances in highly concentrated multicomponent alloys is proposed. By using the unit cell parameter and bulk modulus of the elements involved, the method accurately describes the distortion in the lattice produced by the interaction of the different atomic species. To prove this, density functional theory calculations have been used to provide the description of the lattice in a monophasic BCC MoNbTaVW high entropy alloy and its five sub-quaternary systems at different temperatures. Short-range order is also well described by the new methodology, where the mean error in the predicted atomic coordinates in comparison with the atomistic simulations is in the order of 1–2 pm over all the compositions and temperatures considered. The new method can be applied to tailor solid solution hardening, highly dependent on the distribution of interatomic distances, and guide the design of new high entropy alloys with enhanced properties

A study on improvement of UO2 powder production process for high sintered density

International Nuclear Information System (INIS)

Park, Jin Hoh; Hwang, Sung Tae; Jun, Kwan Sik; Choi, Yoon Dong; Choi, Jong Hyun; Lee, Kyoo Il; Kim, Tae Joon; Jung, Kyung Chae; Kim, Kwang Lak; Kwon, Sang Woon; Kim, Byung Hoh; Hong, Soon Bok

1995-01-01

Various conversion processes were reviewed from the viewpoint of manufacturing cost, product quality and liquid waste. The MDD process was selected a suitable target process for the good quality of UO 2 powder and the recycling availability of nitric acid. The MDD process consists of two steps, double salt preparation [(NH 4 ) 2 UO 2 (NO 3 ) 4 ] from uranyl nitrate solution and thermal decomposition/reduction to UO 2 powder. The reaction mechanism and properties for the intermediates were analyzed to define the proposed operational conditions of the process. The conceptual process was proposed and experimental facility was designed and installed. 12 figs, 7 tabs, 7 refs. (Author)

Leaching of irradiated CANDU UO2 fuel

International Nuclear Information System (INIS)

Vandergraaf, T.T.; Johnson, L.H.; Lau, D.W.P.

1980-01-01

Irradiated fuel, leached at room temperature with distilled water and with slightly chlorinated river water, releases approx. 4% of its cesium inventory over a comparatively sort period of a few days but releases its actinides and rare earths more slowly. The matrix itself dissolves at a rate conservatively calculated to be less than approx. 2 x 10 -6 g UO 2 /cm 2 day and, with time, the leach rates of the various nuclides approach this value

BURNY-SQUID, 2-D Burnup of UO{sub 2} and Mix UO{sub 2} PuO{sub 2} Fuel in X-Y or R-Z Geometry

Energy Technology Data Exchange (ETDEWEB)

Rosa, I; Zara, G; Guidotti, R [ENEL-DCO, Via G.B. Martini, 3, 00198 Rome (Italy)

1974-08-01

1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO{sub 2} and PUO{sub 2}-UO{sub 2} fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed.

Estimated Critical Conditions for UO(Sub 2)F(Sub 2)-H(Sub 2)O Systems in Fully Water-Reflected Spherical Geometry

Energy Technology Data Exchange (ETDEWEB)

Jordan, W.C.

1992-01-01

The purpose of this report is to document reference calculations performed using the SCALE-4.0 code system to determine the critical parameters of UO{sub 2}F{sub 2}-H{sub 2}O spheres. The calculations are an extension of those documented in ORNL/CSD/TM-284. Specifically, the data for low-enriched UO{sub 2}F{sub 2}-H{sub 2}O spheres have been extended to highly enriched uranium. These calculations, together with those reported in ORNL/CSD/TM-284, provide a consistent set of critical parameters (k{sub {infinity}}, volume, mass, mass of water) for UO{sub 2}F{sub 2} and water over the full range of enrichment and moderation ratio.

Estimated critical conditions for UO{sub 2}F{sub 2}--H{sub 2}O systems in fully water-reflected spherical geometry

Energy Technology Data Exchange (ETDEWEB)

Jordan, W.C.; Turner, J.C.

1992-12-01

The purpose of this report is to document reference calculations performed using the SCALE-4.0 code system to determine the critical parameters of UO{sub 2}F{sub 2}-H{sub 2}O spheres. The calculations are an extension of those documented in ORNL/CSD/TM-284. Specifically, the data for low-enriched UO{sub 2}F{sub 2}-H{sub 2}O spheres have been extended to highly enriched uranium. These calculations, together with those reported in ORNL/CSD/TM-284, provide a consistent set of critical parameters (k{sub {infinity}}, volume, mass, mass of water) for UO{sub 2}F{sub 2} and water over the full range of enrichment and moderation ratio.

Dissolution of unirradiated UO2 fuel in synthetic groundwater. Final report (1996-1998)

International Nuclear Information System (INIS)

Ollila, K.

1999-05-01

This study was a part of the EU R and D programme 1994-1998: Nuclear Fission Safety, entitled 'Source term for performance assessment of spent fuel as a waste form'. The research carried out at VTT Chemical Technology was focused on the effects of granitic groundwater composition and redox conditions on UO 2 solubility and dissolution mechanisms. The synthetic groundwater compositions simulated deep granitic fresh and saline groundwaters, and the effects of the near-field material, bentonite, on very saline groundwater. Additionally, the Spanish granite/bentonite water was used. The redox conditions (Eh), which are obviously the most important factors that influence on UO 2 solubility under the disposal conditions of spent fuel, varied from strongly oxidising (air-saturated), anaerobic (N 2 , O 2 2 , low Eh). The objective of the air-saturated dissolution experiments was to yield the maximum solution concentrations of U, and information on the formation of secondary phases that control the concentrations, with different groundwater compositions. The static batch solubility experiments of long duration (up to 1-2 years) were performed using unirradiated UO 2 pellets and powder. Under anaerobic and reducing conditions, the solubilities were also approached from oversaturation. The results of the oxic, air-saturated dissolution experiments with UO 2 powder showed that the increase in the salinity ( -5 M, were at the level of the theoretical solubility of schoepite or another uranyl oxide hydrate, e.g. becquerelite (possibly Na-polyuranate). The higher alkalinity of the fresh (Allard) composition increased the aqueous U concentration. Only some kind of oxidised U-phase (U 3 O 8 -UO 3 ) was identified with XRD when studying possible secondary phases after the contact time of one year with all groundwater compositions. Longer contact times are needed to identify secondary phases predicted by modelling (EQ3/6). In the anoxic dissolution experiments with UO 2 pellets, the

UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die

Science.gov (United States)

Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.

2018-02-01

The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.

Effect of sintering condition on the grain growth of Cr{sub 2}O{sub 3} doped UO{sub 2} pellets

Energy Technology Data Exchange (ETDEWEB)

Oh, Jang Soo; Kim, Keon Sik; Kim, Dong Joo; Kim, Jong Hun; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this paper, Cr{sub 2}O{sub 3} doped UO{sub 2} pellets were fabricated by two-step sintering process. The grain growth of pellet is related to dwell time in a hydrogen atmosphere during sintering process. A large grain pellet can minimize fission gas release and deform easily at an elevated temperature. So, the recent development of nuclear fuel pellet materials is mainly focused on the large grain pellets. The various methods of fabrication processes for large grain UO{sub 2} pellets have been investigated extensively. Those parameters include the additives, sintering temperature, sintering time, sintering atmosphere, and so on. Cr-doped UO{sub 2} pellet is one of the promising candidates for PCI remedy. It was shown that the grain size and softness of UO{sub 2} pellets could be enhanced by doping Cr or Cr compound in UO{sub 2}. Various in-pile test results revealed that the PCI properties were enhanced considerably [4]. In the sintering process of Cr-doped UO{sub 2} pellet, it was known that tight adjusting of sintering atmosphere is most important to achieve large grain pellet. The relevant research revealed that the doped Cr{sub 2}O{sub 3} became liquid phase in optimized oxygen potential and that liquid phase promoted the grain growth. Recently, KAERI has shown that grain size of Cr-doped UO{sub 2} pellet could be more enlarged by adjusting process parameters. In this paper, we introduced a sintering process which can form a liquid phase for a large grain growth in Cr{sub 2}O{sub 3} doped UO{sub 2} pellet. The study on the effect of dwell time in H{sub 2} atmosphere during sintering process on the grain structure of sintered pellet is also a part of this work. In order to obtain large grain in pellet, it is important to increase amount of Cr that can form a liquid phase for grain growth by increasing dwell time in a hydrogen atmosphere during sintering process.

Retrieval of interatomic separations of molecules from laser-induced high-order harmonic spectra

Energy Technology Data Exchange (ETDEWEB)

Le, Van-Hoang; Nguyen, Ngoc-Ty [Department of Physics, University of Pedagogy, 280 An Duong Vuong, Ward 5, Ho Chi Minh City (Viet Nam); Jin, C; Le, Anh-Thu; Lin, C D [J. R. Macdonald Laboratory, Department of Physics, Kansas State University, Manhattan, KS 66506 (United States)

2008-04-28

We illustrate an iterative method for retrieving the internuclear separations of N{sub 2}, O{sub 2} and CO{sub 2} molecules using the high-order harmonics generated from these molecules by intense infrared laser pulses. We show that accurate results can be retrieved with a small set of harmonics and with one or few alignment angles of the molecules. For linear molecules the internuclear separations can also be retrieved from harmonics generated using isotropically distributed molecules. By extracting the transition dipole moment from the high-order harmonic spectra, we further demonstrated that it is preferable to retrieve the interatomic separation iteratively by fitting the extracted dipole moment. Our results show that time-resolved chemical imaging of molecules using infrared laser pulses with femtosecond temporal resolutions is possible.

Effect of titania addition on hot hardness of UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Sengupta, A.K. E-mail: arghya@apsara.barc.ernet.in; Basak, C.B.; Jarvis, T.; Bhagat, R.K.; Pandey, V.D.; Majumdar, S

2004-02-15

Large grain UO{sub 2} is a potential fuel for LWR's for achieving extended burn up. Large grains are obtained by addition of dopants like Nb{sub 2}O{sub 5}, TiO{sub 2}, Cr{sub 2}O{sub 3}, V{sub 2}O{sub 5} etc. However, presence of such dopants might affect the thermophysical and thermomechanical properties of the fuel. In the present investigation the effect of TiO{sub 2} addition on the hot hardness (H) of sintered UO{sub 2} fuel has been studied from ambient to 1573 K in vacuum. TiO{sub 2} content was varied from 0.01 to 0.15 w/o resulting in a grain size (G) variation of 9 to 94 {mu}m. With increase in grain size (or TiO{sub 2} content) H first decreases, attains a minima and then increases further. The increase is more prominent at lower temperature (<773 K) than that at higher temperatures. H vs. G{sup -1/2} plots indicates the same type of variation like other oxide ceramics with H minima at an intermediate grain size at low temperature. The intrinsic hardness and softening coefficient of UO{sub 2} indicate cubic dependence on TiO{sub 2} content.

Crystal-field effect in UO2

International Nuclear Information System (INIS)

Gajek, Z.; Lahalle, M.P.; Krupa, J.C.; Mulak, J.

1988-01-01

Simple ab initio model perturbation calculations of the crystal-field parameters for the U 4+ ion in UO 2 crystals are reported. The crystal-field parameters obtained, B 0 4 = -7130 cm -1 and B 0 6 = 2890 cm -1 , turn out to be much lower in value, particularly the first one, than those usually assumed for this compound. They are found, however, to agree with new spectroscopic data and recent inelastic neutron scattering measurements. (orig.)

Unirradiated UO2 in irradiated zirconium alloy sheathing

International Nuclear Information System (INIS)

MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

1979-07-01

Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

Thermal properties of UO2 from density functional theory: role of strong correlations

International Nuclear Information System (INIS)

Panigrahi, Puspamitra; Kaur Gurpreet; Valsakumar, M.C.

2011-01-01

We report a study of ground state magnetic structure of Uranium-dioxide (UO 2 ) using ab initio calculations employing PAW pseudopotentials and Dudarev's version of GGA+U formalism as implemented in VASP to take into account the strong on-site Coulomb correlation among the localized Uranium-5f electrons. By choosing the value of the Hubbard parameter U eff to be 4.0 eV, we have confirmed the experimental observation that the ground state of UO 2 is an insulator with an anti-ferromagnetic (AFM) ordering. We study systematically the ground state structural, electronic, and magnetic properties of UO 2 and focus on the structure sensitive thermal properties such as specific heat, thermal expansion and comment on the calculation of thermal conductivity. (author)

Kinetics of UO2(s) dissolution under reducing conditions: Numerical modelling

International Nuclear Information System (INIS)

Puigdomenech, I.; Casas, I.; Bruno, J.

1990-05-01

A numerical model is presented that describes the dissolution and precipitation of UO 2 (s) under reducing conditions. For aqueous solutions with pH>4, main reaction is: UO 2 (s)+2H 2 O↔U(OH) 4 (aq). The rate constant for the precipitation reaction is found to be log(k p )=-1.2±0.2 h -1 m -2 , while the value for the rate constant of the dissolution reaction is log(k d )=-9.0±0.2 mol/(1 h m 2 ). Most of the experiments reported in the literature show a fast initial dissolution of a surface film of hexavalent uranium oxide. Making the assumption that the chemical composition of the surface coating is U 3 O 7 (s), we have derived a mechanism for this process, and its rate constants have been obtained. The influence of HCO 3 - and CO 3 2- on the mechanism of dissolution and precipitation of UO 2 (s) is still unclear. From the solubility measurements reported, one may conclude that the identity of the aqueous complexes in solution is not well known. Therefore it is not possible to make a mechanistic interpretation of the kinetic data in carbonate medium. (orig.)

Application of pulsed electron beam vaporization to studies of UO2

International Nuclear Information System (INIS)

Benson, D.A.

1977-06-01

A method for determining the pressure versus internal energy coordinates of the liquid-vapor saturation curve is applied to the study of UO 2 . The experimental details and results of an initial series of tests are described. A comparison of the measurement results to models of the UO 2 equation of state illustrates the role of the heat capacity in describing the P--E characteristics of the state surface. A discussion of the available heat capacity information suggests that additional modeling and measurements of the heat capacity may be needed to give a complete temperature and energy dependent state surface description. Because of these modeling uncertainties, a method of thermodynamically describing the P(V, E) state surface entirely through the use of dynamic vapor measurements is given. Such a model satisfies transient thermomechanical analysis requirements. Next the effect of the state surface on one type of core disruptive reactor analysis is examined. And finally, the property determinations and models for UO 2 are reviewed with requirements for future work being outlined

Innovative microstructures in ThO2-UO2 system

International Nuclear Information System (INIS)

Kutty, T.R.G.; Sengupta, A.K.; Majumdar, S.; Sah, D.N.; Kamath, H.S.

2005-01-01

The basic properties that really matter to the nuclear scientists are those that have greatest influence on microstructure: crystal structure, defects concentration and phase stability. The role of microstructure and crystal defects in determining the engineering properties are always acknowledged. Microstructure of nuclear fuels controls the in-pile fuel behavior like fission gas release, plasticity, in-pile creep and swelling. Conventional nuclear ceramic fabrication process consists of a number of stages, including calcination, milling, incorporating additives, pressing, drying and densification. Since each of these steps affects the microstructure of fuel pellets they must all be understood and a more holistic approach is required when processing nuclear ceramics compared to metals and polymers. It is possible to obtain a wide range of microstructures for ThO 2 -UO 2 system if a proper fabrication route is chosen. It is possible to tailor microstructure as per our requirement so that an improved behaviour during irradiation is expected. The improvement in plasticity and fission gas release can be attained by modifying the microstructure during fabrication. This paper deals with fabrication of ThO 2 -UO 2 pellets of varying U content and its characterization with the help of optical microscopy, XRD, SEM and EPMA. The microstructures are characterized in terms grain size, pore size and its distribution and homogeneity of uranium. (author)

Thermodynamic mixing properties of the UO{sub 2}–HfO{sub 2} solid solution: Density functional theory and Monte Carlo simulations

Energy Technology Data Exchange (ETDEWEB)

Yuan, Ke, E-mail: keyuan@umich.edu [Department of Earth and Environmental Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Ewing, Rodney C. [Department of Earth and Environmental Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Becker, Udo [Department of Earth and Environmental Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

2015-03-15

HfO{sub 2} is a neutron absorber and has been mechanically mixed with UO{sub 2} in nuclear fuel in order to control the core power distribution. During nuclear fission, the temperature at the center of the fuel pellet can reach above 1300 K, where hafnium may substitute uranium and form the binary solid solution of UO{sub 2}–HfO{sub 2}. UO{sub 2} adopts the cubic fluorite structure, but HfO{sub 2} can occur in monoclinic, tetragonal, and cubic structures. The distribution of Hf and U ions in the UO{sub 2}–HfO{sub 2} binary and its atomic structure influence the thermal conductivity and melting point of the fuel. However, experimental data on the UO{sub 2}–HfO{sub 2} binary are limited. Therefore, the enthalpies of mixing of the UO{sub 2}–HfO{sub 2} binary with three different structures were calculated in this study using density functional theory and subsequent Monte Carlo simulations. The free energy of mixing was obtained from thermodynamic integration of the enthalpy of mixing over temperature. From the ΔG of mixing, a phase diagram of the binary was obtained. The calculated UO{sub 2}–HfO{sub 2} binary forms extensive solid solution across the entire compositional range, but there are a variety of possible exsolution phenomena associated with the different HfO{sub 2} polymorphs. As the structure of the HfO{sub 2} end member adopts lower symmetry and becomes less similar to cubic UO{sub 2}, the miscibility gap of the phase diagram expands, accompanied by an increase in cell volume by 7–10% as the structure transforms from cubic to monoclinic. Close to the UO{sub 2} end member, which is relevant to the nuclear fuel, the isometric uranium-rich solid solutions exsolve as the fuel cools, and there is a tendency to form the monoclinic hafnium-rich phase in the matrix of the isometric, uranium-rich solid solution phase.

Vapour pressure studies of uranium dioxide UO{sub 2} by the effusion method; Mesure de la tension de vapeur du bioxyde d'uranium UO{sub 2} par la methode d'effusion

Energy Technology Data Exchange (ETDEWEB)

Ohse, R W [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1965-07-01

A high temperature apparatus for vapour pressure measurements by Knudsen effusion method is described. Sample is heated in a tungsten cell in an electronic bombardment furnace. Several critical factors affecting the accuracy of measurements such as: - temperature distribution and measurement in the effusion cell, - CLAUSING factor and molecular flow, - compatibility between cell material and sample heated, are discussed with careful attention. Vapour pressure of UO{sub 2} has been studied between 2200 and 2800 K. Experimental points fit a curve expressed by: logP{sub mm} = 12.4264 - (3.3184/T * 10{sup 4}/T) which is in good agreement with previous results of literature. (author) [French] On decrit un appareil destine a la mesure des tensions de vapeur par la methode d'effusion de KNUDSEN. L'echantillon contenu dans une cellule en tungstene est chauffe par bombardement electronique. Apres examen critique des divers facteurs affectant l'exactitude des mesures, a savoir: - homogeneite et mesure de la temperature dans la cellule d'effusion, - facteur de 'CLAUSING' et loi de distribution en cosinus des molecules effusees, - compatibilite a chaud entre le materiau de la cellule et le materiau etudie. On a procede a la mesure de la tension de vapeur de UO{sub 2} qui est relativement bien connue. Entre 2200 et 2800 K les points experimentaux se placent sur une courbe: logP{sub mm} = 12.4264 - (3.3184/T * 10{sup 4}/T) en bon accord avec les valeurs citees dans la litterature. (auteur)

Preparation of Fluidization Feed of UO2 Pellets by Oxidation

International Nuclear Information System (INIS)

Rachmat-Pratomo; H, Didiek; Suwondo, B; Sigit

2000-01-01

The investigation of oxidation of uranium dioxide (UO 2 ) pellets to thetri uranium octoxide (U 3 O 8 ) powder had been carried. Several factor suchtemperature, time of oxidation and the concentration of air are important.The oxidation of UO 2 pellet are carried out on electric furnace atatmosphere as media. The oxidation temperature started at 300 o C, 400 o C,500 o C, and 600 o C along 1 hour. The time oxidation removed to 2 hours and3 hours. The efficiency of oxidation are the ratio of the weight of thepowder product are the uranium content, true density, and specific surfacearea. Result the optimum temperature are 500 o C along 3 hours, uraniumcontent : 84.78%, true density: 8.8293 g/cm 3 and specific surface area :0.389071 m 2 /g. (author)

Development of AUC-based process at BARC for production of free-flowing and sinterable UO2 powder

International Nuclear Information System (INIS)

Keni, V.S.; Ghosh, S.K.; Ganguly, C.; Majumdar, S.

1994-01-01

Ammonium uranium carbonate (AUC) process has been developed and industrially used in Germany for preparation of free-flowing and sinterable UO 2 powder for fabrication of UO 2 fuel pellets for light water reactors (LWR). Efforts are underway at Bhabha Atomic Research Centre (BARC) for developing AUC-based process which would yield free-flowing UO 2 powder suitable for direct pelletisation and sintering to very high density (> 96% T.D.) UO 2 fuel pellets for pressurised heavy water reactors (PHWRs) in India. The first phase of this work has been completed jointly by Chemical Engineering Division (ChED) and Radiometallurgy Division (RMD) in batches of 1.5 kg. It was possible to fabricate UO 2 pellets of density 93-95% T.D. on a reproducible basis. At ChED, process parameters have been optimised for fabrication of AUC with suitable physical properties in batches of 1.5 kg (U), starting with nuclear pure uranyl nitrate solution. At RMD calcination parameters of AUC was optimised in batches of 500 g for obtaining free-flowing UO 2 powder, suitable for direct pelletisation and sintering. The pelletisation and sintering have been carried out at Radiometallurgy Division in batches of 1-1.5 kg. The maximum achievable density of UO 2 pellets has been in the range of 95.5-96% T.D. (author). 11 refs

Topologically identical, but geometrically isomeric layers in hydrous α-, β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O and anhydrous Rb[UO2(AsO3OH)(AsO2(OH)2)

Science.gov (United States)

Yu, Na; Klepov, Vladislav V.; Villa, Eric M.; Bosbach, Dirk; Suleimanov, Evgeny V.; Depmeier, Wulf; Albrecht-Schmitt, Thomas E.; Alekseev, Evgeny V.

2014-07-01

The hydrothermal reaction of uranyl nitrate with rubidium nitrate and arsenic (III) oxide results in the formation of polymorphic α- and β-Rb[UO2(AsO3OH)(AsO2(OH)2)]·H2O (α-, β-RbUAs) and the anhydrous phase Rb[UO2(AsO3OH)(AsO2(OH)2)] (RbUAs). These phases were structurally, chemically and spectroscopically characterized. The structures of all three compounds are based upon topologically identical, but geometrically isomeric layers. The layers are linked with each other by means of the Rb cations and hydrogen bonding. Dehydration experiments demonstrate that water deintercalation from hydrous α- and β-RbUAs yields anhydrous RbUAs via topotactic reactions.

Microscopic appearance analysis of raw material used for the production of sintered UO2 by scanning electron microscope

International Nuclear Information System (INIS)

Liu feiming

1992-01-01

The paper describes the microscopic appearance of UO 2 , U 3 O 8 , ADU and AUC powders used for the production of sintered UO 2 slug of nuclear fuel component of PWR. The characteristic analysis of the microscopic appearance observed by scanning electron microscope shows that the quality and finished product rate of sintered UO 2 depend on the appearance characteristic of the active Uo 2 powder, such as grade size and its distribution, spherulitized extent, surface condition and heap model etc.. The addition of U 3 O 8 to the UO 2 powder improves significantly the quality and the finished product rate. The mechanism of this effect is discussed on the basis of the microscopic appearance characteristic for two kinds of powder

The influence of porosity on the thermal conductivity of irradiated UO2 fuel

International Nuclear Information System (INIS)

Bakker, K.; Kwast, H.; Cordfunke, E.H.P.

1994-12-01

The influence of porosity on the thermal conductivity of irradiated UO 2 fuel has been determined with the Finite Element Method (FEM). Light-microscopy photographs were made of the fuel. The pore shape and the pore distribution are entered in the FEM program from these photographs. The two dimensional (2D) thermal conductivity in the plane of the photograph is obtained from the FEM calculations. The 2D thermal conductivity, that has no physical meaning itself, is the lower limit of the three dimensional (3D) thermal conductivity. For three well defined pore shapes the relation is determined between the 2D thermal conductivity and the 3D thermal conductivity. From these computations a simple relation is obtained that transfers the 2D thermal conductivity into the 3D thermal conductivity, independent of the pore shape. The influence of porosity on the 3D thermal conductivity of irradiated UO 2 fuel and UO 2 fuel doped with Nb 2 O 5 was computed with the FEM. (orig.)

Spent-fuel special-studies progress report: probable mechanisms for oxidation and dissolution of single-crystal UO2 surfaces

International Nuclear Information System (INIS)

Wang, R.

1981-03-01

Due to the complexity of the structural, microstructural and compositional characteristics of spent fuel, basic leaching and dissolution mechanisms were studied with UO 2 matrix material, specifically with single-crystal UO 2 , to isolate individual contributory factors. The effects of oxidation and oxidation-dissolution were investigated in different oxidation conditions, such as in air, oxygenated solutions and deionized water containing H 2 O 2 . In addition, the effects of temperature on dissolution of UO 2 were studied in autoclaves at 75 and 150 0 C. Also, oxidation and dissolution measurements were investigated via electrochemical methods to determine if those techniques could be applied to the characterization of leaching and dissolution of spent fuel in a hot cell. Finally, the effects of radiation were explored since the radiolysis of water may create a localized oxidizing condition at or near the spent fuel-solution interface, even in neutral or reducing conditions as commonly found in deep geological environments. The oxidation and oxidation-dissolution mechanisms for UO 2 are proposed as follows: The UO 2 surface is first oxidized in solution to form a UO/sub 2+x/ surface layer several angstroms thick. This oxidized surface has a high dissolution rate since the UO/sub 2+x/ reacts with the dissolved O 2 , or H 2 O 2 , to form uranyl complex ions in a U(VI) state. As the uranyl ions exceed the solubility limits in solution, they become hydrolyzed to form solid deposits and suspended particles of UO 3 hydrates. The thickness and porosity of the deposited UO 3 hydrate surface-film is dependent on temperature, pH and deposition time. A long-term dissolution rate is then determined by the nature of the surface film, such as porosity, solubility and mechanical properties

Qualification of power determination and in-pile measurements in the UO{sub 2} Gd{sub 2} 0{sub 2} fuel irradiation test IFA 636

Energy Technology Data Exchange (ETDEWEB)

Tverberg, T.; Volkov, B.; Kim, J-C.

2004-04-15

IFA-S36 is irradiated with the main objective of extending the database on the performance of UO{sub 2}Gd{sub 2}O{sub 2} fuel (with 8% absorbing gadolinia isotopes) compared with commercial UO{sub 2}. The rig carries 6 rods in the lower cluster (including three Gd-doped fuel rods) and 3 rods in the upper cluster (one rod with Gd-doped fuel). The rods are instrumented with expansion thermometers (ETs), fuel and cladding elongation detectors (EFs and ECs) and pressure transducers (PFs). Repeated calorimetric power measurements, physics calculations by the HELIOS code and gamma scans of selected rods in both clusters enabled the power and burnup determination to be qualified and corrected. The data suggest that as of May 2004 the power ratings in both fuels are much alike and burnups are about 30 and 34 MW/kgUO{sub 2} in the Gd-doped and ordinary UO{sub 2} rods respectively. Analysis of in-pile measurements compared with calculations shows that neutron absorption affects fuel temperature, power and burnup radial distributions in Gd-doped fuel at BOL compared with UO{sub 2} fuel. Sensitivity analyses performed with the HELIOS and FTEMP3 codes show that fuel centreline temperature in Gd-doped fuel is influenced by radial power depression, depletion of fissile materials and absorbing Gd isotopes as well as thermal conductivity of the fuel matrix and its degradation during irradiation. Analysis of the fuel dimension changes revealed densification only in the UO{sub 2} fuel whereas fuel elongation measurements in the Gd-doped fuel rods indicated essentially constant swelling with burnup. At burnups above 5 MWd/kgUO{sub 2} the swelling rate was about 0.5-O.fi % DELTAV/V per 10 MWd/kgUO{sub 2} for both fuel types. Internal pressure measured in the Gd-doped rod at BOL showed slight fuel densification and possibly He gas absorption, whereas derived swelling rate was somewhat Iarger than values obtained from the fuel elongation measurements. Cladding elongation measurements

Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

International Nuclear Information System (INIS)

Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

1984-02-01

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

Surface modelling on heavy atom crystalline compounds: HfO2 and UO2 fluorite structures

International Nuclear Information System (INIS)

Evarestov, Robert; Bandura, Andrei; Blokhin, Eugeny

2009-01-01

The study of the bulk and surface properties of cubic (fluorite structure) HfO 2 and UO 2 was performed using the hybrid Hartree-Fock density functional theory linear combination of atomic orbitals simulations via the CRYSTAL06 computer code. The Stuttgart small-core pseudopotentials and corresponding basis sets were used for the core-valence interactions. The influence of relativistic effects on the structure and properties of the systems was studied. It was found that surface properties of Mott-Hubbard dielectric UO 2 differ from those found for other metal oxides with the closed-shell configuration of d-electrons

Obtainment of UO{sub 2} ex AUC (ammonium uranyl carbonate) from uranyl fluoride; Obtencion de UO{sub 2} ex AUC (carbonato de uranil amonio) partiendo de fluoruro de uranilo

Energy Technology Data Exchange (ETDEWEB)

Fuente, M de la; Gonzalez, A G; Gonzalez Scardaone, S; Perez de Perel, L; Marajofsky, A [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Combustibles Nucleares

1997-12-31

It is proposed the production of enriched UO{sub 2} powder starting from UF6 with the desired isotopic concentration in order to avoid the possible segregation inconveniences that take place in the mixture of enriched and natural UO{sub 2} powders. In this work is shown the feasibility of obtaining powders with direct sinterability, through the precipitation of uranyl fluoride solutions (UF{sub 6} hydrolysis). The AUC is a crystalline forerunner used in the powder production line. A simulated hydrolyzed UF{sub 6} solution was obtained by means of the dissolution of UO{sub 3} with FH acid. The precipitation operation was carried out in a discontinuous operation device, with simultaneous pumping. The precipitating media is achieved by adding simultaneously NH{sub 3} (g) and CO{sub 2} (g), using ejector nozzles during precipitation. Mother waters pH during precipitation stay between 8.5 and 9.2 and the temperature of operation is around 323 K. The AUC calcination, reduction and passivation took place during the same operation. The reduction was carried out at three different temperatures, 823 K, 933 K and 1003 K in H{sub 2} reducing atmosphere. The passivation was carried out at 343 K. The main problem of this process is free fluorine that could remain in the powder. It would inhibit its use as nuclear fuel, since the international specifications do not tolerate more than 20 mg/kg. However, the determinations carried out in all the cases, showed that it was completely eliminated during calcination. The ex AUC UO{sub 2} powders obtained from solution of F{sub 2} (UO{sub 2}) were fluoride free, showed specific areas within specification and good sinterability. Therefore it is possible to fabricate enriched powders using a humid process from F{sub 6}U, to be used without problems of segregation due to the origin of the powder mix in PWR fuels type (CAREM). (author). 3 figs.

Atomic transport properties in UO2 and mixed oxides (U,Pu)O2

International Nuclear Information System (INIS)

Matzke, H.

1987-01-01

Atomic diffusion processes in UO 2 and in the fast-breeder reactor fuel, (U,Pu)O 2 are reviewed. Emphasis is given to the slower-moving species, i.e. U and Pu. Self-diffusion, chemical diffusion, diffusion in a thermal gradient, enhancement of diffusion by radiation and fission and the operative diffusion mechanisms are discussed. The main parameter, besides the temperature, is the oxygen-to-metal ratio (O/M ratio) of the oxide. The experimental results are compared with recent calculations reported elsewhere in this volume. Also treated are effects of the possible lambda-transition at ca.2600 K in UO 2 on high-temperature kinetic processes. The present knowledge on the diffusion and mobility of fission products with emphasis on volatile and gaseous elements, and of other actinides with emphasis on their valence states are treated. Gaps in our knowledge are pointed out and the relevance of the available results for oxide fuel during reactor operation is discussed. Whereas much is known for the as-produced 'virgin' fuel, more results are urgently needed for oxides with higher burn-ups containing a few per cent fission products. Finally, technological applications of the diffusion results are treated. As an example, important savings in cost, energy and time in fuel sintering were recently achieved based on basic studies of diffusion properties of UO 2 . (author)

Creep behavior of UO2 above 20000C

International Nuclear Information System (INIS)

Slagle, O.D.

1978-01-01

A series of high temperature creep measurements were made for UO 2 in the temperature range from 2000 0 C to the melting temperature. The effects of temperature, stress and accrued strain on the creep rate have been measured. The results indicate that additional creep mechanisms are being activated at the higher temperatures

The MgSeO4-UO2SeO4-H2O system at 25 deg C

International Nuclear Information System (INIS)

Serezhkina, L.B.; Serezhkin, V.N.

1984-01-01

The method of isothermal solubility at 25 deg C has been used to study MgSeO 4 -UO 2 SeO 4 -H 2 O system. Formation of the new compound Mg 2 (UO 2 ) 3 (SeO 4 ) 5 X32H 2 O, congruently soluble in water is stated. Thermographic and X-ray diffraction investigations of the prepared magnesium selenato-uranylate and products of its dehydration are conducted

Microspheres of UO2, ThO2 and PuO2 for the high temperature reactor

International Nuclear Information System (INIS)

Brandau, T.; Brandau, E.

2010-01-01

Up to the end of the eighties of last century, the so called ''Kernels'', microspheres with a diameter of about 300 μm as sintered out of ThO 2 and UO 2 have been produced by a special vibrational dropping process. After coating and embedding in carbon the pebble fuel balls with a diameter of 60 mm included 40.000 UO 2 - or ThO 2 -microspheres in the core. Since the early nineties BRACE is developing the processing of microspheres with a broad range of materials for applications in chemical, pharmaceutical, electronic, cosmetic and food industries. One of the developing areas is the production of microspheres out of metal oxides, where different processes as sol-gel-, suspension- or mixed processes are used. (orig.)

Determination of uranium content and its impurities in the AUC and UO2 powders

International Nuclear Information System (INIS)

Boybul; Arif Nugroho

2012-01-01

The analysis of uranium (U) content and its impurities in the ammonium uranyl carbonate (AUC) and uranium dioxide (UO 2 ) produced from research reactor fuel element production installation, PT. BATAN Teknologi have been carried out. Uranium content in the powders was analyzed by potentiometric titration methods and impurity contents was analyzed by atomic absorption spectrophotometer (AAS) and by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). The purpose of this study was to determine of impurity elements in the AUC and UO 2 powder resulting from the production process if it meets the required specifications. It is reported that U content in the AUC is 48.62 wt% and that in the UO 2 is 88.08 wt%. The precision and accuracy analysis of the U content is 0,235% and 0,151%. In case of impurities in the AUC powders, it is reported that the analytical results of Zn, Ni, Cd, Co, Mn, Mg, Fe, Cu and Cr at 10.15 ppm, 1.12 ppm, not detection, not detection, not detection, 0.30 ppm, 216.07 ppm, not detection, and 31.36 ppm, respectively, while that UO 2 are 11.31 ppm, 72.14 ppm, not detection, not detection, 6.25 ppm, 8.65 ppm, 298.24 ppm, 12.75 ppm and 32, 23 ppm. The U and impurity contents in both the AUC and UO 2 fulfill the specification of nuclear fuel for RSG-GAS research reactor. (author)

Thermal conductivity of sintered UO{sub 2} under in-pile conditions; Conductibilite thermique de l'UO{sub 2} fritte dans les conditions d'utilisation en pile

Energy Technology Data Exchange (ETDEWEB)

Stora, J P; Bernardy De Sigoyer, B; Delmas, R; Deschamps, P; Lavaud, B; Ringot, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The temperature distribution in a stack of sintered UO{sub 2} cylinders has been studied both in the laboratory where the heat energy is produced by an axial heating element, and in-pile, where the heating is due solely to nuclear effects. Under a high thermal gradient the UO{sub 2} cracks both along radial planes and along pseudo-cylindrical surfaces: these latter act as thermal barriers to the heat flow, It is therefore an apparent thermal conductivity k{sub a}(T), lower than the intrinsic value k(T) of this parameter which is measured. The efficiency of these barriers decreases when the gap decreases and when the external pressure acting on the cracked stack increases: in the limiting case, for high values of the binding strain, k{sub a}(T) {approx_equal} k(T). In the domain of phonon conduction (T {<=} 1350 deg C), the expression kw.cm{sup -1}.C{sup -1}=1/(11+0.024*T) accounts for the real thermal conductivity. Above 1350 deg C the thermal conductivity increases. Two in-pile measurements up to 1250 deg C carried out using cartridges fitted with thermocouples confirm, within the limits of experimental error, the above expression and the qualitative effects of the binding strains. Similar tests have been carried out-of-pile and in-pile on the real shape of the EL-4 fuel 'pencils'. Out-of-pile, the influence of the initial free gap, of the nature of the gas filing the 'pencil' and of the external pressure have been studied; the results are compatible with the above interpretation; It appears that an external pressure of 60 kg/cm{sup 2} is insufficient to restore completely the thermal conductivity of the fuel. (authors) [French] La distribution de temperature dans un empilement de cylindres d'UO{sub 2} fritte est etudiee a la fois au laboratoire, ou l'energie calorifique est produite par un element chauffant axial, et en pile, ou l'echauffement est uniquement nucleaire. Sous gradient thermique eleve, l'UO{sub 2} se fracture a la fois suivant des plans radiaux et

Effects of UO2 fuel microstructure and density on fuel in-reactor performance

International Nuclear Information System (INIS)

Hansson, L.

1988-02-01

The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO{sub 2}