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Sample records for fresh uo2 fuel

  1. The effect of fuel chemistry on UO2 dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda; Hanson, Brady; Miller, William

    2016-08-01

    The dissolution rate of both unirradiated UO2 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 infiltration into 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 have on the dissolution rate of unirradiated UO2 under repository conditions and compare them to the rates predicted by current dissolution models. Both unirradiated UO2 and UO2 doped with varying concentrations of Gd2O3, to simulate used fuel composition after long time periods where radiolysis has minor contributions to dissolution, were examined. In general, a rise in temperature increased the dissolution rate of UO2 and had a larger effect on pure UO2 than on those doped with Gd2O3. Oxygen dependence was observed in the UO2 samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO2 matrix showed 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 O2 concentrations in the leachate where the rates would typically be elevated.

  2. The effect of fuel chemistry on UO2 dissolution

    Science.gov (United States)

    Casella, Amanda; Hanson, Brady; Miller, William

    2016-08-01

    The dissolution rate of both unirradiated UO2 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 UO2 under oxidizing repository conditions and compare them to the rates predicted by current dissolution models. Both unirradiated UO2 and UO2 doped with varying concentrations of Gd2O3, 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 UO2 and had a larger effect on pure UO2 than on those doped with Gd2O3. Oxygen dependence was observed in the UO2 samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO2 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 O2 concentrations in the leachate where the rates would typically be elevated.

  3. ADAGIO technique: From UO 2 fuels to MOX fuels

    Science.gov (United States)

    Pontillon, Y.; Desgranges, L.; Poulesquen, A.

    2009-03-01

    The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning 'Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas'). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of 85Kr release kinetics which was developed for UO 2 was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO 2 matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.

  4. Synthesis and sintering of UN-UO2 fuel composites

    Science.gov (United States)

    Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A.; Tyburska-Püschel, Beata; Meyer, Mitch; Xu, Peng; Lahoda, Edward J.; Butt, Darryl P.

    2015-11-01

    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, UO2, 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 UO2 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% UO2 in a planetary ball mill. UN and UN - UO2 composite pellets were sintered in Ar - (0-1 at%) N2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO2 composite pellets were also sintered in Ar - 100 ppm N2 to assess the effects of temperature (1700-2000 °C) on the final grain morphology and phase concentration.

  5. Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

    Science.gov (United States)

    Phillips, W. M.

    1973-01-01

    The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

  6. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  7. Low-enriched fuel particle performance review. [UO2

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance.

  8. Determination of Gd concentration profile in UO2-Gd2O3 fuel pellets

    CERN Document Server

    Tobia, D; Milano, J; Butera, A; Kempf, R; Bianchi, L; Kaufmann, F

    2014-01-01

    A transversal mapping of the Gd concentration was measured in UO2-Gd2O3 nuclear fuel pellets by electron paramagnetic resonance spectroscopy (EPR). The quantification was made from the comparison with a Gd2O3 reference sample. The nominal concentration in the pellets is UO2: 7.5 % Gd2O3. A concentration gradient was found, which indicates that the Gd2O3 amount diminishes towards the edges of the pellets. The concentration varies from (9.3 +/- 0.5)% in the center to (5.8 +/- 0.3)% in one of the edges. The method was found to be particularly suitable for the precise mapping of the distribution of Gd3+ ions in the UO2 matrix.

  9. Preparation, characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

    Science.gov (United States)

    Stennett, Martin C.; Corkhill, Claire L.; Marshall, Luke A.; Hyatt, Neil C.

    2013-01-01

    The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO2 fuel analogue, CeO2, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO2 powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO2 fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO3.

  10. TiO2 doped UO2 fuels sintered by spark plasma sintering

    Science.gov (United States)

    Yao, Tiankai; Scott, Spencer M.; Xin, Guoqing; Lian, Jie

    2016-02-01

    UO2 fuels doped with oxide additives Cr2O3 and TiO2 display larger grain size, improving fission product retention capability and thus accident tolerance. Spark plasma sintering (SPS) was applied to consolidate TiO2-doped UO2 fuel pellets with 0.5 wt % dopant concentration, above its solubility, in order to induce eutectic phase formation and promote sintering kinetics. The grain size can reach 80 μm by sintering at 1700 °C for 20 min, and liquid U-Ti-O eutectic phase occurs at the triple junction of grain boundaries and significantly improves grain growth during sintering. The oxide additive also impedes the reduction of the initial hyperstoichiometric fuel powders to more stoichiometric fuel pellets upon SPS process. Thermal-mechanical properties of the sintered doped fuel pellets including thermal conductivity and hardness are measured and compared with undoped fuel pellets. The enlarged grain size (80 μm) and densification within short sintering duration highlight the immense possibility of SPS in fabricating large grained UO2 fuel pellets to improve fuel performance.

  11. Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO2

    Science.gov (United States)

    Martin, P. M.; Vathonne, E.; Carlot, G.; Delorme, R.; Sabathier, C.; Freyss, M.; Garcia, P.; Bertolus, M.; Glatzel, P.; Proux, O.

    2015-11-01

    X-ray Absorption Spectroscopy (XAS) was used to study the behavior of krypton as a function of its concentration in UO2 samples implanted with Kr ions. For a 0.5 at.% krypton local concentration, by combining XAS results and DFT + U calculations, we show that without any thermal treatment Kr atoms are mainly incorporated in the UO2 lattice as single atoms inside a neutral bound Schottky defect with O vacancies aligned along the (100) direction (BSD1). A thermal treatment at 1273 K induces the precipitation of dense Kr nano-aggregates, most probably solid at room temperature. In addition, 26 ± 2% of the Kr atoms remain inside BSD1 showing that Kr-BSD1 complex is stable up to this temperature. Consequently, the (in-)solubility of krypton in UO2 has to be re-evaluated. For high Kr concentration (8 at.%), XAS signals show that Kr atoms have precipitated in nanometer-sized aggregates with internal densities ranging between 4.15(7) g cm-3 and 3.98(5) g cm-3 even after annealing at 873 K. By neglecting the effect due to the UO2 matrix, the corresponding krypton pressures at 300 K were equal to 2.6(3) GPa and 2.0(2) GPa, respectively. After annealing at 1673 K, regardless of the initial Kr concentration, a bi-modal distribution is observed with solid nano-aggregates even at room temperature and larger cavities only partially filled with Kr. These results are very close to those observed in UO2 fuel irradiated in reactor. In this study we show that a rare gas can be used as a probe to investigate the defect creation and their stability in UO2.

  12. Water corrosion of spent nuclear fuel: radiolysis driven dissolution at the UO2/water interface.

    Science.gov (United States)

    Springell, Ross; Rennie, Sophie; Costelle, Leila; Darnbrough, James; Stitt, Camilla; Cocklin, Elizabeth; Lucas, Chris; Burrows, Robert; Sims, Howard; Wermeille, Didier; Rawle, Jonathan; Nicklin, Chris; Nuttall, William; Scott, Thomas; Lander, Gerard

    2015-01-01

    X-ray diffraction has been used to probe the radiolytic corrosion of uranium dioxide. Single crystal thin films of UO(2) were exposed to an intense X-ray beam at a synchrotron source in the presence of water, in order to simultaneously provide radiation fields required to split the water into highly oxidising radiolytic products, and to probe the crystal structure and composition of the UO(2) layer, and the morphology of the UO(2)/water interface. By modeling the electron density, surface roughness and layer thickness, we have been able to reproduce the observed reflectivity and diffraction profiles and detect changes in oxide composition and rate of dissolution at the Ångström level, over a timescale of several minutes. A finite element calculation of the highly oxidising hydrogen peroxide product suggests that a more complex surface interaction than simple reaction with H(2)O(2) is responsible for an enhancement in the corrosion rate directly at the interface of water and UO(2), and this may impact on models of long-term storage of spent nuclear fuel.

  13. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  14. DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Kessinger, G.; Thompson, M.

    2009-08-07

    The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

  15. An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

    Energy Technology Data Exchange (ETDEWEB)

    Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

    2007-06-01

    Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted ‘traditional’ fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of

  16. A model for evolution of oxygen potential and stoichiometry deviation in irradiated UO 2 fuel

    Science.gov (United States)

    Ozrin, V. D.

    2011-12-01

    A model for radial redistribution of oxygen in irradiated UO 2 fuel under conditions of temperature and fission rate gradients has been developed. The oxygen transport in irradiated fuel is considered as a two-scale problem. On the local scale defined by the grain size, irradiated fuel is considered as a multi-phase system including solid solution of fission products in UO 2 matrix, solid precipitates (metal phase, grey phase of complex ternary compounds, the phase of condensed CsI) formed at the gas/solid interface and the gas phase in the intergranular bubbles. Intraganular transport of fission products is described by a set of diffusion equations which are supplemented by the condition of partial thermochemical equilibrium in the subsystem "precipitates & gas phase". The boundary conditions are formulated basing on thermochemical equilibrium on the interface of subsystems "solid solution" and "precipitates & gas phase". Calculation of the partial thermochemical equilibrium yields local values of the oxygen chemical potential and the deviation from fuel stoichiometry. On the global scale defined by the fuel pellet size, spatial variations of the oxygen potential caused by the temperature gradients or the presence of sources/sinks at the pellet boundary determine thermal diffusion fluxes resulting in redistribution of oxygen. The whole set of equations describing local equilibration and the transport in the local and global scales is solved in a self-consistent manner. The model results for radial distribution of oxygen potential of UO 2 calculated for typical reactor operating conditions and the fuel burnup up ˜100 MW d/kg HM are in satisfactory agreement with experimental data.

  17. Compatibility study between U-UO2 cermet fuel and T91 cladding

    Science.gov (United States)

    Mishra, Sudhir; Kaity, Santu; Khan, K. B.; Sengupta, Pranesh; Dey, G. K.

    2016-12-01

    Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO2 cermet fuel and T91 cladding material. These diffusion couples were annealed at 923-1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope (SEM), X-ray energy dispersive spectroscope (EDS) and electron probe microanalyser (EPMA). It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U6(Fe,Cr) and U(Fe,Cr)2 intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.

  18. Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Steven D. Herrmann; Michael F. Simpson

    2006-08-01

    A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

  19. Microstructural Analysis of Irradiated UO2 Dispersed Fuel during High Temperature Failure%UO2弥散型燃料辐照后高温失效时显微分析

    Institute of Scientific and Technical Information of China (English)

    伍晓勇; 王斐; 温榜

    2012-01-01

    The blister annealing has been performed on UO2 dispersed fuel plate under various temperature conditions. The microstructure of UO2 particles changing with the rise of temperature is studied by using the microscope. Fission gas releases at high temperature and the bubbles burst the coat to generate cracks and result in the local fracture and fall off of UO2 particles. A lot of holes occur in the reaction layer outside of UO2 coat because of the diffusion reaction intensified by the high temperature.%对辐照后燃料板进行退火试验,用显微镜观察了UO2颗粒的微观形貌随着温度的升高的变化情况.在高温下裂变气体膨胀,在UO2颗粒内气孔贯通形成裂纹穿破涂层,引起UO2颗粒局部破碎脱落,UO2涂层外的反应层由于高温加剧的扩散反应还形成了明显的孔洞.

  20. Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

    Science.gov (United States)

    Piro, M. H. A.; Banfield, J.; Clarno, K. T.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

    2013-10-01

    Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102 GW d t(U)-1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

  1. Artificial neural network modeling for fission gas release in LWR UO 2 fuel under RIA conditions

    Science.gov (United States)

    Koo, Yang-Hyun; Oh, Jae-Yong; Lee, Byung-Ho; Tahk, Young-Wook; Song, Kun-Woo

    2010-10-01

    A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO 2 fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.

  2. Role of microstructure and surface defects on the dissolution kinetics of CeO2, a UO2 fuel analogue.

    OpenAIRE

    Corkhill, C.L; Bailey, D. J.; Tocino, F.Y.; Stennett, M.C.; Miller, J. A.; Provis, J.P.; Travis, K.P.; Hyatt, N.C.

    2016-01-01

    The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesised CeO2 analogue for UO2 fuel. Dissolution was performed on: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vac...

  3. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Science.gov (United States)

    Foad, Basma; Takeda, Toshikazu

    2015-12-01

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO2 and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  4. Radiation induced dissolution of UO 2 based nuclear fuel - A critical review of predictive modelling approaches

    Science.gov (United States)

    Eriksen, Trygve E.; Shoesmith, David W.; Jonsson, Mats

    2012-01-01

    Radiation induced dissolution of uranium dioxide (UO 2) nuclear fuel and the consequent release of radionuclides to intruding groundwater are key-processes in the safety analysis of future deep geological repositories for spent nuclear fuel. For several decades, these processes have been studied experimentally using both spent fuel and various types of simulated spent fuels. The latter have been employed since it is difficult to draw mechanistic conclusions from real spent nuclear fuel experiments. Several predictive modelling approaches have been developed over the last two decades. These models are largely based on experimental observations. In this work we have performed a critical review of the modelling approaches developed based on the large body of chemical and electrochemical experimental data. The main conclusions are: (1) the use of measured interfacial rate constants give results in generally good agreement with experimental results compared to simulations where homogeneous rate constants are used; (2) the use of spatial dose rate distributions is particularly important when simulating the behaviour over short time periods; and (3) the steady-state approach (the rate of oxidant consumption is equal to the rate of oxidant production) provides a simple but fairly accurate alternative, but errors in the reaction mechanism and in the kinetic parameters used may not be revealed by simple benchmarking. It is essential to use experimentally determined rate constants and verified reaction mechanisms, irrespective of whether the approach is chemical or electrochemical.

  5. Studies of the UO 2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

    Science.gov (United States)

    Shiozawa, S.; Ichikawa, M.; Fujishiro, T.

    1988-06-01

    Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO 2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

  6. Resolving the H 2 effect on radiation induced dissolution of UO 2-based spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Jonsson, Mats

    2010-01-01

    In recent years, the impact of H2 on α-radiation induced dissolution of UO2-based spent nuclear fuel has been studied and debated extensively. Experimental results on the effect of H2 on the concentration of H2O2 during α-radiolysis have been shown to disagree with numerical simulations. For this reason, the reaction scheme used in simulations of aqueous radiation chemistry has sometimes been questioned. In this work, we have studied the impact of H2 on the H2O2 concentration in α-irradiated aqueous solution using numerical simulations. The effects of H2 pressure, α-dose rate and HCO3- concentration were investigated by performing systematic variations in these parameters. The simulations show that the discrepancy between the previously published experimental result and numerical simulations is due to the use of a homogeneous dose rate (the energy is assumed to be equally distributed in the whole volume). Taking the actual dose rate of the α-irradiated volume into account, the simulation is in perfect agreement with the experimental results. This shows that the H2 effect is strongly α-dose rate dependent, and proves the reliability of the reaction scheme used in the simulations. The simulations also show that H2 influences the H2O2 concentration under α-radiolysis. The magnitude of the effect depends on the dose rate and the H2 pressure as well as on the concentration of HCO 3-. The impact of the radiolytic H2 effect on the rate of α-radiation induced dissolution of spent nuclear fuel is discussed along with other (α- and γ-) radiation induced processes capable of reducing the concentration of uranium in solution. The radiolytic H2 effect is quantitatively compared to the previously presented noble metal catalyzed H2 effect. This comparison shows that the noble metal catalyzed H2 effect is far more efficient than the radiolytic H2 effect. Reduction of U(VI) in solution due to low dose rate γ-radiolysis in the presence of H2 is proposed to be the cause of

  7. What Is the Actual Local Crystalline Structure of Uranium Dioxide, UO2? A New Perspective for the Most Used Nuclear Fuel.

    Science.gov (United States)

    Desgranges, L; Ma, Y; Garcia, Ph; Baldinozzi, G; Siméone, D; Fischer, H E

    2017-01-03

    Up to now, uranium dioxide, the most used nuclear fuel, was said to have a Fm3̅m crystalline structure from 30 to 3000 K, and its behavior was modeled under this assumption. However, recently X-ray diffraction experiments provided atomic pair-distribution functions of UO2, in which UO distance was shorter than the expected value for the Fm3̅m space group. Here we show neutron diffraction results that confirm this shorter UO bond, and we also modeled the corresponding pair-distribution function showing that UO2 has a local Pa3̅ symmetry. The existence of a local lower symmetry in UO2 could explain some unexpected properties of UO2 that would justify UO2 modeling to be reassessed. It also deserves more study from an academic point of view because of its good thermoelectric properties that may originate from its particular crystalline structure.

  8. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  9. Hydrogen suppresses UO 2 corrosion

    Science.gov (United States)

    Carbol, Paul; Fors, Patrik; Gouder, Thomas; Spahiu, Kastriot

    2009-08-01

    Release of long-lived radionuclides such as plutonium and caesium from spent nuclear fuel in deep geological repositories will depend mainly on the dissolution rate of the UO 2 fuel matrix. This dissolution rate will, in turn, depend on the redox conditions at the fuel surface. Under oxidative conditions UO 2 will be oxidised to the 1000 times more soluble UO 2.67. This may occur in a repository as the reducing deep groundwater becomes locally oxidative at the fuel surface under the effect of α-radiolysis, the process by which α-particles emitted from the fuel split water molecules. On the other hand, the groundwater corrodes canister iron generating large amounts of hydrogen. The role of molecular hydrogen as reductant in a deep bedrock repository is questioned. Here we show evidence of a surface-catalysed reaction, taking place in the H 2-UO 2-H 2O system where molecular hydrogen is able to reduce oxidants originating from α-radiolysis. In our experiment the UO 2 surface remained stoichiometric proving that the expected oxidation of UO 2.00 to UO 2.67 due to radiolytic oxidants was absent. As a consequence, the dissolution of UO 2 stopped when equilibrium was reached between the solid phase and U 4+ species in the aqueous phase. The steady-state concentration of uranium in solution was determined to be 9 × 10 -12 M, about 30 times lower than previously reported for reducing conditions. Our findings show that fuel dissolution is suppressed by H 2. Consequently, radiotoxic nuclides in spent nuclear fuel will remain immobilised in the UO 2 matrix. A mechanism for the surface-catalysed reaction between molecular hydrogen and radiolytic oxidants is proposed.

  10. Role of Microstructure and Surface Defects on the Dissolution Kinetics of CeO2, a UO2 Fuel Analogue.

    Science.gov (United States)

    Corkhill, Claire L; Bailey, Daniel J; Tocino, Florent Y; Stennett, Martin C; Miller, James A; Provis, John L; Travis, Karl P; Hyatt, Neil C

    2016-04-27

    The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesized CeO2 analogue for UO2 fuel. Dissolution was performed on the following: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vacancy defects and on crushed CeO2 particles of different size fractions. BET surface area measurements were used as an indicator of reactive surface site concentration. Cerium stoichiometry, determined using X-ray Photoelectron Spectroscopy (XPS) and supported by X-ray Diffraction (XRD) analysis, was used to determine oxygen vacancy concentration. Upon dissolution in nitric acid medium at 90 °C, a quantifiable relationship was established between the concentration of high energy surface sites and CeO2 dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-x gave rise to rates that were an order of magnitude greater than for CeO2 with fewer oxygen vacancies. While enhanced solubility of Ce(3+) influenced the dissolution, it was shown that replacement of vacancy sites by oxygen significantly affected the dissolution mechanism due to changes in the lattice volume and strain upon dissolution and concurrent grain boundary decohesion. These results highlight the significant influence of defect sites and grain boundaries on the dissolution kinetics of UO2 fuel analogues and reduce uncertainty in the long term performance of spent fuel in geological disposal.

  11. Development of an Innovative High-Thermal Conductivity UO2 Ceramic Composites Fuel Pellets with Carbon Nano-Tubes Using Spark Plasma Sintering

    Energy Technology Data Exchange (ETDEWEB)

    Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States); Wu, Kuang-Hsi [Florida International Univ. (FIU), Miami, FL (United States); Tulenko, James [Univ. of Florida, Gainesville, FL (United States)

    2014-03-10

    Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. Despite its numerous advantages such as high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation, it suffers from low thermal conductivity that can result in large temperature gradients within the UO2 fuel pellet, causing it to crack and release fission gases. Thermal swelling of the pellets also limits the lifetime of UO2 fuel in the reactor. To mitigate these problems, we propose to develop novel UO2 fuel with uniformly distributed carbon nanotubes (CNTs) that can provide high-conductivity thermal pathways and can eliminate fuel cracking and fission gas release due to high temperatures. CNTs have been investigated extensively for the past decade to explore their unique physical properties and many potential applications. CNTs have high thermal conductivity (6600 W/mK for an individual single- walled CNT and >3000 W/mK for an individual multi-walled CNT) and high temperature stability up to 2800°C in vacuum and about 750°C in air. These properties make them attractive candidates in preparing nano-composites with new functional properties. The objective of the proposed research is to develop high thermal conductivity of UO2–CNT composites without affecting the neutronic property of UO2 significantly. The concept of this goal is to utilize a rapid sintering method (5–15 min) called spark plasma sintering (SPS) in which a mixture of CNTs and UO2 powder are used to make composites with different volume fractions of CNTs. Incorporation of these nanoscale materials plays a fundamentally critical role in controlling the performance and stability of UO2 fuel. We will use a novel in situ growth process to grow CNTs on UO2 particles for rapid sintering and develop UO2-CNT composites. This method is expected to provide a uniform distribution of CNTs at various volume fractions so that a high

  12. Corrigendum to "Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel"

    Science.gov (United States)

    Piro, M. H. A.; Banfield, J.; Clarno, K.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

    2016-09-01

    Figs. 7-9 in "Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel" [1] have a consistent error corresponding to the relative proportions of iodine. Reported concentrations of iodine in the original manuscript are approximately ten times higher than expected, and are comparable in atomic proportions to cesium. One would expect that the amount of cesium would be about one order of magnitude greater than iodine based on the difference in fission yields of 235U and 239Pu. A practical consequence of this error would affect the predicted quantity and chemical composition of iodine on the fuel surface, which is related to iodine-induced stress corrosion cracking [2].

  13. Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

    Science.gov (United States)

    Braun, James; Guéneau, Christine; Alpettaz, Thierry; Sauder, Cédric; Brackx, Emmanuelle; Domenger, Renaud; Gossé, Stéphane; Balbaud-Célérier, Fanny

    2017-04-01

    Silicon carbide-silicon carbide (SiC/SiC) composites are considered to replace the current zirconium-based cladding materials thanks to their good behavior under irradiation and their resistance under oxidative environments at high temperature. In the present work, a thermodynamic analysis of the UO2±x/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO2±x) is investigated in the 1500-1970 K temperature range. The reaction leads to the formation of mainly uranium silicides and carbides phases along with CO and SiO gas release. Knudsen Cell Mass Spectrometry is used to measure the gas release occurring during the reaction between UO2+x and SiC powders as function of time and temperature. These experimental conditions are representative of an open system. Diffusion couple experiments with pellets are also performed to study the reaction kinetics in closed system conditions. In both cases, a limited chemical reaction is observed below 1700 K, whereas the reaction is enhanced at higher temperature due to the decomposition of SiC leading to Si vaporization. The temperature of formation of the liquid phase is found to lie between 1850 < T < 1950 K.

  14. Dissolution of Irradiated Commercial UO2 Fuels in Ammonium Carbonate and Hydrogen Peroxide

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z.; Johnsen, Amanda M.; McNamara, Bruce K.; Hanson, Brady D.; Chenault, Jeffrey W.; Carson, Katharine J.; Peper, Shane M.

    2011-01-18

    We propose and test a disposition path for irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide media. We demonstrate on a 13 g scale that >98% of the irradiated fuel dissolves. Subsequent expulsion of carbonate from the dissolver solution precipitates >95% of the plutonium, americium, curium, and substantial amounts of fission products, effectively partitioning the fuel at the dissolution step. Uranium can be easily recovered from solution by any of several means, such as ion exchange, solvent extraction, or direct precipitation. Ammonium carbonate can be evaporated from solution and recovered for re-use, leaving an extremely compact volume of fission products, transactinides, and uranium. Stack emissions are predicted to be less toxic, less radioactive, chemically simpler, and simpler to treat than those from the conventional PUREX process.

  15. Experimental evidence of oxygen thermo-migration in PWR UO2 fuels during power ramps using in-situ oxido-reduction indicators

    Science.gov (United States)

    Riglet-Martial, Ch.; Sercombe, J.; Lamontagne, J.; Noirot, J.; Roure, I.; Blay, T.; Desgranges, L.

    2016-11-01

    The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U4O9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

  16. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Energy Technology Data Exchange (ETDEWEB)

    Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  17. Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO2 +/- x: Implications for nuclear fuel performance modeling

    Energy Technology Data Exchange (ETDEWEB)

    Giovanni Pastore; Michael R. Tonks; Derek R. Gaston; Richard L. Williamson; David Andrs; Richard Martineau

    2014-03-01

    Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2x nonstoichiometrywere used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated. 2014 Elsevier B.V. All rights

  18. Effect of Al(OH)3 on the sintering of UO2-Gd2O3 fuel pellets with addition of U3O8 from recycle

    Science.gov (United States)

    dos Santos, Lauro Roberto; Durazzo, Michelangelo; Urano de Carvalho, Elita Fontenele; Riella, Humberto Gracher

    2017-09-01

    The incorporation of gadolinium as burnable poison directly into nuclear fuel is important for reactivity compensation, which enables longer fuel cycles. The function of the burnable poison fuel is to control the neutron population in the reactor core during its startup and the beginning of the fuel burning cycle to extend the use of the fuel. The implementation of UO2-Gd2O3 poisoned fuel in Brazil has been proposed according to the future requirements established for the Angra-2 nuclear power plant. The UO2 powder used is produced from the Ammonium Uranyl Carbonate (AUC). The incorporation of Gd2O3 powder directly into the AUC-derived UO2 powder by dry mechanical blending is the most attractive process, because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. The cause of the low densities is the bad sintering behavior of the UO2-Gd2O3 mixed fuel, which shows a blockage in the sintering process that hinders the densification. This effect has been overcome by microdoping of the fuel with small quantities of aluminum. The process for manufacturing the fuel inevitably generates uranium-rich scraps from various sources. This residue is reincorporated into the production process in the form of U3O8 powder additions. The addition of U3O8 also hinders densification in sintering. This study was carried out to investigate the influence of both aluminum and U3O8 additives on the density of fuel pellets after sintering. As the effects of these additives are counterposed, this work studied the combined effect thereof, seeking to find an applicable composition for the production process. The experimental results demonstrated the effectiveness of aluminum, in the form of Al(OH)3, as an additive to promote increase in the densification of the (U,Gd)O2 pellets during sintering, even with high additions of U3O8 recycled from the manufacturing process.

  19. An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

    Science.gov (United States)

    Cartas, Andrew R.

    The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

  20. The Manufacture of W-UO2 Fuel Elements for NTP Using the Hot Isostatic Pressing Consolidation Process

    Science.gov (United States)

    Broadway, Jeramie; Hickman, Robert; Mireles, Omar

    2012-01-01

    NTP is attractive for space exploration because: (1) Higher Isp than traditional chemical rockets (2)Shorter trip times (3) Reduced propellant mass (4) Increased payload. Lack of qualified fuel material is a key risk (cost, schedule, and performance). Development of stable fuel form is a critical path, long lead activity. Goals of this project are: Mature CERMET and Graphite based fuel materials and Develop and demonstrate critical technologies and capabilities.

  1. Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

    OpenAIRE

    Serrano Purroy, D.; Casas Pons, Ignasi; Gonzalez Robles, E.; Glatz, Jean Paul; Wegen, D.H.; Clarens Blanco, Frederic; Giménez Izquierdo, Francisco Javier; Pablo Ribas, Joan de; Martínez Esparza, A.

    2013-01-01

    The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used. For uranium and actinides, the results showed that U, Np, Am and Cm ga...

  2. Effect of additives in sintering UO2-7wt%Gd{sub 2}O{sub 3} fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Santos, L.R., E-mail: lauro@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil); Riella, H.G., E-mail: riella@enq.ufsc.br [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Departamento de Engenharia Quimica e Alimentos

    2009-07-01

    Gadolinium has been used as burnable poison for reactivity control in modern PWRs. The incorporation of Gd{sub 2}O{sub 3} powder directly into the UO{sub 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{sub 2}- Gd{sub 2}O{sub 3} generates scraps that should be reused. The main scraps are green and sintered pellets, which must be calcined to U{sub 3}O{sub 8} to return to the fabrication process. Also, the incorporation of Gd{sub 2}O{sub 3} in UO{sub 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{sub 2}-Gd{sub 2}O{sub 3} fuel. Aluminum hydroxide Al(OH){sub 3} was also incorporated to the fuel as a sintering aid. The results shown that the use of 2000 ppm of Al(OH){sub 3} as additive allow to fabricate good pellets with up to 10 wt% of recycled scraps. (author)

  3. Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

    Science.gov (United States)

    Serrano-Purroy, D.; Casas, I.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; Clarens, F.; Giménez, J.; de Pablo, J.; Martínez-Esparza, A.

    2013-03-01

    The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used.For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample.Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample.Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

  4. Lattice parameter changes associated with the rim-structure formation in high burn-up UO 2 fuels by micro X-ray diffraction

    Science.gov (United States)

    Spino, J.; Papaioannou, D.

    2000-10-01

    Radial variations of the lattice parameter and peak width of two high burn-up UO 2-fuels (67 and 80 GWd/tM) were measured by a specially developed micro-X-ray diffraction technique, allowing spectra acquisition with 30 μm spatial resolution. The results showed a significant but constant peak broadening, and a lattice parameter that increased towards the pellet edge and decreased again within the rim-zone. This lattice contraction coincided with other property changes in the rim region, i.e., porosity increase, hardness decrease and Xe depletion. In terms of local burn-ups, the lattice contraction followed the rate of the matrix Xe depletion measured by EMPA, exceeding greatly the contraction rate due to dissolved fission products. The observed behaviour can be equally explained by a saturation of single interstitials with subsequent recombination with excess vacancies, as by the saturation and enlargement of dislocation loops. The concentration and sizes of defects involved and their possible relation to the rim structure formation are discussed.

  5. Can redox sensitive radionuclides be immobilized on the surface of spent nuclear fuel? - A model study on the reduction of Se(IV) aq on Pd-doped UO 2 under H 2 atmosphere

    Science.gov (United States)

    Puranen, Anders; Trummer, Martin; Jonsson, Mats

    2009-08-01

    Spent nuclear fuel contains noble metal particles composed of fission products (Pd, Mo, Ru, Tc, Rh and Te, often referred to as ɛ-particles). Studies have shown that these particles play a major role in catalyzing oxidative dissolution as well as H 2 reduction of the oxidized UO 2 fuel matrix, depending on the conditions. Thus it is possible that these particles also could have a major impact on the state of other redox sensitive radionuclides (such as the long lived fission product 79Se) present in spent nuclear fuel. In this study, Pd-doped UO 2 pellets are used to simulate noble metal particles inclusions in spent nuclear fuel and the effect on dissolved selenium in the form of selenite (250 μM selenite) in simulated ground water solution (10 mM NaCl, 10 mM NaHCO 3) at 1 and 10 bar hydrogen pressure. The selenite was found to be reduced to elemental Se, forming colloidal particles. At hydrogen pressures of 10 bar, the rate of selenite reduction was found to be linearly correlated to the fraction of Pd in the UO 2 pellets. No selenium was detected on the surface of the pellets. For the lowest Pd loading (0.1% Pd) the selenite reduction does not appear to proceed to completion indicating that the surface becomes less active.

  6. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    Science.gov (United States)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  7. Model Development of UO2-Zr Dispersion Plate-Type Fuel Behavior at Early Phase of Severe Accident and Molten Fuel Meat Relocation%UO2-Zr弥散燃料板严重事故早期行为与熔融芯体迁移模型研究

    Institute of Scientific and Technical Information of China (English)

    张卓华; 彭诗念; 于俊崇

    2014-01-01

    基于UO2-Zr弥散燃料板的结构与材料特性,利用已有的扩散、Nb-Zr反应以及UO2-Zr等材料学相关文献研究了UO2-Zr弥散燃料板严重事故过程中的氧化、固相反应以及熔融物迁移等特殊过程的机理模型,能为含UO2-Zr弥散燃料板堆芯的严重事故行为特性研究与安全分析提供参考.

  8. Radiation chemical synthesis and characterization of UO 2 nanoparticles

    Science.gov (United States)

    Roth, Olivia; Hasselberg, Hanna; Jonsson, Mats

    2009-01-01

    In a deep repository for spent nuclear fuel, U(VI)(aq) released upon dissolution of the fuel matrix could, in reducing parts of the system, be converted to U(IV) species which might coalesce and form nanometer-sized UO 2 particles. This type of particles is expected to have different properties compared to bulk UO 2(s). Hence, their properties, in particular the capacity for oxidant consumption, must be investigated in order to assess the effects of formation of such particles in a deep repository. In this work, methods for radiation chemical synthesis of nanometer-sized UO 2 particles, by electron- and γ-irradiation of U(VI) solutions, are presented. Electron-irradiation proved to be the most efficient method, showing high conversions of U(VI) and yielding small particles with a narrow size distribution (22-35 nm). Stable colloidal suspensions were obtained at low pH and ionic strength (pH 3, I = 0.03). Furthermore, the reactivity of the produced UO 2 particles towards H 2O 2 is investigated. The U(IV) fraction in the produced particles was found to be ˜20% of the total uranium content, and the results show that the UO 2 nanoparticles are significantly more reactive than micrometer-sized UO 2 when it comes to H 2O 2 consumption, the major part of the H 2O 2 being catalytically decomposed on the particle surface.

  9. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  10. Fabrication of micro-cell UO2-Mo pellet with enhanced thermal conductivity

    Science.gov (United States)

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

    2015-07-01

    As one of accident tolerant fuel pellets which should have features of good thermal conductivity and high fission product retention, a micro-cell UO2-Mo pellet has been studied in the aspect of fabrication and thermal property. It was intended to develop the compatible process with conventional UO2 pellet fabrication process. The effects of processing parameters such as the size and density of UO2 granule and the size of Mo powder have been studied to produce sound and dense pellet with completely connected uniform Mo cell-walls. The micro-cell UO2-Mo pellet consists of many Mo micro-cells and UO2 in them. The thermal conductivity of the micro-cell UO2-Mo pellet was measured and compared to those of the UO2 pellet and the UO2-Mo pellet with dispersed form of Mo particles. The thermal conductivity of the micro-cell UO2-Mo pellet was much enhanced and was found to be influenced by the Mo volumetric fraction and pellet integrity. A continuous Mo micro-cell works as a heat conducting channel in the pellet, greatly enhancing the thermal conductivity of the micro cell UO2-Mo pellet.

  11. Can H2 enhance the oxidative dissolution of UO2?

    Science.gov (United States)

    Barreiro Fidalgo, Alexandre; Jonsson, Mats

    2016-08-01

    Understanding the mechanism and kinetics of spent nuclear fuel dissolution in water is of key-importance for the safety assessment of deep geological repositories for spent nuclear fuel [1-5]. For UO2-based fuel, radiation induced oxidative dissolution of the fuel matrix is of considerable importance as this will enhance the release of fission products and actinides by several orders of magnitude [2-4]. This process has been studied extensively over several decades and can now be considered to be fairly well understood [2-6]. The aqueous radiolysis product identified as mainly being responsible for the oxidative dissolution of UO2 is H2O2[7]. In addition to oxidation of U(IV) to U(VI), H2O2 also undergoes catalytic decomposition on the UO2-surface [8,9]. In fact, it has been shown that catalytic decomposition is the major route on UO2-pellets [10,11]. In recent years it has been shown that this process involves the formation of surface bound hydroxyl radicals [12]. The mechanism of the catalytic decomposition is depicted in reaction (1-3).

  12. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  13. Chemical state of fission products in irradiated UO 2

    Science.gov (United States)

    Imoto, S.

    1986-08-01

    The chemical state of fission products in irradiated UO 2 fuel has been estimated for FBR as well as LWR on the basis of equilibrium calculation with the SOLGASMIX-PV code. The system considered for the calculation is composed of a gas phase, a CaF 2 type oxide phase, three grey phases, a noble metal alloy, a mixed telluride phase and several other phases each consisting of single compound. The distribution of elements into these phases and the amount of chemical species in each phase at different temperatures are obtained as a function of oxygen potential for LWR and FBR. Changes of the chemical potential of the fuel-fission products system during burnup are also evaluated with particular attention to the difference between LWR and FBR. Some informations obtained by the calculation are compared with the results of post irradiation examination of UO 2 fuels.

  14. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

    Science.gov (United States)

    Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Winckel, S. Van; Goll, W.; Manzel, R.

    2005-10-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G bar (O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G bar (O2) of the fuel increased from -370 kJ mol-1 at r/r0 = 0.1 to -293 kJ mol-1 at r/r0 = 0.975. Thus, the Δ G bar (O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  15. Determination of U3O8 in UO2 by infrared spectroscopy

    Directory of Open Access Journals (Sweden)

    Liliane Aparecida Silva

    Full Text Available Abstract The oxygen-uranium (O-U system has various oxides, such as UO2, U4O9, U3O8, and UO3. Uranium dioxide is the most important one because it is used as nuclear fuel in nuclear power plants. UO2 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 UO2. The presence of other oxides in the manufacturing of UO2 powder or sintered pellets is a critical factor. Fourier Transform Infrared Spectroscopy (FTIR was used to identify U3O8 in samples of UO2 powder. UO2 can be identified by bands at 340 cm-1 and 470 cm-1, and U3O8 and UO3 by bands at 735 cm-1, 910 cm-1, respectively. The methodology for sample preparation for FTIR spectra acquisition is presented, as well as the calibration for quantitative measurement of U3O8 in UO2. The content of U3O8 in partially calcined samples of UO2 powder was measured by FTIR with good agreement with X-rays diffractometry (XRD.

  16. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  17. Investigation on high temperature vapor pressure of UO 2 containing simulated fission-product elements

    Science.gov (United States)

    Yano, T.; Ohtsubo, A.; Ishii, T.

    1984-06-01

    During the hypothetical core disruptive accident (HCDA) of a fast breeder reactor (FBR), the temperature of the fuel would rise above 3000 K. The experimental data concerning the saturated fuel vapor pressure are necessary for the analysis of the HCDA. In this study, the UO 2 containing Cs, Ba, Ag, or Sn was used to simulate the irradiated fuel in the FBR. The saturated vapor pressure of pure UO 2 and UO 2 containing Cs, Ba, Ag, or Sn at 3000 to 5000 K was measured dynamically with a pulse laser and a torsion pendulum. The surface of a specimen on the pendulum was heated to eject vapor by the injection of a giant pulse ruby laser beam. The pressure of the ejected vapor was measured by both the maximum rotation angle of the pendulum and the duration of vapor ejection. The saturated vapor pressure was theoretically calculated by using the ejected vapor pressure. The surface temperature of the specimen was estimated from the irradiated energy density measured with a laser energy meter. The saturated vapor pressure of UO 2 at 3640 to 5880 K measured in this study was near the extrapolated value of Ackermann's low temperature data. The vapor pressure of UO 2 containing Cs, Ba, Ag or Sn was higher than that of UO 2. The saturated vapor pressure of UO 2 and a solid fission products system was calculated by using these experimental data.

  18. Realistic bandwidth estimation in the theoretically predicted radionuclide inventory of PWR-UO2 spent fuel derived from reactor design and operating data

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan

    2017-06-01

    Nuclear energy for power generation produces heat-generating high- and intermediate level radioactive waste (HLW and ILW) for which a safe solution for the handling and disposal has to be found. Currently, many European countries consider the final disposal of HLW and ILW in deep geological formations as the most preferable option. In Germany the main stream of HLW and ILW include spent fuel assemblies from nuclear power plants (NPPs), the vitrified waste and compacted metallic waste of the fuel assembly structural parts originate from reprocessing plants. An important task that occurs within the framework of the Product Quality Control (PQC) of nuclear waste is the assessment of the compliance of any reprocessed waste product inventory with the prescribed limits for each relevant radionuclide (RN). The PQC task is to verify the required quality and safety of nuclear waste prior to transportation to a German repository and to avert the disposal of non-conform waste packages. The verification is usually based on comparing the declared radionuclide inventory of the waste with the presumed or expected composition, which is estimated, based on the known history of the waste and its processing. The difficulty of such estimations for radioactive components from nuclear fuel assemblies is that reactor design parameters and operating histories can have a significant influence on the nuclide inventory of any individual fuel assembly. Thus, knowledge of these parameters is a key issue to determine the realistic concentration ranges, or bandwidths, of the radionuclide inventory. As soon as a governmental decision on the construction of a high-level waste repository will be made, comprehensive radionuclide inventories of the wastes assigned for the deposition will be required. The list of final repository relevant radionuclide is based on the safety assessment for this particular repository, thus it is likely to comprise more-or-less the same radionuclides that need to be

  19. Preparing UO2 kernels by gelcasting

    Institute of Scientific and Technical Information of China (English)

    GUO Wenli; LIANG Tongxiang; ZHAO Xingyu; HAO Shaochang; LI Chengliang

    2009-01-01

    A process named gel-casting has been developed for the production of dense UO2 kernels for the high-ten-temperature gas-cooled reactor. Compared with the sol-gel process, the green microspheres can be got by dispersing the U3O8 slurry in gelcasting process, which means that gelcasting is a more facilitative process with less waste in fabricating UO2 kernels. The heat treatment.

  20. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

    Indian Academy of Sciences (India)

    Seyed Amir Hossein Feghhi; Marzieh Rezazadeh; Yachine Kadi; Claudio Tenreiro; Morteza Aref; Zohreh Gholamzadeh

    2013-01-01

    The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of 233U converted from 232Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.

  1. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Science.gov (United States)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  2. Behavior of UO2 and fissium in sodium vapor atmosphere at temperatures up to 2800 C

    Science.gov (United States)

    Feuerstein, H.; Oschinski, J.

    1986-11-01

    An experimental technique was developed to study the behavior of fuel and fission products in out-of-pile tests in a sodium vapor atmosphere. Evaporation rates of UO2 were measured up to 2800 C. The evaporation is found to depend on temperature and the active surface. Evaporation restructures the surface of the samples, however no new active surface is formed; UO2 can form well shaped crystals and curious erosion products. The efficiency of the used condenser/filter lines is 99.99%. In an HCDA, all the evaporated substances condense in the sodium pool. Thermal reduction of the UO2 reduces the oxygen potential of the system. The final composition at 2500 C is UO1.95. The only influence of the sodium vapor is found for the diffusion of UO2 into the thoria of the crucible. Compared with experiments in an atmosphere of pure argon, the diffusion rate is reduced.

  3. H 2O 2 and radiation induced dissolution of UO 2 and SIMFUEL pellets

    Science.gov (United States)

    Nilsson, Sara; Jonsson, Mats

    2011-03-01

    Dissolution of the UO 2 matrix is of major importance in the safety assessment of a future deep repository for spent nuclear fuel. The aim of this work is to elucidate if the observed differences in dissolution rates between SIMFUEL and UO 2 can be attributed to differences in oxidant reactivity towards these two materials. To elucidate this, the oxidative dissolution of U(VI) and consumption of H 2O 2 have been studied for UO 2 and SIMFUEL pellets under N 2 and H 2 atmosphere. The H 2O 2 and U(VI) concentrations have been measured as a function of reaction time. In addition, γ-radiation induced dissolution UO 2 and SIMFUEL pellets have been studied. The experiments show that while the reactivity of the two types of pellets towards H 2O 2 is almost identical and in good agreement with the previously determined rate constant for the reaction, the dissolution rates differ considerably. The significantly lower rate of dissolution of the SIMFUEL pellet is attributed to an increased fraction of catalytic decomposition of H 2O 2. The radiation chemical experiments reveal a similar but less pronounced difference between the two types of pellets. This implies that the relative impact of the radiolytic oxidants in radiation induced UO 2 dissolution differs between a pure UO 2 pellet and SIMFUEL.

  4. The influence of particle size on the kinetics of UO 2 oxidation in aqueous powder suspensions

    Science.gov (United States)

    Roth, Olivia; Bönnemark, Tobias; Jonsson, Mats

    2006-07-01

    Previous studies have indicated that the rate of a heterogeneous liquid-solid reaction depends on the size of the solid particles. It has been suggested that both the pre-exponential factor and the activation energy depend on the particle size. The processes involved in dissolution of UO 2 have been extensively studied because of their importance for the safety analysis of a future deep repository for spent nuclear fuel and in many of these studies powder suspensions of UO 2 are used as a model system. Therefore, it is of importance to investigate and quantify the particle size effect on the kinetics of UO 2 oxidation in order to enable comparison of data from studies on different solid substrates. In this work the influence of particle size on the second order rate constant and on the activation energy of the reaction between MnO4- and UO 2 was studied using aqueous UO 2-particle suspensions of four different size distributions. A comparative study of the activation energy for the reaction using a UO 2 pellet was also performed.

  5. Adsorption behaviour of PuF6 on UO2F2 by the use of 236Pu

    Science.gov (United States)

    Sato, Nobuaki; Matsuda, Minoru; Mitsugashira, Toshiaki; Kirishima, Akira

    2010-03-01

    To know the behavior of plutonium in the fluoride volatility process (FLUOREX PROCESS) for the spent nuclear fuel, both UO2 and PuO2 are fluorinated by fluorine forming volatile UF6 and PuF6, respectively. Then PuF6 is separated and recovered from UF6 by using adsorption materials such as uranyl fluoride UO2F2. In this paper, adsorption behavior of PuF6on UO2F2 was examined by the use of 236Pu tracer. First, the stability of UO2F2 in F2atmosphere was analyzed by TG-DTA method showing that uranium volatilized completely over 350 °C by the formation of UF6 and the adsorption of plutonium by UO2F2 should be done at temperatures lower than 250 °C. The behavior of PtF6 as a chemical analogue of PuF6 was also conducted for comparison and it showed that the deposition of PtF4 on UO2F2 at 200 °C. When the 236Pu doped U3O8 was reacted with 10%F2-He gas, the PuF6 vaporized at ca. 600 °C. Then adsorption of 236Pu on UO2F2 was observed by α ray measurement. The adsorption mechanism of Pu on UO2F2 was discussed with experimental data and thermodynamic consideration.

  6. Development Status of a CVD System to Deposit Tungsten onto UO2 Powder via the WCI6 Process

    Science.gov (United States)

    Mireles, O. R.; Kimberlin, A.; Broadway, J.; Hickman, R.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) is under development for deep space exploration. NTP's high specific impulse (> 850 second) enables a large range of destinations, shorter trip durations, and improved reliability. W-60vol%UO2 CERMET fuel development efforts emphasize fabrication, performance testing and process optimization to meet service life requirements. Fuel elements must be able to survive operation in excess of 2850 K, exposure to flowing hydrogen (H2), vibration, acoustic, and radiation conditions. CTE mismatch between W and UO2 result in high thermal stresses and lead to mechanical failure as a result UO2 reduction by hot hydrogen (H2) [1]. Improved powder metallurgy fabrication process control and mitigated fuel loss can be attained by coating UO2 starting powders within a layer of high density tungsten [2]. This paper discusses the advances of a fluidized bed chemical vapor deposition (CVD) system that utilizes the H2-WCl6 reduction process.

  7. Numerical characterization of micro-cell UO2sbnd Mo pellet for enhanced thermal performance

    Science.gov (United States)

    Lee, Heung Soo; Kim, Dong-Joo; Kim, Sun Woo; Yang, Jae Ho; Koo, Yang-Hyun; Kim, Dong Rip

    2016-08-01

    Metallic micro-cell UO2 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 UO2 fuels for further enhancement of thermal performance are lacking. Here, we numerically investigated the thermal conductivities of micro-cell UO2sbnd 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 UO2sbnd Mo pellets (5 vol% Mo) with modified geometries increased the thermal conductivity of the current UO2 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.

  8. On the redox reactivity of doped UO2 pellets - Influence of dopants on the H2O2 decomposition mechanism

    Science.gov (United States)

    Pehrman, Reijo; Trummer, Martin; Lousada, Cláudio M.; Jonsson, Mats

    2012-11-01

    The reactivity of doped UO2 such as SIMFUEL, Y2O3 doped UO2 and Y2O3/Pd doped UO2 towards H2O2 has been shown to be fairly similar to that of pure UO2. However, the oxidative dissolution yield, i.e. the ratio between the amount of dissolved uranium and the amount of consumed H2O2 is significantly lower for doped UO2. The rationale for the observed differences in dissolution yield is a difference in the ratio between the rates of the two possible reactions between H2O2 and the doped UO2. In this work we have studied the effect of doping on the two possible reactions, electron-transfer and catalytic decomposition. The catalytic decomposition was studied by monitoring the hydroxyl radical production (the primary product) as a function of time. The redox reactivity of the doped pellets was studied by using MnO4- and IrCl62- as model oxidants, only capable of electron-transfer reactions with the pellets. In addition, the activation energies for oxidation of UO2 and SIMFUEL by MnO4- were determined experimentally. The experiments show that the rate of catalytic decomposition of H2O2 varies by 30% between the most and least reactive material. This is a negligible difference compared to the difference in oxidative dissolution yield. The redox reactivity study shows that doping UO2 influences the redox reactivity of the pellet. This is further illustrated by the observed activation energy difference for oxidation of UO2 and SIMFUEL by MnO4-. The redox reactivity study also shows that the sensitivity to dopants increases with decreasing reduction potential of the oxidant. These findings imply that the relative impact of radiolytic oxidants in oxidative dissolution of spent nuclear fuel must be reassessed taking the actual fuel composition into account.

  9. The heat capacity and enthalpy of condensed UO 2: Critical review and assessment

    Science.gov (United States)

    Hyland, G. J.; Ohse, R. W.

    1986-09-01

    Having established the role of the heat capacity, Cp( T), of condensed UO 2 in various FBR accident scenarios, e.g. HCDA and PAHR, and having noted the unsatisfactory state of present knowledge concerning this basic thermophysical property of the fuel, all existing enthalpy and heat capacity data are collated and assessed, and certain recommendations made. The conventional method of obtaining Cp( T) by analytical differentiation of some adopted fit to this enthalpy data is then critically examined. The attendant problems are illustrated both for solid UO 2, where the contribution to Cp( T) from the weak, sigmoidal, enthalpy structure (which is just discernible in the data of Hein and Flagella) is missed and for molten UO 2, where not even the direction of the trend of Cp( T) with T can be definitively established, resulting, upon extrapolation to 5000 K, in Cp values which can differ by as much as 60 J mol -1K -1. Some recent progress towards a more acceptable, "model-independent" approach, known as quasi-local linear regression (QLLR), is then reviewed and applied to enthalpy data of UO 2 on both sides of its melting point, Tm. In the case of solid UO 2, a pronounced heat capacity peak, extending over about 100 K and centred on 2610 K., is revealed, whose magnitude and location is very similar to that found in other fluorite structured materials near 0.8 Tm wherein it indicates a (Bredig) transition to a state characterised by giant ionic conductivities. Whilst it is impossible to establish any definite T-dependence for the Cp(QLLR) values in molten UO 2, the tendency to slightly decrease appears to marginally outweigh the converse, in qualitative accord with the dependence advocated by Hoch and Vernardakis. In the post-transitional region Tt< T< Tm the opposite holds, as is necessary for consistency between the independently established T-dependences of the thermal conductivity and diffusivity, which requires that Cp( T) increases with T faster than the density

  10. Radiation-induced decomposition of U(VI) phases to nanocrystals of UO"2 [rapid communication

    Science.gov (United States)

    Utsunomiya, Satoshi; Ewing, Rodney C.; Wang, Lu-Min

    2005-12-01

    U 6+-phases are common alteration products, under oxidizing conditions, of uraninite and the UO 2 in spent nuclear fuel. These U 6+-phases are subjected to a radiation field caused by the α-decay of U, or in the case of spent nuclear fuel, incorporated actinides, such as 239Pu and 237Np. In order to evaluate the effects of α-decay events on the stability of the U 6+-phases, we report, for the first time, the results of ion beam irradiations (1.0 MeV Kr 2+) of U 6+-phases. The heavy-particle irradiations are used to simulate the ballistic interactions of the recoil-nucleus of an α-decay event with the surrounding structure. The Kr 2+-irradiation decomposed the U 6+-phases to UO 2 nanocrystals at doses as low as 0.006 displacements per atom (dpa). U 6+-phases accumulate substantial radiation doses (˜1.0 displacement per atom) within 100,000 yr if the concentration of incorporated 239Pu is as high as 1 wt.%. Similar nanocrystals of UO 2 were observed in samples from the natural fission reactors at Oklo, Gabon. Multiple cycles of radiation-induced decomposition to UO 2 followed by alteration to U 6+-phases provide a mechanism for the remobilization of incorporated radionuclides.

  11. Spectroscopy of the UO2+ cation and the delayed ionization of UO2

    Science.gov (United States)

    Merritt, Jeremy M.; Han, Jiande; Heaven, Michael C.

    2008-02-01

    Vibronically resolved spectra for the UO2+ cation have been recorded using the pulsed field ionization zero electron kinetic energy (PFI-ZEKE) technique. For the ground state, long progressions in both the bending and symmetric stretch vibrations were observed. Bend and stretch progressions of the first electronically excited state were also observed, and the origin was found at an energy of 2678cm-1 above the ground state zero-point level. This observation is consistent with a recent theoretical prediction [Infante et al., J. Chem. Phys. 127, 124308 (2007)]. The ionization energy for UO2, derived from the PFI-ZEKE spectrum, namely, 6.127(1)eV, is in excellent agreement with the value obtained from an earlier photoionization efficiency measurement. Delayed ionization of UO2 in the gas phase has been reported previously [Han et al., J. Chem. Phys. 120, 5155 (2004)]. Here, we extend the characterization of the delayed ionization process by performing a quantitative study of the ionization rate as a function of the energy above the ionization threshold. The ionization rate was found to be 5×106s-1 at threshold, and increased linearly with increasing energy in the range investigated (0-1200cm-1).

  12. 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.

  13. Methodology for Producing a Uniform Distribution of UO2 in a Tungsten Matrix

    Science.gov (United States)

    Tucker, Dennis S.; O'Conner, Andrew; Hickman, Rickman; Broadway, Jeramie; Belancik, Grace

    2015-01-01

    Current work at NASA's Marshall Space Flight Center (MSFC) is focused on the development CERMET fuel materials for Nuclear Thermal Propulsion (NTP). The CERMETs consist of uranium dioxide (UO2) fuel particles embedded in a tungsten (W) metal matrix. Initial testing of W-UO2 samples fabricated from fine angular powders performed reasonably well, but suffered from significant fuel loss during repeated thermal cycling due to agglomeration of the UO2 (1). The blended powder mixtures resulted in a non-uniform dispersion of the UO2 particles in the tungsten matrix, which allows rapid vaporization of the interconnected UO2 from the sample edges into the bulk material. Also, the angular powders create areas of stress concentrations due to thermal expansion mismatch, which eventually cracks the tungsten matrix. Evenly coating spherical UO2 particles with chemical vapor deposited (CVD) tungsten prior to consolidation was previously demonstrated to provide improved performance. However, the CVD processing technology is expensive and not currently available. In order to reduce cost and enhance performance, a powder coating process has been developed at MSFC to produce a uniform distribution of the spherical UO2 particles in a tungsten matrix. The method involves utilization of a polyethylene binder during mixing which leads to fine tungsten powders clinging to the larger UO2 spherical particles. This process was developed using HfO2 as a surrogate for UO2. Enough powder was mixed to make 8 discs (2cm diameter x 8mm thickness) using spark plasma sintering. A uniaxial pressure of 50 MPa was used at four different temperatures (2 samples at each temperature). The first two samples were heated to 1400C and 1500C respectively for 5 minutes. Densities for these samples were less than 85% of theoretical, so the time at temperature was increased to 20 minutes for the remaining samples. The highest densities were achieved for the two samples sintered at 1700C (approx. 92% of

  14. Development of the MFPR model for fission gas release in irradiated UO2 under transient conditions

    Directory of Open Access Journals (Sweden)

    Veshchunov Michael S.

    2017-01-01

    Full Text Available The fission gas release microscopic model of the mechanistic code MFPR is further developed for modelling of enhanced release from irradiated UO2 fuel under transient conditions of the power ramp tests, along with the microstructure evolution characterised by the formation of a new population of large intragranular bubbles with a rather wide size distribution (from 30 to 500 nm, observed in transient-tested UO2 fuel samples. Implementation of the additional microscopic mechanisms results in a notable improvement of the code predictions (in comparison with the previous code version for the fractional gas release in the Risø ramp tests with three different hold times of 3, 40 and 62 h at the terminal linear power of ≈40 kW/m.

  15. Effects of Fe(II) and hydrogen peroxide interaction upon dissolving UO2 under geologic repository conditions.

    Science.gov (United States)

    Amme, M; Bors, W; Michel, C; Stettmaier, K; Rasmussen, G; Betti, M

    2005-01-01

    Iron redox cycling is supposed to be one of the major mechanisms that control the geochemical boundary conditions in the near field of a geologic repository for UO2 spent nuclear fuel. This work investigates the impact of reactions between hydrogen peroxide (H2O2) and iron (Fe2+/Fe3+) on UO2 dissolution. The reaction partners were contacted with UO2 in oxygen-free batch reactor tests. The interaction in absence of UO2 gives a stoichiometric redox reaction of Fe2+ and H2O2 when the reactants are present in equal concentration. Predomination of H202 results in its delayed catalytic decomposition. With UO2 present, its dissolution is controlled by either a slow mechanism (as typical for anoxic environments) or uranium peroxide precipitation, depending strongly on the reactant ratio. Uranium peroxide (UO4 x nH2O, m-studtite), detected on UO2 surfaces after exposure to H2O2, was not found on the surfaces exposed to solutions with stoichometric Fe(II)/ H2O2 ratios. This suggests that H2O2 was deactivated in redox reactions before a formation of UO4 took place. ESR measurements employing the spin trapping technique revealed only the DMPO-OH adduct within the first minutes after the reaction start (high initial concentrations of the OH radical); however, in the case of Fe(II) and H2O2 reacting at 10(-4) mol/L with UO2, dissolved oxygen and Fe2+ concentrations indicate the participation of further Fe intermediates and, therefore, Fenton redox activities.

  16. Update on Fresh Fuel Characterization of U-Mo Alloys

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Burkes; D. M. Wachs; D. D. Keiser; M. A. Okuniewski; J. F. Jue; F. J. Rice; R. Prabhakaran

    2009-03-01

    The need to provide more accurate property information on U-Mo fuel alloys to operators, modellers, researchers, fabricators, and government increases as success of the GTRI Reactor Convert program continues. This presentation provides an update on fresh fuel characterization activities that have occurred at the INL since the RERTR 2008 conference in Washington, D.C. The update is particularly focused on properties recently obtained and on the development progress of new measurement techniques. Furthermore, areas where useful and necessary information is still lacking is discussed. The update deals with mechanical, physical, and microstructural properties for both integrated and separate effects. Appropriate discussion of fabrication characteristics, impurities, thermodynamic response, and effects on the topic areas are provided, along with a background on the characterization techniques used and developed to obtain the information. Efforts to measure similar characteristics on irradiated fuel plates are discussed.

  17. 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

  18. SCDAP/RELAP5分析UO2-Zr板型元件严重事故的方法研究%Approach for Simulating Severe Accident of UO2-Zr Plate by SCDAP/RELAP5

    Institute of Scientific and Technical Information of China (English)

    张卓华; 彭诗念; 黄善仿; 于俊崇

    2013-01-01

    SCDAP/RELAP5是一种常见的机理性严重事故分析程序,能够分析多种类型的堆芯构件.通过对比分析SCDAP/RELAP5程序模拟棒形燃料元件与板型燃料元件堆芯在严重事故下行为的分析模型,结合UO2-Zr板型状元件堆芯的特性,提出了运用并改进SCDAP/RELAP5程序模拟UO2-Zr板型元件堆芯在严重事故下行为的研究方案.对程序结构的分析结果表明,SCDAP/RELAP5程序部分结构和模型适用于对UO2-Zr板型元件进行基本的严重事故分析,但需要通过创建新部件、研究新模型,并与已有模型的重新组合搭配才能较为精准地模拟UO2-Zr板型元件严重事故的实际行为.%As a common mechanistic code for safety analysis of severe accident,SCDAP/RELAP5 can simulate many types of core components phenomenon during severe accidents.Comparison of simulation model of fuel behavior under severe accident between fuel rod and ATR plate is described in this paper and the approach for simulating severe accident of UO2-Zr plate is concluded by combining structure properties of UO2-Zr.It is concluded that the basic analysis of severe accident of UO2-Zr plate could be achieved by S/R code from the code simulation.However,new core structure,new model of fuel behavior and combination of existing model should be developed in S/R code to simulate the precise core behavior of reactor assembled with UO2-Zr plate under severe accidents.

  19. Natural analogues to the spent fuel behaviour of radioactive wastes (MATRIX, FASES I y II projects); Analogos naturales de la liberacion y migracion del UO2 y elementos metalicos asociados (Proyecto MATRIX, FASES I y II)

    Energy Technology Data Exchange (ETDEWEB)

    Perez del Villa, L.; Campos, R.; Garralon, A.; Crespo, M. T.; Quejido, J. A.; Cozar, J. S.; Arcos, D.; Bruno, J.; Grive, M.; Domenech, C.; Duro, L.; Ruiz Sanchez-Prro, J.; Marin, F.; Izquierdo, A.; Cattetero, G.; Ortuno, F.; Floria, E.

    2005-07-01

    Uranium ore deposits have been extensively studied as natural analogues to the spent fuel behaviour of radioactive wastes. These investigations constitute an essential element of both national and international research programmes applied to the assessment of HLNW repositories and their interaction with the environment. The U ore deposit of Mina Fe (Ciudad Rodrigo, Salamanca) is hosted in highly fractured schistose rocks, a geological setting that has not been envisaged in the ENRESA option for nuclear waste disposal. However, the processes occurring at Mina Fe maintain some analogies with those occurring in a HLNW repository: The existence of large U concentrations as pitchblende (UO{sub 2}+x), which is chemically analogous to the main component of spent nuclear fuel, which has an oxidation degree of 2.25 < x < 2.66 as a result of radiolytic oxidation. The solubility behaviour of pitchblende as a result of interaction with groundwaters of varying chemical composition can be used to validate predictive models for spent fuel stability under severe alteration conditions. Some of the weathering products of pitchblende are similar to those that have been identified during the experimental oxidative dissolution of UO{sub 2}, Sim fuel, as well as natural uraninite and pitchblende. This is a subject that has been previously investigated in other research projects. Fe(III)-oxy hydroxides in the oxidised zone of the deposit could be similar to the spent fuel container corrosion products that could be formed under redox transition conditions. These corrosion products may act as radionuclide and trace metal scavengers. (Author)

  20. Ab initio calculation of oxygen self-diffusion coefficient in uranium dioxide UO2

    Science.gov (United States)

    Dorado, Boris; Garcia, Philippe; Torrent, Marc

    Uranium dioxide UO2 is the most widely used nuclear fuel worldwide and its atomic transport properties are relevant to practically all engineering aspects of the material. Although transport properties have already been studied in UO2 by means of first-principles calculations, the ab initio determination of self-diffusion coefficients has up to now remained unreachable because the relevant computational tools were neither available or adapted. The present work reports our results related to the ab initio calculation of the oxygen self-diffusion coefficient in UO2. We first determine the Gibbs free energies of formation of oxygen charged defects by calculating both the electronic and vibrational (hence entropic) contributions. Then, we use the transition state theory in order to compute the effective jump frequency of the defects, which in turn provides us with the value of the pre-exponential factor. The results are compared to self-diffusion data obtained experimentally with a careful monitoring of the relevant thermodynamic conditions (oxygen partial pressure, temperature, impurity content).

  1. Onset conditions for flash sintering of UO2

    Science.gov (United States)

    Raftery, Alicia M.; Pereira da Silva, João Gustavo; Byler, Darrin D.; Andersson, David A.; Uberuaga, Blas P.; Stanek, Christopher R.; McClellan, Kenneth J.

    2017-09-01

    In this work, flash sintering was demonstrated on stoichiometric and non-stoichiometric uranium dioxide pellets at temperatures ranging from room temperature (26 °C) up to 600 °C . The onset conditions for flash sintering were determined for three stoichiometries (UO2.00, UO2.08, and UO2.16) and analyzed against an established thermal runaway model. The presence of excess oxygen was found to enhance the flash sintering onset behavior of uranium dioxide, lowering the field required to flash and shortening the time required for a flash to occur. The results from this study highlight the effect of stoichiometry on the flash sintering behavior of uranium dioxide and will serve as the foundation for future studies on this material.

  2. Evaluation of sintering effects on SiC-incorporated UO2 kernels under Ar and Ar-4%H2 environments

    Science.gov (United States)

    Silva, Chinthaka M.; Lindemer, Terrence B.; Hunt, Rodney D.; Collins, Jack L.; Terrani, Kurt A.; Snead, Lance L.

    2013-11-01

    Silicon carbide (SiC) is suggested as an oxygen getter in UO2 kernels used for tristructural isotropic (TRISO) particle fuels and to prevent kernel migration during irradiation. Scanning electron microscopy and X-ray diffractometry analyses performed on sintered kernels verified that an internal gelation process can be used to incorporate SiC in UO2 fuel kernels. Even though the presence of UC in either argon (Ar) or Ar-4%H2 sintered samples suggested a lowering of the SiC up to 3.5-1.4 mol%, respectively, the presence of other silicon-related chemical phases indicates the preservation of silicon in the kernels during sintering process. UC formation was presumed to occur by two reactions. The first was by the reaction of SiC with its protective SiO2 oxide layer on SiC grains to produce volatile SiO and free carbon that subsequently reacted with UO2 to form UC. The second process was direct UO2 reaction with SiC grains to form SiO, CO, and UC. A slightly higher density and UC content were observed in the sample sintered in Ar-4%H2, but both atmospheres produced kernels with ˜95% of theoretical density. It is suggested that incorporating CO in the sintering gas could prevent UC formation and preserve the initial SiC content.

  3. Optimization of a Wcl6 CVD System to Coat UO2 Powder with Tungsten

    Science.gov (United States)

    Belancik, Grace A.; Barnes, Marvin W.; Mireles, Omar; Hickman, Robert

    2015-01-01

    In order to achieve deep space exploration via Nuclear Thermal Propulsion (NTP), Marshall Space Flight Center (MSFC) is developing W-UO2 CERMET fuel elements, with focus on fabrication, testing, and process optimization. A risk of fuel loss is present due to the CTE mismatch between tungsten and UO2 in the W-60vol%UO2 fuel element, leading to high thermal stresses. This fuel loss can be reduced by coating the spherical UO2 particles with tungsten via H2/WCl6 reduction in a fluidized bed CVD system. Since the latest incarnation of the inverted reactor was completed, various minor modifications to the system design were completed, including an inverted frit sublimer. In order to optimize the parameters to achieve the desired tungsten coating thickness, a number of trials using surrogate HfO2 powder were performed. The furnace temperature was varied between 930 C and 1000degC, and the sublimer temperature was varied between 140 C and 200 C. Each trial lasted 73-82 minutes, with one lasting 205 minutes. A total of 13 trials were performed over the course of three months, two of which were re-coatings of previous trials. The powder samples were weighed before and after coating to roughly determine mass gain, and Scanning Electron Microscope (SEM) data was also obtained. Initial mass results indicated that the rate of layer deposition was lower than desired in all of the trials. SEM confirmed that while a uniform coating was obtained, the average coating thickness was 9.1% of the goal. The two re-coating trials did increase the thickness of the tungsten layer, but only to an average 14.3% of the goal. Therefore, the number of CVD runs required to fully coat one batch of material with the current configuration is not feasible for high production rates. Therefore, the system will be modified to operate with a negative pressure environment. This will allow for better gas mixing and more efficient heating of the substrate material, yielding greater tungsten coating per trial.

  4. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  5. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  6. Comparative study of the structural and electrochemical properties of noble metal inclusions in a UO2 matrix

    Science.gov (United States)

    Stumpf, S.; Petersmann, T.; Seibert, A.; Gouder, T.; Huber, F.; Brendebach, B.; Denecke, M. A.

    2010-03-01

    The intention of the presented study is to elucidate the influence of noble metal inclusions (fission products) on the structure as well as on the electrochemical properties of spent nuclear fuel (SNF). To this aim, thin UO2 films doped with metal inclusions such as Pd, Mo and Au are prepared by sputter deposition. The films are characterized by spectroscopic (XPS, EXAFS, XRD) as well as by microscopic (AFM, SEM) methods. In a next step the electrochemical properties of these model systems are comparatively investigated by cyclo voltammetry (CV). The sputter technique in combination with the heating treatment of the films allows the formation of a crystalline UO2 matrix as it is found in SNF. The co-deposition with Au results in the dispersion of the pure metal in the oxide matrix. Pd as well as Mo are oxidized due to the deposition at RT. Heating the films involves a further oxidation of MoO2 to MoO3. By contrast Pd agglomerates and forms metallic -phases as it is found in SNF. Electrochemical investigations of the UO2-Pd samples indicate an inhibiting influence of Pd on the oxidative dissolution of UO2. When it comes to the formation of secondary phases under reducing conditions such influence is passivated. The precipitates finally dominate the overall redox behaviour of the model system.

  7. Speciation of residual carbon contained in UO2

    Science.gov (United States)

    Ziouane, Yannis; Arab-Chapelet, Bénédicte; Tamain, Christelle; Lalleman, Sophie; Delahaye, Thibaud; Leturcq, Gilles

    2016-12-01

    UO2 powders were synthesized thanks to oxalic precipitation (platelet morphology) and sol-gel route and completely characterized. A secondary phase was found depending on the calcination atmospheres. This phase has been identified by Raman spectroscopy as graphitic material (i.e. carbon-based secondary compound) and quantified by thermogravimetric analyses. Its amount varies with the calcination atmosphere. The presence of this secondary phase has no significant effect on the lattice parameter and its specific surface area.

  8. Post-irradiation examinations and high-temperature tests on undoped large-grain UO2 discs

    Science.gov (United States)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.

    2015-07-01

    Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants -in the form of thin circular discs - were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/tHM. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO2 discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/tHM. Detailed characterizations of some of these irradiated large-grain UO2 discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO2 discs irradiated under the same conditions. Examination of the high burn-up large-grain UO2 discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO2 discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/tHM discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel's microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms at high burn-up and where bubbles appear during the low

  9. Microbeam x-ray absorption spectroscopy study of chromium in large-grain uranium dioxide fuel

    Science.gov (United States)

    Mieszczynski, C.; Kuri, G.; Bertsch, J.; Martin, M.; Borca, C. N.; Delafoy, Ch; Simoni, E.

    2014-09-01

    Synchrotron-based microprobe x-ray absorption spectroscopy (XAS) has been used to study the local atomic structure of chromium in chromia-doped uranium dioxide (UO2) grains. The specimens investigated were a commercial grade chromia-doped UO2 fresh fuel pellet, and materials from a spent fuel pellet of the same batch, irradiated with an average burnup of ~40 MW d kg-1. Uranium L3-edge and chromium K-edge XAS have been measured, and the structural environments of central uranium and chromium atoms have been elucidated. The Fourier transform of uranium L3-edge extended x-ray absorption fine structure shows two well-defined peaks of U-O and U-U bonds at average distances of 2.36 and 3.83 Å. Their coordination numbers are determined as 8 and 11, respectively. The chromium Fourier transform extended x-ray absorption fine structure of the pristine UO2 matrix shows similar structural features with the corresponding spectrum of the irradiated spent fuel, indicative of analogous chromium environments in the two samples studied. From the chromium XAS experimental data, detectable next neighbor atoms are oxygen and uranium of the cation-substituted UO2 lattice, and two distinct subshells of chromium and oxygen neighbors, possibly because of undissolved chromia particles present in the doped fuels. Curve-fitting analyses using theoretical amplitude and phase-shift functions of the closest Cr-O shell and calculations with ab initio computer code FEFF and atomic clusters generated from the chromium-dissolved UO2 structure have been carried out. There is a prominent reduction in the length of the adjacent Cr-O bond of about 0.3 Å in chromia-doped UO2 compared with the ideal U-O bond length in standard UO2 that would be expected because of the change in effective Coulomb interactions resulting from replacing U4+ with Cr3+ and their ionic size differences. The contraction of shortest Cr-U bond is ~0.1 Å relative to the U-U bond length in bulk UO2. The difference in the

  10. Molecular Dynamics Simulation of Thermal Transport in UO2 Containing Uranium, Oxygen, and Fission-product Defects

    Science.gov (United States)

    Liu, X.-Y.; Cooper, M. W. D.; McClellan, K. J.; Lashley, J. C.; Byler, D. D.; Bell, B. D. C.; Grimes, R. W.; Stanek, C. R.; Andersson, D. A.

    2016-10-01

    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, thereby 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 are 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 UO2 +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 are 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-temperature model typically used in fuel-performance codes

  11. Leaching of UO2 pellets doped with alpha-emitters (238/239Pu) in synthetic deep Callovian-Oxfordian groundwater

    Science.gov (United States)

    Tribet, M.; Jégou, C.; Broudic, V.; Marques, C.; Rigaux, P.; Gavazzi, A.

    2010-03-01

    The reactivity of a polycrystalline UO2 surface under alpha irradiation in contact with groundwater is investigated, in the hypothesis of direct disposal of spent fuel in a deep geological repository. Two series of plutonium-doped UO2 samples (specific alpha activity of 18 and 385 MBq·g-1UO2) were leached in a synthetic Callovian-Oxfordian deep groundwater under anoxic conditions (Ar/CO2 3000 ppm, 3.5 bar relative pressure) to assess both the impact of alpha radiolysis of water and the complexing capacity of the groundwater ions on the dissolution of UO2. This study follows a prior one performed in pure and carbonated waters. Firstly, technical developments were necessary for the analyses in the groundwater solution because of its high salt concentrations: quantification limits were determined for the measurement of uranium and radiolytic H2O2 traces in this medium. Secondly, given the very high reactivity of these samples in the presence of air and in order to minimize any prior surface oxidation, a strict experimental protocol was followed, based on high-temperature annealing in Ar + 4% H2 with preleaching cycles. Each type of UO2 pellet was then leached under static conditions for 30 days (anoxic conditions, deep groundwater solutions). Results on the evolution of uranium releases are presented. For the lowest alpha activity (18 MBq·g-1UO2), uranium releases in groundwater were below the quantification limit of 2 × 10-8 mol·L-1 with a kinetic phosphorescence analyzer, even after 30 days. However, for higher alpha activity (385 MBq·g-1UO2) the uranium releases begin to exceed the quantification limit after 14 days of leaching and then increase exponentially. This increase is comparable to results previously obtained in carbonated solutions.

  12. UO(2) 2+ speciation determines uranium toxicity and bioaccumulation in an environmental Pseudomonas sp. isolate.

    Science.gov (United States)

    Vanengelen, Michael R; Field, Erin K; Gerlach, Robin; Lee, Brady D; Apel, William A; Peyton, Brent M

    2010-04-01

    In the present study, experiments were performed to investigate how representative cellulosic breakdown products, when serving as growth substrates under aerobic conditions, affect hexavalent uranyl cation (UO(2) (2+)) toxicity and bioaccumulation within a Pseudomonas sp. isolate (designated isolate A). Isolate A taken from the Cold Test Pit South (CTPS) region of the Idaho National Laboratory (INL), Idaho Falls, ID, USA. The INL houses low-level uranium-contaminated cellulosic material and understanding how this material, and specifically its breakdown products, affect U-bacterial interactions is important for understanding UO(2) (2+) fate and mobility. Toxicity was modeled using a generalized Monod expression. Butyrate, dextrose, ethanol, and lactate served as growth substrates. The potential contribution of bicarbonate species present in high concentrations was also investigated and compared with toxicity and bioaccumulation patterns seen in low-bicarbonate conditions. Isolate A was significantly more sensitive to UO(2) (2+) and accumulated significantly more UO(2) (2+) in low-bicarbonate concentrations. In addition, UO(2) (2+) growth inhibition and bioaccumulation varied depending on the growth substrate. In the presence of high bicarbonate concentrations, sensitivity to UO(2) (2+) inhibition was greatly mitigated, and did not vary between the four substrates tested. The extent of UO(2) (2+) accumulation was also diminished. The observed patterns were related to UO(2) (2+) aqueous complexation, as predicted by MINTEQ (ver. 2.52) (Easton, PA, USA). In the low- bicarbonate medium, the presence of positively charged and unstable UO(2) (2+)-hydroxide complexes explained both the greater sensitivity of isolate A to UO(2) (2+), and the ability of isolate A to accumulate significant amounts of UO(2) (2+). The exclusive presence of negatively charged and stable UO(2) (2+)-carbonate complexes in the high bi-carbonate medium explained the diminished sensitivity of

  13. UO2核芯制备工艺中U的回收再利用研究%Recycling of Uranium in Process of Preparation UO2 Kernels by Sol-Gel Method

    Institute of Scientific and Technical Information of China (English)

    郝少昌; 赵兴宇; 马景陶; 陈晓彤; 王阳; 邓长生

    2012-01-01

    介绍了清华大学核能与新能源技术研究院制备高温气冷堆燃料元件UO2核芯的溶胶凝胶工艺各种废料的产生情况.针对各种废料的特性,分别进行处理,实现回收再利用.对工艺中产生的含U废液先进行沉淀处理,经1次煅烧后溶解性差,必须将沉淀进行煅烧-还原-煅烧工序后才能实现再利用.对最终烧结球中的不合格品直接进行煅烧处理,将UO2转化为U3 O8,即可再利用.各种废料回收后均可作为原料生产UO2核芯,综合回收率为99.98%.%This report describes the various waste generated in the preparation process of UO2 kernels for HTGR fuel elements by sol-gel method in Institute of Nuclear and New Energy Technology (INET), Tsinghua University. The U-containing liquid waste could be firstly precipitated by a strong alkaline. It is crucial that the precipitate must be treated through calcination-reduction-calcination such that recovered U3 O8 powders can be used effectively; otherwise, if the precipitate was directly calcined to U3 Os,the U3O8 powders would have poor solubility. UO2 microspheres that do not meet the technical specifications and treated as not acceptable in quality could be calcined directly and U3O8 powders thus obtained could be used satisfactorily. The U3O8 recycled from those wastes can be reused as raw materials to produce UO2 kernels meeting the technical requirements. The overall U recovery yield is 99.

  14. UO2中铌行为模拟研究%Simulation on Nb Behavior in UO2

    Institute of Scientific and Technical Information of China (English)

    张永彬; 蒙大桥; 朱正和

    2007-01-01

    通过拟合Nb2O5的晶体结构,建立了铌的经验势.模拟计算孤立铌杂质表明:最近邻Nb-O间距为2.13A,均匀向铌靠近.近邻Nb-U间距为3.84(A),次近邻Nb-U间距为5.48(A),小于正常晶格的U-O间距以及U-U间距.铌附近间隙原子形成能全部增加.铌对周围氧原子施加了额外的束缚,氧间隙形成能随着Nb-O距离的增加而减小,但都大于完整的UO2的氧问隙形成能.引入少量铌元素后其影响范围扩大很多,采用静态过渡态理论计算的铌的空位机制扩散激活能为6.76eV,UO2中铌的扩散几乎不可能.多个铌杂质计算表明:铌离子倾向于团聚,形成替位缺陷簇.根据以上模拟计算,提出如下机制:铌以置换固溶体形式存在于UO2中,为了保持电中性,含铌的UO2中存在相应的氧间隙原子,氧化层总体上保持CaF2结构.铌形成二聚体或者四聚体,联结成网络,使氧的扩散势垒增大,降低了氧化速率.

  15. Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carlo Parisi; Emanuele Negrenti

    2017-02-01

    In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

  16. Combined effects of Fe(II) and oxidizing radiolysis products on UO2 and PuO2 dissolution in a system containing solid UO2 and PuO2

    Science.gov (United States)

    Amme, Marcus; Pehrman, Reijo; Deutsch, Rudolf; Roth, Olivia; Jonsson, Mats

    2012-11-01

    The stability of UO2 spent nuclear fuel in an oxygen-free geological repository depends on the absence of oxidizing reaction partners in the near field. This work investigates the reactions between the products of water radiolysis by alpha radiation and Fe(II) an the effect on UO2 dissolution. Solid 238PuO2 powder and UO2 pellet were allowed to react in Fe(II) solution in oxygen-free batch reactor tests and kinetics of the subsequent redox reactions were measured. Depending on the concentration of Fe(II) (tests with 10-5 and 10-4 mol L-1 were made), the induced redox reactions took place between 20 and 400 h. Dissolved uranium concentrations went first through a minimum caused by reduction, followed by a maximum caused by radiolytic oxidation, and eventually reached another minimum, probably due to sorption on precipitated Fe(III). Plutonium concentrations were decreasing steadily after going through a maximum about 70 h from the start of the experiments. The results show that in the presence of the strong alpha-radiolytic field induced by the presence of solid 238Pu, the behavior of the system is largely governed by Fe(II) as it controls the H2O2 concentration, reduces U(VI) in solution and drives the Fenton reaction leading to the oxidation of Pu(IV).

  17. Structural effects in UO2 thin films irradiated with fission-energy Xe ions

    Science.gov (United States)

    Popel, A. J.; Lebedev, V. A.; Martin, P. G.; Shiryaev, A. A.; Lampronti, G. I.; Springell, R.; Kalmykov, S. N.; Scott, T. B.; Monnet, I.; Grygiel, C.; Farnan, I.

    2016-12-01

    Uranium dioxide thin films have been successfully grown on LSAT (Al10La3O51Sr14Ta7) substrates by reactive magnetron sputtering. Irradiation by 92 MeV 129Xe23+ ions to simulate fission damage that occurs within nuclear fuels caused microstructural and crystallographic changes. Initially flat and continuous thin films were produced by magnetron sputtering with a root mean square roughness of 0.35 nm determined by AFM. After irradiation, this roughness increased to 60-70 nm, with the films developing discrete microstructural features: small grains (∼3 μm), along with larger circular (up to 40 μm) and linear formations with non-uniform composition according to the SEM, AFM and EDX results. The irradiation caused significant restructuring of the UO2 films that was manifested in significant film-substrate mixing, observed through EDX analysis. Diffusion of Al from the substrate into the film in unirradiated samples was also observed.

  18. Formation of studtite during the oxidative dissolution of UO2 by hydrogen peroxide: a SFM study.

    Science.gov (United States)

    Clarens, F; de Pablo, J; Díez-Pérez, I; Casas, I; Giménez, J; Rovira, M

    2004-12-15

    Understanding the formation of alteration phases on the surface of spent nuclear fuel, such as those observed during leaching experiments, is necessary in order to predict the concentration of radionuclides in the near-field of a final repository. Hydrogen peroxide has been identified as one of the oxidants formed by the radiolysis of water in the presence of spent nuclear fuel; especially due to alpha activity. The presence of this species in solution can contribute to the formation of uranium peroxide secondary phases. In this work, we have studied the oxidative dissolution of synthetic UO2 disks in hydrogen peroxide solutions of two different concentrations (5 x 10(-4) and 5 x 10(-6) mol dm(-3)), both at pH 5.8 +/- 0.1. The solid surface evolution of the disks has been followed by means of ex-situ scanning force microscope (SFM) measurements, and uranium concentration in solution has been determined by inductively coupled plasma mass spectrometry. During the first stage of the experiment, SFM images indicate that only UO2 dissolution is occurring. After 142 h, a secondary phase is observed on the surface of the solid at 5 x 10(-4) mol dm(-3) hydrogen peroxide concentration. This secondary phase has been identified by X-ray diffraction as studtite (UO4 x 4H2O). From the analysis of SFM topographic profiles at different elapsed times, a precipitation rate for the studtite has been estimated to be in the range of (8-32) x 10(-10) mol m(-2) s(-1).

  19. Atomistic modelling of residual stress at UO2 surfaces.

    Science.gov (United States)

    Arayro, Jack; Tréglia, Guy; Ribeiro, Fabienne

    2016-01-13

    Modelling oxide surface behaviour is of both technological and fundamental interest. In particular, in the case of the UO2 system, which is of major importance in the nuclear industry, it is essential to account for the link between microstructure and macroscopic mechanical properties. Indeed micromechanical models at the mesoscale need to be supplied by the energetic and stress data calculated at the nanoscale. In this framework, we present a theoretical study, coupling an analytical model and thermostatistical simulation to investigate the modifications induced by the presence of a surface regarding atomic relaxation and energetic and stress profiles. In particular, we show that the surface effective thickness as well as the stress profile, which are required by micromechanical approaches, are strongly anisotropic.

  20. Low temperature synthesis and sintering of d-UO2 nanoparticles.

    Energy Technology Data Exchange (ETDEWEB)

    Nenoff, Tina Maria; Ferreira, Summer Rhodes; Robinson, David B. (Sandia National Laboratories, Livermore CA); Jacobs, Benjamin W. (Sandia National Laboratories, Livermore CA); Provencio, Paula Polyak; Huang, Jian Yu

    2010-12-01

    We report on the novel room temperature method of synthesizing advanced nuclear fuels; a method that virtually eliminates any volatility of components. This process uses radiolysis to form stable nanoparticle (NP) nuclear transuranic (TRU) fuel surrogates and in-situ heated stage TEM to sinter the NPs. The radiolysis is performed at Sandia's Gamma Irradiation Facility (GIF) 60Co source (3 x 10{sup 6} rad/hr). Using this method, sufficient quantities of fuels for research purposes can be produced for accelerated advanced nuclear fuel development. We are focused on both metallic and oxide alloy nanoparticles of varying compositions, in particular d-U, d-U/La alloys and d-UO2 NPs. We present detailed descriptions of the synthesis procedures, the characterization of the NPs, the sintering of the NPs, and their stability with temperature. We have employed UV-vis, HRTEM, HAADF-STEM imaging, single particle EDX and EFTEM mapping characterization techniques to confirm the composition and alloying of these NPs.

  1. Science and Technology of Reactor——Brief Introduction to the Research Program of In-pile Irradiation Test for Advanced Process UO2 Pellets

    Institute of Scientific and Technical Information of China (English)

    ZHANGPei-sheng; WANGHua-rong

    2003-01-01

    In order to develop advanced PWR fuel assembly it is of great importance to carry out in-pile irradiation test UO2 PWR pellets manufactured with advanced process.A research program of the in-pile irradiation test has been planned.The main contents of the program are;1)to develop in-pile testing facility cooled directly by primary coolant in research reactor;2)to design thin fuel element.

  2. Active mechanism of low-temperature sintering to UO2+x pellets%UO2+x芯块低温烧结的活化机理

    Institute of Scientific and Technical Information of China (English)

    李锐

    2012-01-01

    报道了UO2+x芯块低温烧结实验的结果.12组芯块在N2+CO2组成的部分氧化气氛下于立式钼丝炉中低温烧结.UO2芯块要获得密度为10.41g/cm3(94.98%理论密度)需在氢气氛中于2073~2273K下烧结,而UO2+x芯块实现该密度的烧结温度可降低400K以上.建立了超氧化铀缺陷模型来研究低温烧结的活化机理.研究发现铀离子扩散系数与气氛中氧分压或是UO2+x中x成正比.利用铀离子的扩散系数,可预测UO2+x芯块在1073、1273、1473和1673K温度下的烧结密度;还可算出x=0.04时,UO2+x芯块在部分氧化气氛下的理论烧成温度.计算所得烧结密度和烧成温度与实验结果符合得很好.%Low-temperature sintering experiments of UO2+x pellets and the results were reported in this paper. 12 groups pellets were sintered in vertical molybdenum wire furnace with N2+CO2 partially-oxidative atmosphere. Densities which achieved 10. 41g/cm3(94. 98% theoretical density) could be sintered at a much lower temperature (more than 400K) while UO2 pellets were sintered at 2073-2273K in hydrogen. A defect model of hyper-stoichiotnetry uranium dioxide was built for studying the active mechanism. The diffusion coefficient was positive to partial pressure of oxygen (Po2) or excess oxygen ions (x in UO2+x). The sintered densities of UO2+x pellets were predicted by diffusion coefficient of uranium ions at 1073, 1273, 1473 and 1673K. The theoretical sintering temperature of UO2+x in partially-oxidative atmosphere was also calculated when x was 0. 04. The results of sintered densities and sintering temperature were agreed to experimental results very well.

  3. Inhibition of radiation induced dissolution of UO2 by sulfide - A comparison with the hydrogen effect

    Science.gov (United States)

    Yang, Miao; Barreiro Fidalgo, Alexandre; Sundin, Sara; Jonsson, Mats

    2013-03-01

    In this work we have studied the influence of H2S on radiation induced dissolution of spent nuclear fuel using simple model systems. The reaction between H2O2 and H2S/HS- has been studied experimentally as well as the effect of H2S/HS- on γ-radiation induced dissolution of a UO2 pellet. The experiments clearly show that the reaction of H2O2 and H2S/HS- is fairly rapid and that H2O2 and H2S/HS- stoichiometry is favorable for inhibition. Radiolysis experiments show that H2S/HS- can effectively protect UO2 from oxidative dissolution. The effect depends on sulfide concentration in combination with dose rate. Autoclave experiments were also conducted to study the role of H2S/HS- in the reduction of U(VI) in the presence and absence of H2 and Pd particles in anoxic aqueous solution. The aqueous solutions were pressurized with H2 or N2 and two different concentrations of H2S/HS- were used in the presence and absence of Pd. No catalytic effect of Pd on the U(VI) reduction by H2S/HS- could be found in N2 atmosphere. U(VI) reduction was found to be proportional to H2S/HS- concentration in H2 and N2 atmosphere. It is clearly shown the Pd catalyzed H2 effect is more powerful than the effect of H2S/HS-. H2S/HS- poisoning of the Pd catalyst is not observed under the present conditions.

  4. Innovative Fresh Water Production Process for Fossil Fuel Plants

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight; Venugopal Jogi

    2005-09-01

    This project concerns a diffusion driven desalination (DDD) process where warm water is evaporated into a low humidity air stream, and the vapor is condensed out to produce distilled water. Although the process has a low fresh water to feed water conversion efficiency, it has been demonstrated that this process can potentially produce low cost distilled water when driven by low grade waste heat. This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. A dynamic analysis of heat and mass transfer demonstrates that the DDD process can yield a fresh water production of 1.03 million gallon/day by utilizing waste heat from a 100 MW steam power plant based on a condensing steam pressure of only 3 Hg. The optimum operating condition for the DDD process with a high temperature of 50 C and sink temperature of 25 C has an air mass flux of 1.5 kg/m{sup 2}-s, air to feed water mass flow ratio of 1 in the diffusion tower, and a fresh water to air mass flow ratio of 2 in the condenser. Operating at these conditions yields a fresh water production efficiency (m{sub fW}/m{sub L}) of 0.031 and electric energy consumption rate of 0.0023 kW-hr/kg{sub fW}. Throughout the past year, the main focus of the desalination process has been on the direct contact condenser. Detailed heat and mass transfer analyses required to size and analyze these heat and mass transfer devices are described. The analyses agree quite well with the current data. Recently, it has been recognized that the fresh water production efficiency can be significantly enhanced with air heating. This type of configuration is well suited for power plants utilizing air-cooled condensers. The experimental DDD facility has been modified with an air heating section, and temperature and humidity data have been collected over a range of flow and thermal conditions. It has been experimentally observed that the fresh water production rate is enhanced when air

  5. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Mohamed Darwish; Diego Acevedo; Jessica Knight

    2003-09-01

    This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system, which is powered by the waste heat from low pressure condensing steam in power plants. The desalination is driven by water vapor saturating dry air flowing through a diffusion tower. Liquid water is condensed out of the air/vapor mixture in a direct contact condenser. A thermodynamic analysis demonstrates that the DDD process can yield a fresh water production efficiency of 4.5% based on a feed water inlet temperature of only 50 C. An example is discussed in which the DDD process utilizes waste heat from a 100 MW steam power plant to produce 1.51 million gallons of fresh water per day. The main focus of the initial development of the desalination process has been on the diffusion tower. A detailed mathematical model for the diffusion tower has been described, and its numerical implementation has been used to characterize its performance and provide guidance for design. The analysis has been used to design a laboratory scale diffusion tower, which has been thoroughly instrumented to allow detailed measurements of heat and mass transfer coefficient, as well as fresh water production efficiency. The experimental facility has been described in detail.

  6. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Mohamed Darwish; Diego Acevedo; Jessica Knight

    2003-09-01

    This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system, which is powered by the waste heat from low pressure condensing steam in power plants. The desalination is driven by water vapor saturating dry air flowing through a diffusion tower. Liquid water is condensed out of the air/vapor mixture in a direct contact condenser. A thermodynamic analysis demonstrates that the DDD process can yield a fresh water production efficiency of 4.5% based on a feed water inlet temperature of only 50 C. An example is discussed in which the DDD process utilizes waste heat from a 100 MW steam power plant to produce 1.51 million gallons of fresh water per day. The main focus of the initial development of the desalination process has been on the diffusion tower. A detailed mathematical model for the diffusion tower has been described, and its numerical implementation has been used to characterize its performance and provide guidance for design. The analysis has been used to design a laboratory scale diffusion tower, which has been thoroughly instrumented to allow detailed measurements of heat and mass transfer coefficient, as well as fresh water production efficiency. The experimental facility has been described in detail.

  7. Cation-Cation Interactions in [(UO2)2(OH)n](4-n) Complexes

    Energy Technology Data Exchange (ETDEWEB)

    Odoh, Samuel O.; Govind, Niranjan; Schreckenbach, Georg; De Jong, Wibe A.

    2013-10-07

    The structures and bonding of gas-phase [(UO2)2(OH)n]4-n (n=2-6) complexes have been studied using density functional theory (DFT), MP2 and CCSD(T) methods with particular emphasis on ground state structures featuring cation-cation interactions (CCIs) between the uranyl groups. An interesting trend is observed in the stabilities of members of this series of complexes. The structures of [(UO2)2(OH)2]2+, [(UO2)2(OH)4] and [(UO2)2(OH)6]2- featuring CCIs are found at higher energies (by 3-20 kcal/mol) in comparison to their conventional μ2-dihydroxo structures. In contrast, the CCI structures of [(UO2)2(OH)3]+ and [(UO2)2(OH)5]- are respectively almost degenerate with and lower in energy than the structures with the μ2-dihydroxo format. The origin of this trend lies in the ‘symmetry’-based need to balance the coordination numbers and effective atomic charges of each uranium center. The calculated IR vibrational frequencies provide signature probes that can be used in differentiating the lowenergy structures and in experimentally confirming the existence of the structures featuring CCIs. Analysis of the bonding in the structures of [(UO2)2(OH)3]+ and [(UO2)2(OH)5]- shows that the CCIs and bridging hydroxo between the dioxo-uranium units are mainly electrostatic in nature.

  8. Energy transfer and biexciton decay in Cs2UO2Cl4 crystals

    NARCIS (Netherlands)

    Krol, D.M.

    1980-01-01

    We have investigated the influence of energy transfer on luminescence properties of Cs2UO2Cl4 crystals at low temperatures. Time-resolved emission spectra and luminescence decay times were measured between 1.5 and 15 K with the use of selective excitation techniques. The luminescence of Cs2UO2Cl4

  9. Energy transfer and biexciton decay in Cs2UO2Cl4 crystals

    NARCIS (Netherlands)

    Krol, D.M.

    1980-01-01

    We have investigated the influence of energy transfer on luminescence properties of Cs2UO2Cl4 crystals at low temperatures. Time-resolved emission spectra and luminescence decay times were measured between 1.5 and 15 K with the use of selective excitation techniques. The luminescence of Cs2UO2Cl4 de

  10. 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

    2017-04-01

    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.

  11. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight

    2004-09-01

    An innovative Diffusion Driven Desalination (DDD) process was recently described where evaporation of mineralized water is driven by diffusion within a packed bed. The energy source to drive the process is derived from low pressure condensing steam within the main condenser of a steam power generating plant. Since waste heat is used to drive the process, the main cost of fresh water production is attributed to the energy cost of pumping air and water through the packed bed. This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. A combined thermodynamic and dynamic analysis demonstrates that the DDD process can yield a fresh water production of 1.03 million gallon/day by utilizing waste heat from a 100 MW steam power plant based on a condensing steam pressure of only 3'' Hg. Throughout the past year, the main focus of the desalination process has been on the diffusion tower and direct contact condenser. Detailed heat and mass transfer analyses required to size and analyze these heat and mass transfer devices are described. An experimental DDD facility has been fabricated, and temperature and humidity data have been collected over a range of flow and thermal conditions. The analyses agree quite well with the current data and the information available in the literature. Direct contact condensers with and without packing have been investigated. It has been experimentally observed that the fresh water production rate is significantly enhanced when packing is added to the direct contact condensers.

  12. 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.

  13. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  14. Alternate Anodes for the Electrolytic Reduction of UO2

    Science.gov (United States)

    Merwin, Augustus; Chidambaram, Dev

    2015-01-01

    The electrolytic reduction process of UO2 employs a platinum anode and a stainless steel cathode in molten LiCl-LiO2 maintained at 973 K (700 °C). The degradation of platinum under the severely oxidizing conditions encountered during the process is an issue of concern. In this study, Inconel 600 and 718, stainless steel alloy 316, tungsten, nickel, molybdenum, and titanium, were investigated though electrochemical polarization techniques, electron microscopy, Raman spectroscopy, and X-ray photoelectron spectroscopy to serve as potential anode materials. Of the various materials investigated, only tungsten exhibited sufficient stability at the required potential in the molten electrolyte. Tungsten anodes were further studied in molten LiCl-LiO2 electrolyte containing 2, 4, and 6 wt pct of Li2O. In LiCl-2 wt pct Li2O tungsten was found to be sufficiently stable to both oxidation and microstructural changes and the stability is attributed to the formation of a lithium-intercalated tungsten oxide surface film. Increase in the concentration of Li2O was found to lead to accelerated corrosion of the anode, in conjunction with the formation of a peroxotungstate oxide film.

  15. Innovative Fresh Water Production Process for Fossil Fuel Plants

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight

    2006-09-29

    This project concerns a diffusion driven desalination (DDD) process where warm water is evaporated into a low humidity air stream, and the vapor is condensed out to produce distilled water. Although the process has a low fresh water to feed water conversion efficiency, it has been demonstrated that this process can potentially produce low cost distilled water when driven by low grade waste heat. This report summarizes the progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. Detailed heat and mass transfer analyses required to size and analyze the diffusion tower using a heated water input are described. The analyses agree quite well with the current data and the information available in the literature. The direct contact condenser has also been thoroughly analyzed and the system performance at optimal operating conditions has been considered using a heated water/ambient air input to the diffusion tower. The diffusion tower has also been analyzed using a heated air input. The DDD laboratory facility has successfully been modified to include an air heating section. Experiments have been conducted over a range of parameters for two different cases: heated air/heated water and heated air/ambient water. A theoretical heat and mass transfer model has been examined for both of these cases and agreement between the experimental and theoretical data is good. A parametric study reveals that for every liquid mass flux there is an air mass flux value where the diffusion tower energy consumption is minimal and an air mass flux where the fresh water production flux is maximized. A study was also performed to compare the DDD process with different inlet operating conditions as well as different packing. It is shown that the heated air/heated water case is more capable of greater fresh water production with the same energy consumption than the ambient air/heated water process at high liquid mass flux. It is also shown that there can be

  16. Modeling conversion of ammonium diuranate (ADU) into uranium dioxide (UO2) powder

    Science.gov (United States)

    Hung, Nguyen Trong; Thuan, Le Ba; Khoai, Do Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

    2016-10-01

    In the paper, Brandon mathematical model that describes the relationship between the essential fabrication parameters [reduction temperature (TR), calcination temperature (TC), calcination time (tC) and reduction time (tR)] and specific surface area of ammonium diuranate (ADU)-derived UO2 powder products was established. The proposed models can be used to predict and control the specific surface area of UO2 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 UO2 powders.

  17. Vacuum extraction based response equipment for recovery of fresh fuel spills from soil.

    Science.gov (United States)

    Halmemies, Sakari; Gröndahl, Siri; Arffman, Mika; Nenonen, Keijo; Tuhkanen, Tuula

    2003-02-28

    Accidental overturns of fuel tankers can have, depending on soil types, severe consequences. This applies, particularly in areas of shallow soils where the groundwater is located 2-4m below the ground surface. By rapid, vacuum extraction based recovery emergency services, which would normally be the first to arrive on the scene, could minimize consequences of fresh fuel spills and even prevent groundwater contamination, the primary purpose of emergency response. Powerful vacuum extraction-based response (PER), equipment has been developed to recover freshly spilt volatile fuels from the soil, primary by emergency services, but also by other trained responders. The main components of mobile PER-equipment are perforated extraction pipes, a recovery vacuum tank, a vacuum pump and an incinerator. The PER-equipment has been tested in summer and sub-zero winter conditions, and in both cases 50-80% of fresh gasoline spilled into sandy soil was recovered during the first 2h of operation. Gasoline was recovered in both liquid and vapor form, and hydrocarbon vapors were destroyed by controlled incineration at a safe distance from the spill. Recovery of less volatile diesel oil is not so effective from the sandy soil, but about 30% of it could be pumped from a fresh pool directly after a seepage time of 15 min.

  18. Biogenic nanoparticulate UO 2: Synthesis, characterization, and factors affecting surface reactivity

    Science.gov (United States)

    Singer, David M.; Farges, François; Brown, Gordon E., Jr.

    2009-06-01

    The surface reactivity of biogenic, nanoparticulate UO 2 with respect to sorption of aqueous Zn(II) and particle annealing is different from that of bulk uraninite because of the presence of surface-associated organic matter on the biogenic UO 2. Synthesis of biogenic UO 2 was accomplished by reduction of aqueous uranyl ions, UO22+ by Shewanella putrefaciens CN32, and the resulting nanoparticles were washed using one of two protocols: (1) to remove surface-associated organic matter and soluble uranyl species (NAUO2), or (2) to remove only soluble uranyl species (BIUO2). A suite of bulk and surface characterization techniques was used to examine bulk and biogenic, nanoparticulate UO 2 as a function of particle size and surface-associated organic matter. The N 2-BET surface areas of the two biogenic UO 2 samples following the washing procedures are 128.63 m 2 g -1 (NAUO2) and 92.56 m 2 g -1 (BIUO2), and the average particle sizes range from 5-10 nm based on TEM imaging. Electrophoretic mobility measurements indicate that the surface charge behavior of biogenic, nanoparticulate UO 2 (both NAUO2 and BIUO2) over the pH range 3-9 is the same as that of bulk. The U L III-edge EXAFS spectra for biogenic UO 2 (both NAUO2 and BIUO2) were best fit with half the number of second-shell uranium neighbors compared to bulk uraninite, and no oxygen neighbors were detected beyond the first shell around U(IV) in the biogenic UO 2. At pH 7, sorption of Zn(II) onto both bulk uraninite and biogenic, nanoparticulate UO 2 is independent of electrolyte concentration, suggesting that Zn(II) sorption complexes are dominantly inner-sphere. The maximum surface area-normalized Zn(II) sorption loadings for the three substrates were 3.00 ± 0.20 μmol m -2 UO 2 (bulk uraninite), 2.34 ± 0.12 μmol m -2 UO 2 (NAUO2), and 2.57 ± 0.10 μmol m -2 UO 2 (BIUO2). Fits of Zn K-edge EXAFS spectra for biogenic, nanoparticulate UO 2 indicate that Zn(II) sorption is dependent on the washing protocol. Zn

  19. 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.

  20. Time-resolved laser fluorescence spectroscopy of UO2(CO3)3(4-).

    Science.gov (United States)

    Jung, E C; Cho, H-R; Baik, M H; Kim, H; Cha, W

    2015-11-21

    The objective of the present study is to examine the luminescence characteristics of UO2(CO3)3(4-) in detail using time-resolved laser fluorescence spectroscopy. The peak wavelengths and lifetime of UO2(CO3)3(4-) were determined at room temperature using the two excitation laser wavelengths of 266 and 448 nm. The peak wavelengths in the luminescence spectrum exhibited hypsochromic shifts compared with those of UO2(2+). The lifetime determined from several samples containing various uranium concentrations was 8.9 ± 0.8 ns. Explanations for the hindrance to the observation of the luminescence spectrum of UO2(CO3)3(4-) in previous investigations are discussed. The representative experimental parameters, which might interrupt the measurement of weak luminescence, are the insertion delay time of the detection device, the overlapped luminescence of the background materials and the primary inner filter effect in the sample solution.

  1. Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, S. B.; Michelhaugh, R. D.; Pope, R. B.; Shappert, L. B.; Singletary, B. H.; Chae, S. M.; Parks, C. V.; Broadhead, B. L.; Schmid, S. P.; Cowart, C. G.

    1997-12-01

    This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies.

  2. Hydrothermal growth and characterization of UO2 single crystals for neutron radiation detection(Conference Presentation)

    Science.gov (United States)

    Mann, Matthew; Hunt, Eric; Young, Christopher; Kimani, Martin; Turner, David; Varga, Stephan; Petrosky, James

    2016-09-01

    There is significant interest in developing efficient, direct conversion, neutron sensitive solid-state radiation detector materials with the ability to discriminate between photon and neutron events. Recently, this has led several research groups to pursue uranium dioxide (UO2) single crystals as a detection material due to the large reaction energy ( 185 MeV) from a neutron induced fission event. The resulting electrical pulse, generated primarily by the energetic fission fragments, is expected to be on the order of 165 MeV, which is much greater than current detection schemes which rely on reaction energies between 2-6 MeV. The primary technical challenge to the successful fabrication of UO2 devices is the lack of high quality (semiconductor grade) single crystals of UO2. The high melting point of UO2 ( 2878°C) precludes the use of traditional melt growth techniques like Czochralski. While exotic melt growth techniques such as arc fusion, cold crucible, and solar furnace have successfully grown UO2, the crystal quality suffers from both thermal strain and oxygen non-stoichiometry, two particularly difficult challenges inherent to uranium oxide materials. Crystal growth of UO2 by the hydrothermal synthesis technique has never been investigated, although the method has been successfully applied to the synthesis of other refractory oxides. In this talk, we will present growth of UO2 single crystals from a variety of hydrothermal solutions at temperatures below 650C. X-ray diffraction confirmed the stoichiometric nature of the samples and X-ray photoelectron spectroscopy determined the photoelectric work function of two crystal orientations. Preliminary proof-of-concept irradiation studies of a simple UO2 resistive detector will also be presented.

  3. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  4. Xenon migration in UO2 under irradiation studied by SIMS profilometry

    Science.gov (United States)

    Marchand, B.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Garnier, C.; Raimbault, L.; Sainsot, P.; Epicier, T.; Delafoy, C.; Fraczkiewicz, M.; Gaillard, C.; Toulhoat, N.; Perrat-Mabilon, A.; Peaucelle, C.

    2013-09-01

    During Pressurized Water Reactor operation, around 25% of the created Fission Products (FP) are Xenon and Krypton. They have a low solubility in the nuclear fuel and can either (i) agglomerate into bubbles which induce mechanical stress in the fuel pellets or (ii) be released from the pellets, increasing the pressure within the cladding and decreasing the thermal conductivity of the gap between pellets and cladding. After fifty years of studies on the nuclear fuel, all mechanisms of Fission Gas Release (FGR) are still not fully understood. This paper aims at studying the FGR mechanisms by decoupling thermal and irradiation effects and by assessing the Xenon behavior for the first time by profilometry. Samples are first implanted with 136Xe at 800 keV corresponding to a projected range of 140 nm. They are then either annealed in the temperature range 1400-1600 °C, or irradiated with heavy energy ions (182 MeV Iodine) at Room Temperature (RT), 600 °C or 1000 °C. Depth profiles of implanted Xenon in UO2 are determined by Secondary Ion Mass Spectrometry (SIMS). It is shown that Xenon is mobile during irradiation at 1000 °C. In contrast, thermal treatments do not induce any Xenon migration process: these results are correlated to the formation of Xenon bubbles observed by Transmission Electron Microscopy. At depths lower than about 40 nm (zone 1), no bubbles are observed, At depths in between 40 nm and 110 nm (zone 2), a large number of small bubbles (around 2 nm in diameter) can be observed. By comparing with the SRIM profile, it appears that this area corresponds to the maximum of the defect profile, The third zone displays two bubble populations. The first population has the same size than the bubbles present in zone 2. The bubble size of the second population is significantly larger (up to around 10 nm). A STEM micrograph is presented in Fig. 4. It highlights the Xenon bubbles more clearly. It appears that the largest bubbles are located mainly near dislocations

  5. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  6. An X-ray photoelectron spectroscopy study of the products of the interaction of gaseous IrF6 with fine UO2F2

    Directory of Open Access Journals (Sweden)

    Prusakov Vladimir N.

    2007-01-01

    Full Text Available Nuclear fuel reprocessing by fluorination, a dry method of regeneration of spent nuclear fuel, uses UO2F2 for the separation of plutonium from gaseous mixtures. Since plutonium requires special treatment, IrF6 was used as a thermodynamic model of PuF6. The model reaction of the interaction of gaseous IrF6 with fine UO2F2 in the sorption column revealed a change of color of the sorption column contents from pale-yellow to gray and black, indicating the formation of products of such an interaction. The X-ray photoelectron spectroscopy study showed that the interaction of gaseous IrF6 with fine UO2F2 at 125 °C results in the formation of stable iridium compounds where the iridium oxidation state is close to Ir3+. The dependence of the elemental compositions of the layers in the sorption column on the penetration depth of IrF6 was established.

  7. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  8. Valence XPS structure and chemical bond in Cs2UO2Cl4

    Directory of Open Access Journals (Sweden)

    Teterin Yury A.

    2016-01-01

    Full Text Available Quantitative analysis was done of the valence electrons X-ray photoelectron spectra structure in the binding energy (BE range of 0 eV to ~35 eV for crystalline dicaesium tetrachloro-dioxouranium (VI (Cs2UO2Cl4. This compound contains the uranyl group UO2. The BE and structure of the core electronic shells (~35 eV-1250 eV, as well as the relativistic discrete variation calculation results for the UO2Cl4(D4h cluster reflecting U close environment in Cs2UO2Cl4 were taken into account. The experimental data show that many-body effects due to the presence of cesium and chlorine contribute to the outer valence (0-~15 eV BE spectral structure much less than to the inner valence (~15 eV-~35 eV BE one. The filled U5f electronic states were theoretically calculated and experimentally confirmed to be present in the valence band of Cs2UO2Cl4. It corroborates the suggestion on the direct participation of the U5f electrons in the chemical bond. Electrons of the U6p atomic orbitals participate in formation of both the inner (IVMO and the outer (OVMO valence molecular orbitals (bands. The filled U6p and the O2s, Cl3s electronic shells were found to make the largest contributions to the IVMO formation. The molecular orbitals composition and the sequence order in the binding energy range 0 eV-~35 eV in the UO2Cl4 cluster were established. The experimental and theoretical data allowed a quantitative molecular orbitals scheme for the UO2Cl4 cluster in the BE range 0-~35 eV, which is fundamental for both understanding the chemical bond nature in Cs2UO2Cl4 and the interpretation of other X-ray spectra of Cs2UO2Cl4. The contributions to the chemical binding for the UO2Cl4 cluster were evaluated to be: the OVMO contribution - 76%, and the IVMO contribution - 24 %.

  9. Precisely Determining Ultralow level UO2(2+) in Natural Water with Plasmonic Nanowire Interstice Sensor.

    Science.gov (United States)

    Gwak, Raekeun; Kim, Hongki; Yoo, Seung Min; Lee, Sang Yup; Lee, Gyoung-Ja; Lee, Min-Ku; Rhee, Chang-Kyu; Kang, Taejoon; Kim, Bongsoo

    2016-01-21

    Uranium is an essential raw material in nuclear energy generation; however, its use raises concerns about the possibility of severe damage to human health and the natural environment. In this work, we report an ultrasensitive uranyl ion (UO2(2+)) detection method in natural water that uses a plasmonic nanowire interstice (PNI) sensor combined with a DNAzyme-cleaved reaction. UO2(2+) induces the cleavage of DNAzymes into enzyme strands and released strands, which include Raman-active molecules. A PNI sensor can capture the released strands, providing strong surface-enhanced Raman scattering signal. The combination of a PNI sensor and a DNAzyme-cleaved reaction significantly improves the UO2(2+) detection performance, resulting in a detection limit of 1 pM and high selectivity. More importantly, the PNI sensor operates perfectly, even in UO2(2+)-contaminated natural water samples. This suggests the potential usefulness of a PNI sensor in practical UO2(2+)-sensing applications. We anticipate that diverse toxic metal ions can be detected by applying various ion-specific DNA-based ligands to PNI sensors.

  10. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Charlton, William S., E-mail: wcharlton@tamu.edu [Nuclear Security Science and Policy Institute, Texas A and M University, 3473 TAMU, College Station, TX 77843 (United States); Menlove, Howard O., E-mail: hmenlove@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T., E-mail: swinhoe@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States)

    2012-07-11

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four {sup 235}U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from {sup 235}U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15 Multiplication-Sign 15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective {sup 235}U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies. - Highlights: Black-Right-Pointing-Pointer Development of new measurement technique called SINRD to improve LWR safeguards. Black-Right-Pointing-Pointer Performed SINRD experiment to measure {sup 235}U and pin diversions in PWR fresh assembly. Black-Right-Pointing-Pointer Excellent agreement of MCNPX and measured results confirmed accuracy of SINRD model. Black-Right-Pointing-Pointer SINRD

  11. Assessing ligand selectivity for uranium over vanadium ions to aid in the discovery of superior adsorbents for extraction of UO2(2+) from seawater.

    Science.gov (United States)

    Ivanov, Alexander S; Bryantsev, Vyacheslav S

    2016-06-28

    Uranium is used as the basic fuel for nuclear power plants, which generate significant amounts of electricity and have life cycle carbon emissions that are as low as renewable energy sources. However, the extraction of this valuable energy commodity from the ground remains controversial, mainly because of environmental and health impacts. Alternatively, seawater offers an enormous uranium resource that may be tapped at minimal environmental cost. Nowadays, amidoxime polymers are the most widely utilized sorbent materials for large-scale extraction of uranium from seawater, but they are not perfectly selective for uranyl, UO2(2+). In particular, the competition between UO2(2+) and VO(2+)/VO2(+) cations poses a significant challenge to the efficient mining of UO2(2+). Thus, screening and rational design of more selective ligands must be accomplished. One of the key components in achieving this goal is the establishment of computational techniques capable of assessing ligand selectivity trends. Here, we report an approach based on quantum chemical calculations that achieves high accuracy in reproducing experimental aqueous stability constants for VO(2+)/VO2(+) complexes with ten different oxygen donor ligands. The predictive power of the developed computational protocol is demonstrated for amidoxime-type ligands, providing greater insights into new design strategies for the development of the next generation of adsorbents with high selectivity toward UO2(2+) over VO(2+)/VO2(+) ions. Importantly, the results of calculations suggest that alkylation of amidoxime moieties present in poly(acrylamidoxime) sorbents can be a potential route to better discrimination between the uranyl and competing vanadium ions in seawater.

  12. Effect of the substitution of f-electron elements on the structure and elastic properties of UO2

    Science.gov (United States)

    Behera, Rakesh K.; Deo, Chaitanya S.; Xu, Haixuan

    2013-02-01

    The chemistry of nuclear reactor fuel initially is complex, and continuous loss of uranium and plutonium, and formation of a broad range of new species due to fission introduce a challenging time-dependence to this chemistry. Lanthanides and/or Actinides substitution on the uranium sublattice occurs (a) during fission, (b) when mixed oxide fuel is used, and (c) when minor Actinides are reprocessed in UO2 matrix fuel as part of a closed nuclear fuel cycle. These fission products and minor Actinides influence a variety of thermo-physical properties, which depend on structure and elastic properties. How these structural and elastic properties vary with Lanthanide and Actinide substitution is not well studied. In this study we use atomic level simulations to investigate the effect of 4+ and 3+ ion substitutions on the structural and elastic properties of urania matrix. Our results show that most of the 4+ ions reduce the overall lattice parameter, while all the 3+ ions considered here increased the lattice parameter of the urania matrix. This effect is guided by the interplay between the elastic and electrostatic effect of the substituted ions. We calculate the chemical expansion and chemical expansion coefficient with the change in concentration based on the ionic radii of the substituted 3+ and 4+ ions. In general, elastic properties are enhanced for 4+ ions substitution and reduced for 3+ ion substitution.

  13. Modeling of Fission Gas Release in UO2

    Energy Technology Data Exchange (ETDEWEB)

    MH Krohn

    2006-01-23

    A two-stage gas release model was examined to determine if it could provide a physically realistic and accurate model for fission gas release under Prometheus conditions. The single-stage Booth model [1], which is often used to calculate fission gas release, is considered to be oversimplified and not representative of the mechanisms that occur during fission gas release. Two-stage gas release models require saturation at the grain boundaries before gas is release, leading to a time delay in release of gases generated in the fuel. Two versions of a two-stage model developed by Forsberg and Massih [2] were implemented using Mathcad [3]. The original Forsbers and Massih model [2] and a modified version of the Forsberg and Massih model that is used in a commercially available fuel performance code (FRAPCON-3) [4] were examined. After an examination of these models, it is apparent that without further development and validation neither of these models should be used to calculate fission gas release under Prometheus-type conditions. There is too much uncertainty in the input parameters used in the models. In addition. the data used to tune the modified Forsberg and Massih model (FRAPCON-3) was collected under commercial reactor conditions, which will have higher fission rates relative to Prometheus conditions [4].

  14. Thermodynamics of fission products in UO2+-x

    Energy Technology Data Exchange (ETDEWEB)

    Nerikar, Pankaj V [Los Alamos National Laboratory

    2009-01-01

    The stabilities of selected fission products - Xe, Cs, and Sr - are investigated as a function of non-stoichiometry x in UO{sub 2{+-}x}. In particular, density functional theory (OFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. In order to reproduce the correct insulating state of UO{sub 2}, the DFT calculations are performed using spin polarization and with the Hubbard U tenn. In general, higher charge defects are more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. CS{sub 2}O is observed as a second stable phase and SrO is found to be soluble in the UO{sub 2} matrix for all stoichiometries. These observations mirror experimentally observed phenomena.

  15. Neutron Flux Depression in the UO{sub 2}-PuO{sub 2}(15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment; Depresion de flujo neutronico en las barras combustibles de UO2-PuO2(15 al 30%) del experimento de irradiacion IVO-FR2-Vg7

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, J.; Fernandez, J. L.

    1983-07-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO{sub 2}-PUO{sub 2} (15 to 30% PUO{sub 2}) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs.

  16. Steady-state and time-dependent luminescence of UO2MoO4

    NARCIS (Netherlands)

    Smit, W.M.A.

    1998-01-01

    Luminescence spectra of UO2MoO4 between 4.2 – 50 K were studied. The 0 – 0 line splitting is explained. Decay functions were derived reproducing the temperature dependence of the trap emissions. The rate of energy transfer among intrinsic centres and to trap centres is shown to be of similar magnitu

  17. Steady-state and time-dependent luminescence of UO2MoO4

    NARCIS (Netherlands)

    Smit, W.M.A.

    1998-01-01

    Luminescence spectra of UO2MoO4 between 4.2 – 50 K were studied. The 0 – 0 line splitting is explained. Decay functions were derived reproducing the temperature dependence of the trap emissions. The rate of energy transfer among intrinsic centres and to trap centres is shown to be of similar

  18. Steady-state and time-dependent luminescence of UO2MoO4

    OpenAIRE

    Smit, W. M. A.

    1998-01-01

    Luminescence spectra of UO2MoO4 between 4.2 – 50 K were studied. The 0 – 0 line splitting is explained. Decay functions were derived reproducing the temperature dependence of the trap emissions. The rate of energy transfer among intrinsic centres and to trap centres is shown to be of similar magnitude.

  19. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    Science.gov (United States)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-01

    The possibility of using UO2-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO2-BeO fuel pellets are estimated.

  20. Classical molecular dynamics investigation of microstructure evolution and grain boundary diffusion in nano-polycrystalline UO2

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2013-07-01

    The High Burnup Structure (HBS) observed at pellet periphery in conventional Light Water Reactor nuclear fuels and around spots presenting high plutonium content in mixed (U, Pu) oxide fuel - MOX fuel - consists of a restructuration of the original grains into smaller ones. The process is often postulated to occur because of the accumulation of irradiation damage and the retention of fission products in the matrix. The computing power nowadays available enables for simulating larger systems at the atomic scale up to the point that nano-polycrystalline material can now be investigated by empirical potential molecular dynamics. Simulations of nano-polycrystalline UO2 structures have been carried out at various temperatures to investigate atom mobility close to grain boundaries. The variation of Arrhénius parameters for the diffusion coefficient of oxygen, uranium and xenon as a function of the distance from a grain boundary was studied, leading to the distinction of three zones: the grain boundary layers (up to 1 nm depth) presenting enhanced diffusion, an intermediate zone (1 to roughly 2 nm depth) with intermediate diffusion values and the bulk of the grains. The following Arrhénius relations for grain boundary diffusion were derived:

  1. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Favalli, Andrea [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory

    2012-06-19

    Gd{sub 2}O{sub 3} burnable poison on the measurement of fresh pressurized water reactor fuel. To empirically determine the response function over the range of historical and future use we have considered enrichments up to 5 wt% {sup 235}U/{sup tot}U and Gd weight fractions of up to 10 % Gd/UO{sub 2}. Parameterized correction factors are presented.

  2. Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hausladen, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Newby, Jason [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-11-01

    The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency s present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the 235U content of the fuel with a moderated 241Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the statistical accuracy of a

  3. Photo-induced low temperature synthesis of nanocrystalline UO2, ThO2 and mixed UO2-ThO2 oxides

    Science.gov (United States)

    Pavelková, Tereza; Čuba, Václav; Šebesta, Ferdinand

    2013-11-01

    Photochemically induced preparation of nanocrystalline uranium and/or thorium oxides is based on UV radiation induced formation of amorphous solid precursor in aqueous solutions containing uranium and/or thorium nitrate and ammonium formate. Subsequent heat treatment under various atmospheres leads to formation of nanocrystalline UO2, ThO2 or UO2-ThO2 solid solution at minimum temperatures in the interval 300-550 °C. The materials consist of nanoparticles from 3 to 15 nm in diameter and with narrow size distribution. The initial solutions contain soluble salts of respective metals and OH radical scavenger (ammonium formate) satisfying the "CHON principle". Solutions may be used without further adjustment (e.g. saturation by inert gases or adjusting pH). Photo-induced precipitation proceeds at room temperature and does not require strict control of reaction conditions (pH, temperature). Due to negligible amounts of carbon in solid precursor formed from any solution, it is possible to prepare crystalline nanomaterials containing U(IV) oxide from solid precursors directly via heat treatment in Ar + H2 atmosphere without pre-calcination in air. Mild heat treatment (450-550 °C) results in formation of oxides with well-developed nanocrystals. In the case of mixed oxides, high level of interaction of both components was observed, resulting in the formation of solid solution U0.56Th0.44O2 at 300 °C or higher.

  4. Reliable Potential Energy Surfaces for the Reactions of H2O with ThO2, PaO2(+), UO2(2+), and UO2(.).

    Science.gov (United States)

    Vasiliu, Monica; Peterson, Kirk A; Gibson, John K; Dixon, David A

    2015-11-19

    The potential energy surfaces for the reactions of H2O with ThO2, PaO2(+), UO2(2+), and UO2(+) have been calculated at the coupled cluster CCSD(T) level extrapolated to the complete basis set limit with additional corrections including scalar relativistic and spin-orbit. The reactions proceed by the formation of an initial Lewis acid-base adduct (H2O)AnO2(0/+/2+) followed by a proton transfer to generate the dihydroxide AnO(OH)2(0/+/2+). The results are in excellent agreement with mass spectrometry experiments and prior calculations of hydrolysis reactions of the group 4 transition metal dioxides MO2. The differences in the energies of the stationary points on the potential energy surface are explained in terms of the charges on the system and the populations on the metal center. The use of an improved starting point for the coupled cluster CCSD(T) calculations based on density functional theory with the PW91 exchange-correlation functional or Brueckner orbitals is described. The importance of including second-order spin-orbit corrections for closed-shell molecules is also described. These improvements in the calculations are correlated with the 5f populations on the actinide.

  5. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  6. UF4水解制备UO2粉末工艺的研究%A STUDY ON THE PREPARATION OF UO2 POWDER BY PYROHYDROLYSIS REACTION OF UF4

    Institute of Scientific and Technical Information of China (English)

    杨光宇; 刘全生; 武爱国; 温国义

    2006-01-01

    本文研究了在H2+H2O混合气流下,UF4水解制备UO2粉末过程中的工艺条件,以及影响UO2粉末物性指标和产品含氟量的主要因素,通过试验确定了UF4水解制备UO2粉末的工艺参数.

  7. A novel gel combustion procedure for the preparation of foam and porous pellets of UO2

    Science.gov (United States)

    Sanjay Kumar, D.; Ananthasivan, K.; Venkata Krishnan, R.; Maji, Dasarath; Dasgupta, Arup

    2017-01-01

    In this study, it has been demonstrated for the first time how sucrose gel-combustion could be used for the preparation of UO2 foam. Further the citrate gel-combustion was gainfully used for preparing porous pellets of UO2. The utility of two-step sintering (1073 K for 30 min and 1473 K for 4 h) for obtaining these porous bodies was demonstrated for the first time. The foams and pellets possessed meso and macro pores. A starting mixture with sucrose to nitrate ratio of 2.4 was found to yield urania foam with adequate crush strength. The porous pellets were found to possess better handling strength, lesser carbon residue and higher overall density than the foam. A citric acid to nitrate ratio 0.25 in the starting mixture, 180 MPa compaction pressure were optimal for obtaining a pellet with 40% porosity.

  8. New insight on the high radiation resistance of UO2 against fission fragments

    Science.gov (United States)

    Szenes, G.

    2016-12-01

    Track radii are derived for semiconductors from a temperature distribution Θ(r) in which the width of the distribution is the only materials parameter. Analysis of track data for GeS, InP, GaAs and GaN show that the projectile velocity has no effect on track radii in semiconductors. Due to the missing velocity effect, the threshold for track formation, Set = 20 keV/nm is high in semiconducting UO2 in the whole range of projectile velocities. This is the origin of the high radiation resistance for fission fragments. Consequences for the simulation experiments with insulating CeO2 are discussed. It is verified that sputtering is described accurately by the Arrhenius equation for various materials including UO2. The ion-induced surface potential has a strong effect on the activation energy.

  9. Low-temperature sintering process for UO2 pellets in partially-oxidative atmosphere

    Institute of Scientific and Technical Information of China (English)

    YANG Xiao-dong; GAO Jia-cheng; WANG Yong; CHANG Xin

    2008-01-01

    Low-temperature sintering(LTS) experiments of UO2 pellets and their results were reported. Moreover, a routine process of LTS for UO2 pellets was primarily established. Being sintered at 1 400 ℃ for 3 h in a partially-oxidative atmosphere, the relative density of the pellet can be up to around 94%. Pellets with such a high density are of benefit for following-up reduction-sintering processes. Orthogonal test indicates that the importance of factors affecting the density decreases in the sequence of partial-oxidative sintering temperature and time, reduction-sintering time and temperature, and sintering atmosphere. It is found that it is helpful to introducing a small amount of water vapor into the sintering atmosphere during the latter stage. It is believed that it is the key factor to raise the O/U ratio of original powder in order to improve the properties of the low-temperature sintered pellets.

  10. Analysis of the Temperature Effect on the Infinite Multiplication Factor for HEU-UAl4 and LEU-UO2 Lattices of GHARR-1

    Directory of Open Access Journals (Sweden)

    E. Alhassan

    2011-03-01

    Full Text Available The purpose of the study is to analyze the temperature effect on the infinite multiplication factor for light water moderated High Enriched Uranium (HEU-UAl4 and Low Enriched Uranium (LEU-UO2 lattices of the Ghana Research Reactor-1 (GHARR-1. To quantify the contribution of each component of the infinite multiplication factor with respect to temperature within the 20 to 140ºC range, cell calculations were performed for the two MNSR typical lattices: the 90.2% enriched HEU-UAl4 and 12.6% enriched LEU- UO2 proposed fuel of the Ghana Research Reactor-1 (GHARR-1 using SCUBA, a locally developed FORTRAN 95 code for the calculations and analysis of temperature coefficients of GHARR-1. It was observed that at the beginning of life of the core, the temperature coefficient of the resonance escape probability and that of the thermal utilization factor, contributed significantly to the negative temperature coefficient of the infinite multiplication factor obtained for both fuels.

  11. New head process for non-HEU 99Mo production based on the oxidation of irradiated UO2-pellets forming soluble U3O8

    Directory of Open Access Journals (Sweden)

    Beyer Gerd Juergen

    2016-01-01

    Full Text Available All fission-based 99Mo producers worldwide are required to convert their 99Mo production processes from using highly enriched uranium to low-enriched uranium. At a recent IAEA meeting in Vienna, problems related to bottlenecks and target modification and optimization of low-enriched uranium-based 99Mo production processes were discussed. Ceramic UO2-pellets (as used in fuel were excluded from the discussion with the argument that this material cannot be dissolved under practically applicable conditions. In this paper, we suggest transforming the non-soluble ceramic UO2 fuel-pellets into the U3O8 form by simple oxidation and the use of the soluble U-oxide modification as the starting material for the 99Mo production processes. Due to the absence of Al, larger target quantities could be processed and the waste volume could still be kept small. The approach is known and proven in nuclear technology. In principle, this new head process can be connected to any of the presently used 99Mo production procedures.

  12. Thermal expansion in UO2 determined by high-energy X-ray diffraction

    Science.gov (United States)

    Guthrie, M.; Benmore, C. J.; Skinner, L. B.; Alderman, O. L. G.; Weber, J. K. R.; Parise, J. B.; Williamson, M.

    2016-10-01

    Here we present crystallographic analyses of high-energy X-ray diffraction data on polycrystalline UO2 up to the melting temperature. The Rietveld refinements of our X-ray data are in agreement with previous measurements, but are systematically located around the upper bound of their uncertainty, indicating a slightly steeper trend of thermal expansion compared to established values. This observation is consistent with recent first principles calculations.

  13. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  14. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  15. Molecular dynamics study of fission gas bubble nucleation in UO2

    Science.gov (United States)

    Liu, X.-Y.; Andersson, D. A.

    2015-07-01

    Molecular dynamics (MD) simulations are used to study helium and xenon gas bubble nucleation in UO2. For helium bubbles, the pressure release mechanism is by creating defects on the oxygen sublattice. Helium atoms diffuse away from the bubbles into nearby bulk UO2, thus forming a diffuse interface. For xenon bubbles, over-pressurized bubbles containing xenon can displace uranium atoms, which tend to aggregate around the xenon bubble as a pressure release mechanism. MD simulations of xenon atoms in pre-existing voids suggest that xenon atoms and the replaced uranium atoms occur in a 1:1 ratio, although kinetic factors may reduce that ratio depending on availability of xenon atoms and vacancies around the bubble. Finally, MD simulations suggest that for small bubbles (1-5 xenon atoms), the xenon bubble nucleus at UO2 grain-boundaries has much lower formation energy compared to that of bubbles of similar sizes in the bulk. However, when the xenon bubble grows into larger sizes, this energy difference is reduced.

  16. Radiolysis and corrosion of 238Pu-doped UO2 pellets in chloride brine

    Indian Academy of Sciences (India)

    M Kelm; E Bohnert

    2002-12-01

    Deaerated 5 M NaCl solution is irradiated in the presence of UO2 pellets with a-radiation from 238Pu. Experiments are conducted with 238Pu doped pellets and others with 238Pu dissolved in the brine. The radiolysis products and yields of mobilized U and Pu from the oxidative dissolution of UO2 are determined. Results found for radiolysis products and for the oxidation/dissolution of pellets immersed in Pu containing brine are similar to results for Pu doped pellets, where the radiation chemical processes occur only in the liquid layer of some 10 m thickness adjacent to the pellet. The yield of radiolysis products is comparable to earlier results, that of mobilized U from the pellets is < 1% of the total amount of oxidized species. Thus, the radiation chemical yield (-value) for mobilized hexavalent U is < 0.01 ions/100 eV. In spite of the low radiation yield for the corrosion, the rate of UO2 dissolution is higher than expected for the concentrations of long-lived oxidizing radiolysis compounds found in the solutions.

  17. Investigating microstructural evolution during the electroreduction of UO2 to U in LiCl-KCl eutectic using focused ion beam tomography

    Science.gov (United States)

    Brown, L. D.; Abdulaziz, R.; Tjaden, B.; Inman, D.; Brett, D. J. L.; Shearing, P. R.

    2016-11-01

    Reprocessing of spent nuclear fuels using molten salt media is an attractive alternative to liquid-liquid extraction techniques. Pyroelectrochemical processing utilizes direct, selective, electrochemical reduction of uranium dioxide, followed by selective electroplating of a uranium metal. Thermodynamic prediction of the electrochemical reduction of UO2 to U in LiCl-KCl eutectic has shown to be a function of the oxide ion activity. The pO2- of the salt may be affected by the microstructure of the UO2 electrode. A uranium dioxide filled "micro-bucket" electrode has been partially electroreduced to uranium metal in molten lithium chloride-potassium chloride eutectic. This partial electroreduction resulted in two distinct microstructures: a dense UO2 and a porous U metal structure were characterised by energy dispersive X-ray spectroscopy. Focused ion beam tomography was performed on five regions of this electrode which revealed an overall porosity ranging from 17.36% at the outer edge to 3.91% towards the centre, commensurate with the expected extent of reaction in each location. The pore connectivity was also seen to reduce from 88.32% to 17.86% in the same regions and the tortuosity through the sample was modelled along the axis of propagation of the electroreduction, which was seen to increase from a value of 4.42 to a value of infinity (disconnected pores). These microstructural characteristics could impede the transport of O2- ions resulting in a change in the local pO2- which could result in the inability to perform the electroreduction.

  18. High temperature investigation of the solid/liquid transition in the PuO2-UO2-ZrO2 system

    Science.gov (United States)

    Quaini, A.; Guéneau, C.; Gossé, S.; Sundman, B.; Manara, D.; Smith, A. L.; Bottomley, D.; Lajarge, P.; Ernstberger, M.; Hodaj, F.

    2015-12-01

    The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2-PuO2-ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2-PuO2-ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

  19. On the catalytic effects of UO 2(s) and Pd(s) on the reaction between H 2O 2 and H 2 in aqueous solution

    Science.gov (United States)

    Nilsson, Sara; Jonsson, Mats

    2008-01-01

    The possible catalytic effects of UO 2 and Pd (as a model for noble metal particles) on the reaction between H 2O 2 and H 2 have been studied experimentally. The experiments were performed in aqueous solution using an autoclave. The aqueous solutions were pressurized with H 2 or N 2 and the H 2O 2 concentration was measured as a function of time. The experiments clearly showed that Pd catalyzes the reaction between H 2O 2 and H 2 while UO 2 has no catalytic effect. The rate constant of the reaction between H 2O 2 and H 2 catalyzed by Pd was found to be close to diffusion controlled and independent of the H 2 pressure in the range 1-40 bar. The impact of the catalytic effect on the reaction between H 2O 2 and H 2 on spent nuclear fuel dissolution is, however, fairly small. Other possible effects of noble metal particles are also discussed, e.g. reduction of U(VI) to U(IV) in the liquid and solid phase.

  20. Influence of palm oil fuel ash on fresh and mechanical properties of self-compacting concrete

    Indian Academy of Sciences (India)

    Hossein Mohammadhosseini; A S M Abdul Awal; Abdul Haq Ehsan

    2015-09-01

    This paper presents experimental results of some fresh and hardened state properties of self-compacting concrete (SCC) incorporating palm oil fuel ash (POFA). Three concrete mixes namely ordinary Portland cement (OPC) concrete i.e. concrete with 100% OPC as control, and concrete with 30% and 60% POFA having different water/binder (w/b) ratios of 0.4, 0.45 and 0.5 were prepared. Filling ability, passing ability and segregation resistance of SCC along with strength properties were determined and compared with those of the OPC based SCC. Test results revealed that replacement of POFA in general decreased the workability of concrete with acceptable range. The compressive strength, however, increased with lower w/b ratio and lower replacement of ash. The splitting tensile and flexural strength values have also followed the same trend. The results obtained and the observation made in this study suggest that POFA can suitably be used as supplementary cementing material in SCC.

  1. An Innovative High Thermal Conductivity Fuel Design

    Energy Technology Data Exchange (ETDEWEB)

    PI: James S. Tulenko; Co-PI: Ronald H. Baney,

    2007-10-14

    Uranium dioxide (UO2) is the most common fuel material in commercial nuclear power reactors. UO2 has the advantages of a high melting point, good high-temperature stability, good chemical compatibility with cladding and coolant, and resistance to radiation. The main disadvantage of UO2 is its low thermal conductivity. During a reactor’s operation, because the thermal conductivity of UO2 is very low, for example, about 2.8 W/m-K at 1000 oC [1], there is a large temperature gradient in the UO2 fuel pellet, causing a very high centerline temperature, and introducing thermal stresses, which lead to extensive fuel pellet cracking. These cracks will add to the release of fission product gases after high burnup. The high fuel operating temperature also increases the rate of fission gas release and the fuel pellet swelling caused by fission gases bubbles. The amount of fission gas release and fuel swelling limits the life time of UO2 fuel in reactor. In addition, the high centerline temperature and large temperature gradient in the fuel pellet, leading to a large amount of stored heat, increase the Zircaloy cladding temperature in a lost of coolant accident (LOCA). The rate of Zircaloy-water reaction becomes significant at the temperature above 1200 oC [2]. The ZrO2 layer generated on the surface of the Zircaloy cladding will affect the heat conduction, and will cause a Zircaloy cladding rupture. The objective of this research is to increase the thermal conductivity of UO2, while not affecting the neutronic property of UO2 significantly. The concept to accomplish this goal is to incorporate another material with high thermal conductivity into the UO2 pellet. Silicon carbide (SiC) is a good candidate, because the thermal conductivity of single crystal SiC is 60 times higher than that of UO2 at room temperature and 30 times higher at 800 oC [3]. Silicon carbide also has the properties of low thermal neutron absorption cross section, high melting point, good chemical

  2. Development of Xe and Kr empirical potentials for CeO2, ThO2, UO2 and PuO2, combining DFT with high temperature MD

    Science.gov (United States)

    Cooper, M. W. D.; Kuganathan, N.; Burr, P. A.; Rushton, M. J. D.; Grimes, R. W.; Stanek, C. R.; Andersson, D. A.

    2016-10-01

    The development of embedded atom method (EAM) many-body potentials for actinide oxides and associated mixed oxide (MOX) systems has motivated the development of a complementary parameter set for gas-actinide and gas-oxygen interactions. 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. The resultant gas potentials were validated against DFT trapping energies and are suitable for simulating combinations of Xe and Kr in solid solutions of CeO2, ThO2, UO2 and PuO2, providing a powerful tool for the atomistic simulation of conventional nuclear reactor fuel UO2 as well as advanced MOX fuels.

  3. Development of Xe and Kr empirical potentials for CeO2, ThO2, UO2 and PuO2, combining DFT with high temperature MD.

    Science.gov (United States)

    Cooper, M W D; Kuganathan, N; Burr, P A; Rushton, M J D; Grimes, R W; Stanek, C R; Andersson, D A

    2016-10-12

    The development of embedded atom method (EAM) many-body potentials for actinide oxides and associated mixed oxide (MOX) systems has motivated the development of a complementary parameter set for gas-actinide and gas-oxygen interactions. 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. The resultant gas potentials were validated against DFT trapping energies and are suitable for simulating combinations of Xe and Kr in solid solutions of CeO2, ThO2, UO2 and PuO2, providing a powerful tool for the atomistic simulation of conventional nuclear reactor fuel UO2 as well as advanced MOX fuels.

  4. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  5. UO2核芯还原炉设计模拟研究%Simulation Study of UO2 Kernel Reduction Furnace Design

    Institute of Scientific and Technical Information of China (English)

    刘马林; 刘业飞; 郭文利; 梁彤祥

    2011-01-01

    以N-S方程和k-ε湍流模型为基础,针对UO2核芯颗粒制备过程中的焙烧还原炉设备,采用计算流体力学方法模拟考察了南非和国内正在使用的两种还原炉体设计及人流速度对内部流场的影响.从模拟结果中可发现,两种炉体设计均无法实现气流在轴向上的均匀分布,而是呈现出炉体顶部气量大、底部气量小的分布状态,这是导致颗粒还原不均匀的原因之一,且这种不均匀性随气速增加变化不大.在分析轴向压力变化影响径向气流分布的基础上对还原炉体进行了改进,提出了一种新型设计,模拟结果证实改进后的炉体设计能够实现径向气流在轴向上更为均匀的分布,因而可推定该新型炉体设计可使炉内不同轴向高度处的颗粒还原更加均匀.%Based on the N-S equations and the k-e turbulence model, different kinds of UO2 kernel reduction furnace equipments in PBMR, South Africa and INET, China were numerically simulated using computational fluid dynamics method. The simulation results show that these two kinds of furnace designs can not be achieved on the uniform distribution of gas flow in the axial direction, but show large volume at the top and small volume at the bottom of the furnace, and this is one of the reasons of non-uniform particle reduction. Improved design was proposed based on the analysis of changes of axial pressure in the furnace. Simulation results demonstrate that the improved furnace design is suitable for obtaining a more uniform distribution of the gas in the axial direction. It can be concluded that the improved furnace design will improve particle reduction effects.

  6. [La(UO2)V2O7][(UO2)(VO4)] the first lanthanum uranyl-vanadate with structure built from two types of sheets based upon the uranophane anion-topology

    Science.gov (United States)

    Mer, A.; Obbade, S.; Rivenet, M.; Renard, C.; Abraham, F.

    2012-01-01

    The new lanthanum uranyl vanadate divanadate, [La(UO2)V2O7][(UO2)(VO4)] was obtained by reaction at 800 °C between lanthanum chloride, uranium oxide (U3O8) and vanadium oxide (V2O5) and the structure was determined from single-crystal X-ray diffraction data. This compound crystallizes in the orthorhombic system with space group P212121 and unit-cell parameters a=6.9470(2) Å, b=7.0934(2) Å, c=25.7464(6) Å, V=1268.73(5) Å3, Z=4. A full matrix least-squares refinement yielded R1=0.0219 for 5493 independent reflections. The crystal structure is characterized by the stacking of uranophane-type sheets [(UO2)(VO4)]-∞2 and double layers [La(UO2)(V2O7)]+∞2 connected through La-O bonds involving the uranyl oxygen of the uranyl-vanadate sheets. The double layers result from the connection of two [La(UO2)(VO4)2]-∞2 sheets derived from the uranophane anion-topology by replacing half of the uranyl ions by lanthanum atoms and connected through the formation of divanadate entities.

  7. Phase equilibria in the UO 2-austenitic steel system up to 3000°C

    Science.gov (United States)

    Kleykamp, Heiko

    1997-08-01

    The pseudobinary UO 2-austenitic steel system was investigated by DTA up to 1500°C, by isothermal annealing up to 2000°C, by induction heating up to 2850°C and by arc melting up to about 3000°C. The system is characterized by a degenerate eutectic at 1433°C on the steel side and a monotectic at 2830°C and about 1 mol% steel. The maximum solubility of steel in solid U0 2 is 0.6 mol%, that in liquid U0 2 at 3000°C is about 4 mol%. U0 2 and steel form (Fe, Mn, Cr) 2O 3 precipitates between 1300 and 2600°C as U02 becomes hypostoichiometric. Liquid steel is stabilized to higher temperatures above its boiling point at 2790°C by dissolution of uranium and decomposes peritectically to liquid UO 2-χ and gas at estimated 3200°C. The critical data of the single-phase U0 2-steel melt based on the application of the Redlich-Kister model are Tc = 4900°C, Pc = 300 bar and χ c, steel = 0.41.

  8. Disordering and dynamic self-organization in stoichiometric UO2 at high temperatures

    Science.gov (United States)

    Annamareddy, Ajay; Eapen, Jacob

    2017-01-01

    Neutron scattering experiments show significant oxygen disorder in UO2 at temperatures above 2000 K. The nature of the disorder, however, has not been ascertained with certainty. Using atomistic simulations and metrics from statistical mechanics we show that the oxygen anions predominantly hop from one native (tetrahedral) lattice site to another, above a characteristic temperature Tα (∼2000 K). Interestingly, we discover two types of disorder - the first one, which is a measure of the fraction of anions that are displaced from their native sites, portrays a monotonic increase with temperature and shows excellent conformity to neutron scattering data. The second metric based on the mean square displacement of the anions in an isoconfigurational ensemble demonstrates a dynamic self-organization behavior in which the anions are spatially correlated to those with similar mobility. This dynamic self-organization, however, experiences a non-monotonic variation with temperature depicting a maximum near the Bredig or λ-transition. We further establish that the thermodynamic metric cp/T, which is equal to the rate of change of entropy with temperature, is a key entropic indicator of the dynamic self-organization among the oxygen anions in UO2 at high temperatures.

  9. Experimental characterization and modelling of UO2 behavior at high temperatures and high strain rates

    Science.gov (United States)

    Salvo, Maxime; Sercombe, Jérôme; Ménard, Jean-Claude; Julien, Jérôme; Helfer, Thomas; Désoyer, Thierry

    2015-01-01

    This work presents an experimental characterization of uranium dioxide (UO2) in compression under Reactivity Initiated Accident (RIA) conditions. Pellet samples were tested at four temperatures (1100, 1350, 1550 and 1700 °C) and at a strain rate varying over 4 decades (10-4-10-3-10-2-10-1 /s). The experimental results show that the stress-strain curves cannot be fitted with a unique power law as it is the case at smaller strain rates (10-9-10-5 /s). A strain-hardening also appears in most of the tests. The microstructural observations show a pronounced evolution of the porosity at the pellet center during the tests. A hyperbolic sine model which accounts for volume variations (pore compressibility) was therefore proposed to describe the behavior of UO2 on a large range of temperatures (1100 - 1700 °C) and strain rates (10-9-10-1 /s). The Finite Element simulations of the compression tests lead to results (maximum stress, axial and hoop strain distribution, porosity distribution) in good agreement with the measurements. The model was then assessed on a database of more than two hundred creep tests.

  10. Why a steady state void size distribution in irradiated UO2? A modeling approach

    Science.gov (United States)

    Maillard, S.; Martin, G.; Sabathier, C.

    2016-05-01

    In UO2 pellets irradiated in standard water reactor, Xe nano-bubbles nucleate, grow, coarsen and finally reach a quasi steady state size distribution: transmission electron microscope (TEM) observations typically report a concentration around 10-4 nm-3 and a radius around 0.5 nm. This phenomenon is often considered as a consequence of radiation enhanced diffusion, precipitation of gas atoms and ballistic mixing. However, in UO2 thin foils irradiated with energetic ions at room temperature, a nano-void population whose size distribution reaches a similar steady state can be observed, although quasi no foreign atoms are implanted nor significant cation vacancy diffusion expected in conditions. Atomistic simulations performed at low temperature only address the first stage of the process, supporting the assumption of void heterogeneous nucleation: 25 keV sub-cascades directly produce defect aggregates (loops and voids) even in the absence of gas atoms and thermal diffusion. In this work a semi-empirical stochastic model is proposed to enlarge the time scale covered by simulation up to damage levels where every point in the material undergoes the superposition of a large number of sub-cascade impacts. To account for the accumulation of these impacts, simple rules inferred from the atomistic simulation results are used. The model satisfactorily reproduces the TEM observations of nano-voids size and concentration, which paves the way for the introduction of a more realistic damage term in rate theory models.

  11. The Shipment of UO2 Spent Fuel as First Batch by HWRR

    Institute of Scientific and Technical Information of China (English)

    1995-01-01

    4.4TheShipmentofUO_2SpentFuelasFirstBatchbyHWRRYangDaoliang72groupsofUO_2spentfuelassembliesloadedin6shippingcaskswereshipped...

  12. Influence of the Electronic Structure and Optical Properties of CeO2 and UO2 for Characterization with UV-Laser Assisted Atom Probe Tomography

    Energy Technology Data Exchange (ETDEWEB)

    Billy Valderrama; H.B. Henderson; C. Yablinsky; J. Gan; T.R. Allen; M.V. Manuel

    2015-09-01

    Oxide materials are used in numerous applications such as thermal barrier coatings, nuclear fuels, and electrical conductors and sensors, all applications where nanometer-scale stoichiometric changes can affect functional properties. Atom probe tomography can be used to characterize the precise chemical distribution of individual species and spatially quantify the oxygen to metal ratio at the nanometer scale. However, atom probe analysis of oxides can be accompanied by measurement artifacts caused by laser-material interactions. In this investigation, two technologically relevant oxide materials with the same crystal structure and an anion to cation ratio of 2.00, pure cerium oxide (CeO2) and uranium oxide (UO2) are studied. It was determined that electronic structure, optical properties, heat transfer properties, and oxide stability strongly affect their evaporation behavior, thus altering their measured stoichiometry, with thermal conductance and thermodynamic stability being strong factors.

  13. 羟基磷灰石及氟掺杂对UO2+2的吸附性能研究%The Effect of Hydroxyapatite and Fluorine-Substituted on the Adsorption of UO2+2

    Institute of Scientific and Technical Information of China (English)

    李永鹏; 张红平; 林晓艳

    2015-01-01

    The hydroxyapatite(HA)and fluorine -substituted hydroxyapatite(FHA93,FHA200)were synthesized by the wet chemical precipitation method. The structures of these samples were analyzed by Fourier transform infrared spectroscopy,X-rays diffraction,Thermogravimetric analysis and X-ray pho-toelectron spectra. Meanwhile,we also explored the capacities for removal of uranyl( UO2+2 )from model wastewater. The effects of the different factors such as pH values,initial UO2+2 concentration,adsorption time and temperature on the adsorption of UO2+2 by hydroxyapatite and fluorine-substituted hydroxyapa-tite were examined. The results show that the maximum removal rate of HA,FHA93 and FHA200 is about 99. 5%. The equilibrium adsorption data is found to be well described by Langmuir model,and the pseu-do-second -order rate model fits the adsorption kinetic data. When the pH is ranged with 2 ~7, FHA200 exhibites excellent removal rate of UO2+2 . Furthermore,FHA93 reached the saturated adsorption capacity of 98% in 60 min. The thermogravimetric analysis and anti-acids tests also indicate that the FHA200 has the best thermal stability and acid resistance among these three materials. Accordingly,we found that the fluorine-substituted hydroxyapatite is more suitable for the treatment of UO2+2 -containing waste water.%采用湿化学沉淀法合成粉末状羟基磷灰石( HA)和氟掺杂羟基磷灰石( FHA93和FHA200),通过红外光谱、X射线衍射、热失重分析及X射线光电子能谱进行表征,研究了其对模拟废水中UO2+2的吸附去除性能,考察了pH值、UO2+2初始浓度、吸附时间和温度等条件对羟基磷灰石和氟掺杂羟基磷灰石去除UO2+2的影响。结果表明:HA, FHA93和FHA2003种吸附剂对UO2+2的最大去除率均在99.5%左右;吸附平衡数据符合Langmuir模型,吸附动力学数据与准二阶动力学方程较吻合;FHA200在pH值2~7范围内均有较高的UO2+2去除率;FHA93在60 min

  14. RELATION BETWEEN PORE MODEL AND CENTER-LINE TEMPERATURE IN HIGH BURN-UP UO2 PELLET

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2010-06-01

    Full Text Available Relation between pore model and center-line temperature of high burn up UO2 Pellet. Temperature distribution has been evaluated by using different model of pore distribution. Typical data of power distribution and coolant data have been chosen in this study. Different core model and core distribution model have been studied for related temperature, in correlation with high burn up thermal properties. Finite element combined finite different adapted from Saturn-1 has been used for calculating the temperature distribution. The center-line temperature for different pore model and related discussion is presented.   Keywords: pore model, high burn up, UO2 pellet, centerline temperature.

  15. 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.

  16. Dynamical simulations of radiation damage induced by 10 keV energetic recoils in UO 2

    Science.gov (United States)

    Tian, X. F.; Gao, T.; Long, Chongsheng; Li, JiuKai; Jiang, Gang; Xiao, Hongxing

    2011-08-01

    We have performed classical molecular dynamics simulations to simulate the primary damage state induced by 10 keV energetic recoils in UO 2. The numbers versus time and the distance distributions for the displaced uranium and oxygen atoms were investigated with the energetic recoils accelerated along four different directions. The simulations suggest that the direction of the primary knock-on atom (PKA) has no effect on the final primary damage state. In addition, it was found that atomic displacement events consisted of replacement collision sequences in addition to the production of Frenkel pairs. The spatial distribution of defects introduced by 10 keV collision cascades was also presented and the results were similar to that of energetic recoils with lower energy.

  17. Mechanical behaviour near grain boundaries of He-implanted UO2 ceramic polycrystals

    Science.gov (United States)

    Ibrahim, M.; Castelier, É.; Palancher, H.; Bornert, M.; Caré, S.; Micha, J.-S.

    2017-01-01

    For studying the micromechanical behaviour of UO2 and characterising the intergranular interaction, polycrystals are implanted with helium ions, inducing strains in a thin surface layer. Laue X-ray micro-diffraction is used to measure the strain field in this implanted layer with a spatial resolution of about 1 μm. It allows a 2D mapping of the strain field in a dozen of grains. These measurements show that the induced strain depends mainly on the crystal orientation, and can be evaluated by a semi-analytical mechanical model. A mechanical interaction of the neighbouring grains has also been evidenced near the grain boundaries, which has been well reproduced by a finite element model. This interaction is shown to increase with the implantation energy (i.e. the implantation depth): it can be neglected at low implantation energy (60 keV), but not at higher energy (500 keV).

  18. Reductive silylation of Cp*UO2((Mes)PDI(Me)) promoted by Lewis bases.

    Science.gov (United States)

    Kiernicki, J J; Harwood, J S; Fanwick, P E; Bart, S C

    2016-02-21

    Functionalization of the uranyl moiety (UO2(2+)) in Cp*UO2((Mes)PDI(Me)) (1-PDI) ((Mes)PDI(Me) = 2,6-((Mes)N=CMe)2C5H3N; Mes = 2,4,6-triphenylmethyl), which bears a reduced, monoanionic pyridine(diimine) ligand, is reported. Silylating reagents, R3Si-X (R = Me, X = Cl, I, OTf, SPh; R = Ph, X = Cl), effectively add across the strong O=U=O bonds in the presence of the Lewis base, OPPh3, generating products of the form (R3SiO)2UX2(OPPh3)2 (R = Me, X = I (2-OPPh3), Cl (3-OPPh3), SPh (5-OPPh3), OTf (6-OPPh3); R = Ph, X = Cl (4-OPPh3)). During this transformation, reduction to uranium(iv) occurs with loss of (Cp*)2 and (Mes)PDI(Me), each of which acts as a one-electron source. In the reaction, the Lewis base serves to activate the silyl halide, generating a more electrophilic silyl group, as determined by (29)Si NMR spectroscopy, that undergoes facile transfer to the oxo groups. Complete U-O bond scission was accomplished by treating the uranium(iv) disiloxide compounds with additional silylating reagent, forming the family (Ph3PO)2UX4. All compounds were characterized by (1)H NMR, infrared, and electronic absorption spectroscopies. X-ray crystallographic characterization was used to elucidate the structures of 2-OPPh3, 4-OPPh3, 5-OPPh3, and 6-OPPh3.

  19. Synthesis and X-ray diffraction study of (Cs0.5Ba0.25)[UO2(CH3COO)3] and Ba0.5[UO2(CH3COO)3

    Science.gov (United States)

    Serezhkina, L. B.; Vologzhanina, A. V.; Klepov, V. V.; Serezhkin, V. N.

    2011-03-01

    Uranyl triacetate complexes (Cs0.5Ba0.25)[UO2(CH3COO)3] ( I) and Ba0.5[UO2(CH3COO)3] ( II) are synthesized for the first time and their structures are determined by X-ray diffraction. Both compounds crystallize in the cubic crystal system. The crystal data are as follows: a = 17.3289(7) Å, V = 5203.7(4) Å3, space group I213 and Z = 16 ( I); a = 17.0515(8)Å, V = 4957.8(4) Å3, space group I bar 4 3 d, and Z = 16 ( II). In I and II, as in all uranyl triacetates studied earlier, the coordination polyhedron of the uranium atom is a hexagonal bipyramid whose vertices are occupied by the oxygen atoms of the uranyl and three acetate groups. The uranium-containing group belongs to the AB {3/01} ( A = UO{2/2+}, B 01 = CH3COO-) crystal chemical group of uranyl complexes. It was found that compound II is isostructural to the (Rb0.50Ba0.25)[UO2(CH3COO)3] studied earlier.

  20. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  1. The solidification behaviour of the UO2–ThO2 system in a laser heating study

    NARCIS (Netherlands)

    Böhler, R.; Quaini, A.; Capriotti, L.; Cakir, P.; Benes, O.; Boboridis, K.; Guiot, A.; Luzzi, L.; Konings, R.J.M.; Manara, D.

    2014-01-01

    The high-temperature phase diagram of the UO2–ThO2 system has been experimentally revisited in the present study for the first time since 1970, using a laser heating approach combined with fast pyrometry in a thermal arrest method. The melting/solidification temperature, which is of fundamental info

  2. Toward equatorial planarity about uranyl: synthesis and structure of tridentate nitrogen-donor {UO2}2+ complexes.

    Science.gov (United States)

    Copping, Roy; Jeon, Byoungseon; Pemmaraju, C Das; Wang, Shuao; Teat, Simon J; Janousch, Markus; Tyliszczak, Tolek; Canning, Andrew; Grønbech-Jensen, Niels; Prendergast, David; Shuh, David K

    2014-03-01

    The reaction of UO2Cl2·3THF with the tridentate nitrogen donor ligand 2,6-bis(2-benzimidazolyl)pyridine (H2BBP) in pyridine leads to the formation of three different complexes: [(UO2)(H2BBP)Cl2] (1), [(UO)2(HBBP)(Py)Cl] (2), and [(UO2)(BBP)(Py)2] (3) after successive deprotonation of H2BBP with a strong base. Crystallographic determination of 1-3 reveals that increased charge through ligand deprotonation and displacement of chloride leads to equatorial planarity about uranyl as well as a more compact overall coordination geometry. Near-Edge X-ray Absorption Fine Structure (NEXAFS) spectra of 1-3 at the U-4d edges have been recorded using a soft X-ray Scanning Transmission X-ray Microscope (STXM) and reveal the uranium 4d5/2 and 4d3/2 transitions at energies associated with uranium in the hexavalent oxidation state. First-principles Density Functional Theory (DFT) electronic structure calculations for the complexes have been performed to determine and validate the coordination characteristics, which correspond well to the experimental results.

  3. Investigation of oxygen disorder, thermal parameters, lattice vibrations and elastic constants of UO2 and ThO2 at temperatures up to 2 930 K

    DEFF Research Database (Denmark)

    Clausen, Kurt Nørgaard; Hayes, W; Hutchings, M.T.

    1984-01-01

    A knowledge of the thermodynamic properties of UO2 at temperatures in the region 1 500-3 100 K is of importance in reactor safety calculations, yet there are relatively few detailed experimental data available. In particular the major question of whether Frenkel disorder occurs in UO2 at high tem...

  4. Adsorption characteristics of UO(2)(2+) and Th(4+) ions from simulated radioactive solutions onto chitosan/clinoptilolite sorbents.

    Science.gov (United States)

    Humelnicu, Doina; Dinu, Maria Valentina; Drăgan, Ecaterina Stela

    2011-01-15

    Adsorption features of UO(2)(2+) and Th(4+) ions from simulated radioactive solutions onto a novel chitosan/clinoptilolite (CS/CPL) composite as beads have been investigated compared with chitosan cross-linked with epichlorohydrin. The effects of contact time, the initial metal ion concentration, sorbent mass and temperature on the adsorption capacity of the CS-based sorbents were investigated. The adsorption kinetics was well described by the pseudo-second order equation, and the adsorption isotherms were better fitted by the Sips model. The maximum experimental adsorption capacities were 328.32 mg Th(4+)/g composite, and 408.62 mg UO(2)(2+)/g composite. The overall adsorption tendency of CS/CPL composite toward UO(2)(2+) and Th(4+) radiocations in the presence of Cu(2+), Fe(2+) and Al(3+), under competitive conditions, followed the order: Cu(2+)>UO(2)(2+)>Fe(2+)>Al(3+), and Cu(2+)>Th(4+)>Fe(2+)>Al(3+), respectively. The negative values of Gibbs free energy of adsorption indicated the spontaneity of the adsorption of radioactive ions on both the CS/CPL composite and the cross-linked CS. The desorption level of UO(2)(2+) from the composite CS/CPL, by using 0.1M Na(2)CO(3), was around 92%, and that of Th(4+) ions, performed by 0.1M HCl, was around 85%, both values being higher than the desorption level of radiocations from the cross-linked CS, which were 89% and 83%, respectively.

  5. Kinetics study on the dissolution of UO2 particles by microwave and conventional heating in 4 mol/L nitric acid

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The dissolution of UO2 particles in 4 mol·L-1 nitric acid medium at temperatures of 90-110℃ by mi- crowave heating and conventional heating has been investigated, respectively. It is found that the dissolution ratios of UO2 particles by microwave heating were 10%-40% higher than that by conven- tional heating. Kinetics research shows that the dissolution of UO2 particles in 4 mol·L-1 nitric acid is controlled by the diffusion control model for microwave heating and by the surface reaction control model for conventional heating. The diffusion control model for the dissolution of UO2 particles by mi- crowave heating could be explained by the diffuseness on the surface of UO2 particles.

  6. Intergranular fracture in UO2: derivation of traction-separation law from atomistic simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yongfeng Zhang; Paul C Millett; Michael R Tonks; Xian-Ming Bai; S Bulent Biner

    2013-10-01

    In this study, the intergranular fracture behavior of UO2 was studied by molecular dynamics simulations using the Basak potential. In addition, the constitutive traction-separation law was derived from atomistic data using the cohesive-zone model. In the simulations a bicrystal model with the (100) symmetric tilt E5 grain boundaries was utilized. Uniaxial tension along the grain boundary normal was applied to simulate Mode-I fracture. The fracture was observed to propagate along the grain boundary by micro-pore nucleation and coalescence, giving an overall intergranular fracture behavior. Phase transformations from the Fluorite to the Rutile and Scrutinyite phases were identified at the propagating crack tips. These new phases are metastable and they transformed back to the Fluorite phase at the wake of crack tips as the local stress concentration was relieved by complete cracking. Such transient behavior observed at atomistic scale was found to substantially increase the energy release rate for fracture. Insertion of Xe gas into the initial notch showed minor effect on the overall fracture behavior.

  7. Sequestering uranium from UO2(CO3)3(4-) in seawater with amine ligands: density functional theory calculations.

    Science.gov (United States)

    Guo, Xiaojing; Huang, Liangliang; Li, Cheng; Hu, Jiangtao; Wu, Guozhong; Huai, Ping

    2015-06-14

    The polystyrene-supported primary amine -CH2NH2 has shown an at least 3-fold increase in uranyl capacity compared to a diamidoxime ligand on a polystyrene support. This study aims to understand the coordination of substitution complexes from UO2(CO3)3(4-) and amines using density functional theory calculations. Four kinds of amines (diethylamine (DEA), ethylenediamine (EDA), diethylenetriamine (DETA) and triethylenetetramine (TETA)) were selected because they belong to different classes and have different chain lengths. The geometrical structures, electronic structures and the thermodynamic stabilities of various substitution complexes, as well as the trends in their calculated properties were investigated at equilibrium. In these optimized complexes, DEA groups bind to uranyl as monodentate ligands; EDA groups serve as monodentate and bidentate ligands; DETA groups act as monodentate and tridentate ligands; while TETA groups serve as monodentate, bidentate and tridentate ligands. The thermodynamic analysis confirmed that the primary amines coordinate to uranyl more strongly than does the secondary amine. The stabilities of substitution complexes with primary amines were calculated to decrease with increasing chain length of the amine, except for UO2(L2)(2+). Of the complexes analyzed, only UO2L(CO3)2(2-) (L = EDA and DETA) and UO2L2CO3 (L = EDA) were predicted to form from the substitution reactions with UO2(CO3)3(4-) and protonated amines as reactants in aqueous solution. Amines were calculated to be comparable to, or sometimes weaker than, amidoximate in replacing CO3(2-) in UO2(CO3)3(4-) to coordinate to uranium. Therefore, the coordination mechanism, in which amines replace carbonates to bind to uranyl, is not primarily responsible for the experimentally observed 3-fold or greater increase in uranyl capacity of primary amines compared to a diamidoxime ligand. Based on the results of our calculations, we believe that the cation exchange mechanism, in which the

  8. Two new lithium uranyl tungstates Li 2(UO 2)(WO 4) 2 and Li 2(UO 2) 4(WO 4) 4O with framework based on the uranophane sheet anion topology

    Science.gov (United States)

    Obbade, S.; Yagoubi, S.; Dion, C.; Saadi, M.; Abraham, F.

    2004-04-01

    Two new lithium uranyl tangstates Li 2(UO 2)(WO 4) 2 and Li 2(UO 2) 4(WO 4) 4O have been prepared by high-temperature solid state reactions of Li 2CO 3, U 3O 8 and WO 3. For each compound, the crystal structure was determined by single crystal X-ray diffraction data, using a Brucker diffractometer, equipped with a SMART CCD detector and Mo Kα radiation. The crystal structures were solved at room temperature by direct methods followed by Fourier difference techniques, and refined by a least square procedure on the basis of F2 for all independent reflections, to R1=0.035 for 65 refined parameters and 807 reflections with I⩾2 σ( I) for Li 2(UO 2)(WO 4) 2 and to R1=0.051 for 153 refined parameters and 1766 reflections with I⩾2 σ( I) for Li 2(UO 2) 4(WO 4) 4O. The crystal structure of Li 2(UO 2)(WO 4) 2 is formed by perovskite sheets of WO 6 octahedra, one octahedron thickness, connected together by (UO 5) ∞ infinite chains, and creating tunnels parallel to the c-axis. The lithium atoms are localized in the tunnels. The structure can be deduced from that of U MO 5 ( M=Mo, V, Nb) compounds by the replacement of half U atoms by Li. The crystal structure of Li 2(UO 2) 4(WO 4) 4O consists of UO 7 pentagonal bipyramids, UO 6 tetragonal bipyramids and WO 6 distorted octahedra linked together to form a three-dimensional framework creating paralleled channels filled with lithium cations. The structure can also be described by the stacking of layers with the uranophane sheet anion topology similar to those obtained in U MO 5 ( M=Mo, V, Nb, Sb) compounds with an ordered population of pentagons by U and Li and of squares by U and W. The measured conductivities are comparable to those of the better Li + ion conductor solid electrolytes such as LISICON or Li- β-alumina. Crystallographic data: Li 2(UO 2)(WO 4) 2, orthorhombic symmetry, space group Pbcn and unit cell parameters a=7.9372(15) Å, b=12.786(2) Å, c=7.4249(14) Å, ρcal=6.87(2) g/cm 3, ρmes=6.89(1) g/cm 3 and Z

  9. Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.; Gilbert, Andrew J.; Jacomb-Hood, Timothy W.; VanDevender, Brent A.; Wood, Lynn S.

    2017-07-17

    Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a one meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.

  10. Foundations for the definition of MOX fuel quality requirements

    Science.gov (United States)

    Bairiot, H.; Deramaix, P.; Mostin, N.; Trauwaert, E.; Vanderborck, Y.

    1991-02-01

    The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: - the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, - the fuel enrichment process takes place in the manufacturing plant, - the radioactivity of Pu imposes handling constraints, - Pu ages quite rapidly, altering its isotopic composition during storage, - the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart from UO 2 fuel.

  11. Synthesis and structure of (Rb0.50Ba0.25)[UO2(CH3COO)3

    Science.gov (United States)

    Serezhkina, L. B.; Peresypkina, E. V.; Virovets, A. V.; Klepov, V. V.

    2010-03-01

    A new compound (Rb0.50Ba0.25)[UO2(CH3COO)3] is synthesized and its crystal structure is studied by X-ray diffraction. The compound crystallizes in the form of yellow plates belonging to the cubic crystal system. The unit cell parameter a = 17.0367(1) Å, V = 4944.89(5) Å3, space group I bar 4 3 d, Z = 16, and R = 0.0182. The coordination polyhedron of the uranium atom is a hexagonal bipyramid with oxygen atoms of three acetate groups and the uranyl group in the vertices. The crystal chemical formula of the uranium-containing group is AB {3/01}( A = UO{2/2+}, B 01 = CH3COO-). The oxygen atoms of the acetate groups that enter the coordination polyhedron of uranium are bound to barium and rubidium atoms.

  12. First-principles calculations of momentum distributions of annihilating electron-positron pairs in defects in UO2

    Science.gov (United States)

    Wiktor, Julia; Jomard, Gérald; Torrent, Marc; Bertolus, Marjorie

    2017-01-01

    We performed first-principles calculations of the momentum distributions of annihilating electron-positron pairs in vacancies in uranium dioxide. Full atomic relaxation effects (due to both electronic and positronic forces) were taken into account and self-consistent two-component density functional theory schemes were used. We present one-dimensional momentum distributions (Doppler-broadened annihilation radiation line shapes) along with line-shape parameters S and W. We studied the effect of the charge state of the defect on the Doppler spectra. The effect of krypton incorporation in the vacancy was also considered and it was shown that it should be possible to observe the fission gas incorporation in defects in UO2 using positron annihilation spectroscopy. We suggest that the Doppler broadening measurements can be especially useful for studying impurities and dopants in UO2 and of mixed actinide oxides.

  13. A Structural and Spectroscopic Study of the First Uranyl Selenocyanate, [Et4N]3[UO2(NCSe5

    Directory of Open Access Journals (Sweden)

    Stefano Nuzzo

    2016-02-01

    Full Text Available The first example of a uranyl selenocyanate compound is reported. The compound [Et4N]3[UO2(NCSe5] has been synthesized and fully characterized by vibrational and multinuclear (1H, 13C{1H} and 77Se{1H} NMR spectroscopy. The photophysical properties have also been recorded and trends in a series of uranyl pseudohalides discussed. Spectroscopic evidence shows that the U–NCSe bonding is principally ionic. An electrochemical study revealed that the reduced uranyl(V species is unstable to disproportionation and a ligand based oxidation is also observed. The structure of [Et4N]4[UO2(NCSe5][NCSe] is also presented and Se···H–C hydrogen bonding and Se···Se chalcogen–chalcogen interactions are seen.

  14. Simple (17) O NMR method for studying electron self-exchange reaction between UO2 (2+) and U(4+) aqua ions in acidic solution.

    Science.gov (United States)

    Bányai, István; Farkas, Ildikó; Tóth, Imre

    2016-06-01

    (17) O NMR spectroscopy is proven to be suitable and convenient method for studying the electron exchange by following the decrease of (17) O-enrichment in U(17) OO(2+) ion in the presence of U(4+) ion in aqueous solution. The reactions have been performed at room temperature using I = 5 M ClO4 (-) ionic medium in acidic solutions in order to determine the kinetics of electron exchange between the U(4+) and UO2 (2+) aqua ions. The rate equation is given as R = a[H(+) ](-2)  + R', where R' is an acid independent parallel path. R' depends on the concentration of the uranium species according to the following empirical rate equation: R' = k1 [UO(2 +) ](1/2) [U(4 +) ](1/2)  + k2 [UO(2 +) ](3/2) [U(4 +) ](1/2) . The mechanism of the inverse H(+) concentration-dependent path is interpreted as equilibrium formation of reactive UO2 (+) species from UO2 (2+) and U(4+) aqua ions and its electron exchange with UO2 (2+) . The determined rate constant of this reaction path is in agreement with the rate constant of UO2 (2+) -UO2 (+) , one electron exchange step calculated by Marcus theory, match the range given experimentally of it in an early study. Our value lies in the same order of magnitude as the recently calculated ones by quantum chemical methods. The acid independent part is attributed to the formation of less hydrolyzed U(V) species, i.e. UO(3+) , which loses enrichment mainly by electron exchange with UO2 (2+) ions. One can also conclude that (17) O NMR spectroscopy, or in general NMR spectroscopy with careful kinetic analysis, is a powerful tool for studying isotope exchange reactions without the use of sophisticated separation processes. Copyright © 2015 John Wiley & Sons, Ltd.

  15. Raman spectroscopic investigation of thorium dioxide-uranium dioxide (ThO₂-UO₂) fuel materials.

    Science.gov (United States)

    Rao, Rekha; Bhagat, R K; Salke, Nilesh P; Kumar, Arun

    2014-01-01

    Raman spectroscopic investigations were carried out on proposed nuclear fuel thorium dioxide-uranium dioxide (ThO2-UO2) solid solutions and simulated fuels based on ThO2-UO2. Raman spectra of ThO2-UO2 solid solutions exhibited two-mode behavior in the entire composition range. Variations in mode frequencies and relative intensities of Raman modes enabled estimation of composition, defects, and oxygen stoichiometry in these compounds that are essential for their application. The present study shows that Raman spectroscopy is a simple, promising analytical tool for nondestructive characterization of this important class of nuclear fuel materials.

  16. Anomalous dispersion and band gap reduction in UO2+x and its possible coupling to the coherent polaronic quantum state

    Science.gov (United States)

    Conradson, Steven D.; Andersson, David A.; Bagus, Paul S.; Boland, Kevin S.; Bradley, Joseph A.; Byler, Darrin D.; Clark, David L.; Conradson, Dylan R.; Espinosa-Faller, Francisco J.; Lezama Pacheco, Juan S.; Martucci, Mary B.; Nordlund, Dennis; Seidler, Gerald T.; Valdez, James A.

    2016-05-01

    Hypervalent UO2, UO2(+x) formed by both addition of excess O and photoexcitation, exhibits a number of unusual or often unique properties that point to it hosting a polaronic Bose-Einstein(-Mott) condensate. A more thorough analysis of the O X-ray absorption spectra of UO2, U4O9, and U3O7 shows that the anomalous increase in the width of the spectral features assigned to predominantly U 5f and 6d final states that points to increased dispersion of these bands occurs on the low energy side corresponding to the upper edge of the gap bordered by the conduction or upper Hubbard band. The closing of the gap by 1.5 eV is more than twice as much as predicted by calculations, consistent with the dynamical polaron found by structural measurements. In addition to fostering the excitation that is the proposed mechanism for the coherence, the likely mirroring of this effect on the occupied, valence side of the gap below the Fermi level points to increased complexity of the electronic structure that could be associated with the Fermi topology of BEC-BCS crossover and two band superconductivity.

  17. Improving mechanical properties and microstructure development of fiber reinforced ceramic nuclear fuel.

    OpenAIRE

    SACRAMENTO SANTANA, HESDRAS HENRIQUE

    2014-01-01

    At the present work the UO2 fuel production process was extensively studied and analyzed. The objectives of such investigation were to understand and analyze the influence of different additives and the variation of the production process steps on the microstructure and consequently in the mechanical strength of the nuclear fuel pellet. Moreover, an improvement of the qualitative characteristics of the ceramic fuel pellets was also aimed. For this purpose UO2 pellets without ad...

  18. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    Energy Technology Data Exchange (ETDEWEB)

    Miko, David K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  19. Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

    Science.gov (United States)

    McCoy, Kevin; Mays, Claude

    2008-04-01

    The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO 2 fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor kinf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO 2 fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO 2, UO 2 with 4.0 vol.% BeO, and UO 2 with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO 2 fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the kinf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO 2.

  20. Effects of Beryllium and Compaction Pressure on the Thermal Diffusivity of Uranium Dioxide Fuel Pellets

    Science.gov (United States)

    Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Ferreira, R. A. N.

    2017-09-01

    In nuclear reactors, the performance of uranium dioxide (UO2) fuel is strongly dependent on the thermal conductivity, which directly affects the fuel pellet temperature, the fission gas release and the fuel rod mechanical behavior during reactor operation. The use of additives to improve UO2 fuel performance has been investigated, and beryllium oxide (BeO) appears as a suitable additive because of its high thermal conductivity and excellent chemical compatibility with UO2. In this paper, UO2-BeO pellets were manufactured by mechanical mixing, pressing and sintering processes varying the BeO contents and compaction pressures. Pellets with BeO contents of 2 wt%, 3 wt%, 5 wt% and 7 wt% BeO were pressed at 400 MPa, 500 MPa and 600 MPa. The laser flash method was applied to determine the thermal diffusivity, and the results showed that the thermal diffusivity tends to increase with BeO content. Comparing thermal diffusivity results of UO2 with UO2-BeO pellets, it was observed that there was an increase in thermal diffusivity of at least 18 % for the UO2-2 wt% BeO pellet pressed at 400 MPa. The maximum relative expanded uncertainty (coverage factor k = 2) of the thermal diffusivity measurements was estimated to be 9 %.

  1. Redox Chemistry in Radiation Induced Dissolution of Spent Nuclear Fuel : from Elementary Reactions to Predictive Modeling

    OpenAIRE

    Roth, Olivia

    2008-01-01

    The focus of this doctoral thesis is the redox chemistry involved in radiation induced oxidative dissolution of spent nuclear fuel and UO2 (as a model substance for spent nuclear fuel). It is shown that two electron oxidants are more efficient than one electron oxidants in oxidative dissolution of UO2 at low oxidant concentrations. Furthermore, it is shown that H2O2 is the only oxidant that has to be taken into account in radiation induced dissolution of UO2 under deep repository conditions (...

  2. Mechanical properties of UO2 thin films under heavy ion irradiation using nanoindentation and finite element modeling

    Science.gov (United States)

    Elbakhshwan, Mohamed S.; Miao, Yinbin; Stubbins, James F.; Heuser, Brent J.

    2016-10-01

    The mechanical response of UO2 to irradiation is becoming increasingly important due to the shift to higher burn-up rates in the next generation of nuclear reactors. In the current study, thin films of UO2 were deposited on YSZ substrates using reactive-gas magnetron sputtering. Nanoindentation was used to measure the mechanical properties of the as-grown and irradiated films. Finite element modeling was used to account for the substrate effect on the measurements. In order to study the effect of displacement cascades accompanying gas bubbles, 5000 Å UO2 films were irradiated with 600 keV Kr+ ions at 25 °C and 600 °C. These irradiation conditions were used to confine radiation damage effects and implanted gas within the film. Results showed an increase in the film hardness and yield strength with dose, while elastic modulus initially decreased with irradiation and then kept increasing with dose. The change in hardness and elastic modulus is attributed to the introduction of gas bubbles and displacement cascade damage. Irradiation at 600 °C resulted in a decrease in the hardness and elastic modulus after irradiation using 600 keV Kr+ at a dose of 1E14 ions/cm2. Both hardness and elastic modulus then increased with irradiation dose. This behavior is attributed to recrystallization during irradiation at 600 °C and the formation of nanocrystallite regions with diameter and density that increase with dose. The calculation of the critical resolved shear stress (CRSS) demonstrated that nanocrystals are the primary cause for film hardening based on the Orowan hardening mechanism.

  3. Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL

    2012-01-01

    This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

  4. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

    Energy Technology Data Exchange (ETDEWEB)

    Klann, R. T.; Perret, G.; Nuclear Engineering Division

    2007-10-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

  5. Simulation of alpha dose for predicting radiolytic species at the surface of spent nuclear fuel pellets

    OpenAIRE

    Becker Frank; Kienzler Bernhard

    2014-01-01

    In many countries, spent nuclear fuel is considered as a waste form to be disposed of in underground disposal. Under deep host rock conditions, a reducing environment prevails. In the case of water contact, long-term radionuclide release from the fuel depends on dissolution processes of the UO2 matrix. The dissolution rate of irradiated UO2 is controlled by oxidizing processes facilitated by dissolved species formed by alpharadiolysis of water in contact with spent nuc...

  6. A hydrated ion model of [UO2] 2 + in water: Structure, dynamics, and spectroscopy from classical molecular dynamics

    Science.gov (United States)

    Pérez-Conesa, Sergio; Torrico, Francisco; Martínez, José M.; Pappalardo, Rafael R.; Sánchez Marcos, Enrique

    2016-12-01

    A new ab initio interaction potential based on the hydrated ion concept has been developed to obtain the structure, energetics, and dynamics of the hydration of uranyl in aqueous solution. It is the first force field that explicitly parameterizes the interaction of the uranyl hydrate with bulk water molecules to accurately define the second-shell behavior. The [UO2(H2O)5 ] 2 + presents a first hydration shell U-O average distance of 2.46 Å and a second hydration shell peak at 4.61 Å corresponding to 22 molecules using a coordination number definition based on a multisite solute cavity. The second shell solvent molecules have longer mean residence times than those corresponding to the divalent monatomic cations. The axial regions are relatively de-populated, lacking direct hydrogen bonding to apical oxygens. Angle-solved radial distribution functions as well as the spatial distribution functions show a strong anisotropy in the ion hydration. The [UO2(H2O)5 ] 2 + solvent structure may be regarded as a combination of a conventional second hydration shell in the equatorial and bridge regions, and a clathrate-like low density region in the axial region. Translational diffusion coefficient, hydration enthalpy, power spectra of the main vibrational modes, and the EXAFS spectrum simulated from molecular dynamics trajectories agree fairly well with the experiment.

  7. Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Janne Pakrinen; Marat Khafizov; Lingfeng He; Chris Wetland; Jian Gan; Andrew T. Nelson; David H Hurley; Anter El-Azab; Todd R Allen

    2014-11-01

    The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to 5×1016 He2+/cm2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9×1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

  8. Final Report: Manganese Redox Mediation of UO2 Stability and Uranium Fate in the Subsurface: Molecular and Meter Scale Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Tebo, Bradley M. [OSHU; Tebo, Bradley M.

    2014-09-02

    One strategy to remediate U contamination in the subsurface is the immobilization of U via injection of an electron donor, e.g., acetate, which leads to stimulation of the bioreduction of U(VI), the more mobile form of U, to U(IV), the less mobile form. This process is inevitably accompanied by the sequential reductive dissolution of Mn and Fe oxides and often continuing into sulfate-reducing conditions. When these reducing zones, which accumulate U(IV), experience oxidizing conditions, reduced Fe and Mn can be reoxidized forming Fe and Mn oxides that, along with O2, can impact the stability of U(IV). The focus of our project has been to investigate (i) the effects of Mn(II) on the dissolution of UO2 under both reducing and oxidizing conditions, (ii) the oxidative dissolution of UO2 by soluble Mn(III), (iii) the fate of U once it is oxidized by MnO2 in both laboratory and field settings, and (iv) the effects of groundwater constituents on the coupled Mn(II)/U(IV) oxidation process. Additionally, studies of the interaction of Se, found at the DOE site at Rifle, CO, and Mn cycling were initiated to understand if observed seasonal fluctuations of Se and Mn are directly linked and whether any such linkages can affect the stability of U(IV).

  9. Effect of point defects on the thermal conductivity of UO2: molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-07-21

    The thermal conductivity of uranium dioxide (UO2) fuel is an important materials property that affects fuel performance since it is a key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. [1] The thermal conductivity of UO2 nuclear fuel is also affected by fission gas, fission products, defects, and microstructural features such as grain boundaries. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of irradiation induced point defects on the thermal conductivity of UO2, as a function of defect concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel [2].

  10. Measuring the noble metal and iodine composition of extracted noble metal phase from spent nuclear fuel using instrumental neutron activation analysis.

    Science.gov (United States)

    Palomares, R I; Dayman, K J; Landsberger, S; Biegalski, S R; Soderquist, C Z; Casella, A J; Brady Raap, M C; Schwantes, J M

    2015-04-01

    Masses of noble metal and iodine nuclides in the metallic noble metal phase extracted from spent fuel are measured using instrumental neutron activation analysis. Nuclide presence is predicted using fission yield analysis, and radionuclides are identified and the masses quantified using neutron activation analysis. The nuclide compositions of noble metal phase derived from two dissolution methods, UO2 fuel dissolved in nitric acid and UO2 fuel dissolved in ammonium-carbonate and hydrogen-peroxide solution, are compared.

  11. Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

    2012-04-11

    A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

  12. UO2 CORROSION IN HIGH SURFACE-AREA-TO-VOLUME BATCH EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Finch, Robert J.; Wolf, Stephen F.; Hanchar, John M.; Bates, John K.

    1998-05-11

    Unsaturated drip tests have been used to investigate the alteration of unirradiated UO{sub 2} and spent UO{sub 2} fuel in an unsaturated environment, such as may be expected in the proposed repository at Yucca Mountain. In these tests, simulated groundwater is periodically injected onto a sample at 90 C in a steel vessel. The solids react with the dripping groundwater and water condensed on surfaces to form a suite of U(VI) alteration phases. Solution chemistry is determined from leachate at the bottom of each vessel after the leachate stops interacting with the solids. A more detailed knowledge of the compositional evolution of the leachate is desirable. By providing just enough water to maintain a thin film of water on a small quantity of fuel in batch experiments, we can more closely monitor the compositional changes to the water as it reacts to form alteration phases.

  13. Local study of defects during sintering of UO2: image processing and quantitative analysis tools:

    OpenAIRE

    Eric Girard; Jean-Marc Chaix; François Valdivieso; Patrice Goeuriot; Jacques Lechelle

    2008-01-01

    This paper describes the image analysis tools developed and used to quantify the local effects, of heterogeneities during sintering of ceramic materials applied in nuclear fuels. Specific materials, containing a controlled dispersion of well defined heterogeneities (dense or porous aggregates) in the ceramics matrix have been prepared and sintered. In order to characterize the materials in the vicinity of these likely isolated heterogeneities, large SEM images are first acquired around hetero...

  14. Sequence/structure selective thermal and photochemical cleavage of yeast-tRNA(Phe) by UO(2)2+

    DEFF Research Database (Denmark)

    Nielsen, Peter E.; Møllegaard, N E

    1997-01-01

    The uranyl(VI) ion, UO(2)2+, cleaves yeast tRNA(Phe) both thermally and photochemically. Photochemical cleavage takes place at all positions but exhibits maxima at G10, G18, G30, A38, C49 and A62. Furthermore, in the presence of stoichiometric concentrations of citrate, the cleavage is generally...... suppressed except that strong cleavage at positions G10 and C48-U50 persists, indicating the presence of a high-affinity metal-ion binding site. It is proposed that these photocleavage sites reflect the tertiary structure of the yeast tRNA(Phe) molecule in terms of D-loop/T-loop interaction and anticodon...

  15. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  16. Comparison Of 252Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Laboratory

    2017-03-30

    The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six 3He detectors as well as a shielded and collimated ion chamber. The 3He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a 252Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the 252Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The 252Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the 3He detector, which helps reduce the statistical errors in doubles when using the 252Cf interrogation source instead of

  17. Possible Bose-condensated Behavior in a Quantum Phase Originating in a Collective Excitation in the Chemically and Optically Doped Mott-Hubbard System UO2+x

    Energy Technology Data Exchange (ETDEWEB)

    Conradson, Steven D.; Durakiewicz, Tomasz; Espinosa-Faller, Francisco J.; An, Yong Q.; Andersson , David; Bishop, Alan R.; Boland, Kevin S.; Bradley, Joseph A.; Byler, Darrin D.; Clark, David L.; Conradson, Dylan R.; Conradson, Leilani L.; Costello, Alison E.; Hess, Nancy J.; Lander, Gerard H.; Llobet, Anna; Martucci, Mary B.; de Leon, Jose M.; Nordlund, Dennis; Lezama-Pacheco, Juan S.; Proffen, Thomas E.; Rodriguez, George; Schwarz, Daniel E.; Seidler, Gerald T.; Taylor, Antoinette; Trugman, Stuart A.; Tyson, Trevor A.; Valdez, James A.

    2013-09-23

    The pinned charge defects in U4O9, and U3O7 that are the single phase fluoritestructured derivatives of UO2 have been characterized by U L3 EXAFS at 30, 100, and 200 K, xray and neutron pair distribution function analysis, O K edge XAS and non-resonant inelastic xray scattering, and Raman spectroscopy, while mobile charge defects were investigated by femtosecond time-resolved pump-probe laser spectroscopy on single crystal UO2 between 7 and 300 K. The results from all of these measurements show highly complex and anomalous behaviors, which we attribute to a charge-lattice instability in UO2 that most likely originates in the intersection of the ground U(IV) and a proximate uranyl-like excited state in a conic section, causing a breakdown of the Born-Oppenheimer approximation. Furthermore, the photoinduced quasiparticles undergo a gap-opening condensation between 50 and 60 K. Doped UO2 may therefore exhibit novel correlated electron physics that extends beyond that of the cuprate-manganite-pnictide family of compounds.

  18. High-pressure high-temperature equations of state of UO2 and ThO2

    Science.gov (United States)

    Chidester, B.; Campbell, A. J.; Fischer, R. A.; Reaman, D. M.; Heinz, D. L.; Prakapenka, V.

    2013-12-01

    The actinide elements uranium and thorium are important from the standpoint of heat production in the deep Earth. However, the host mineral phases and distribution of these elements in the mantle are not well constrained. Here we investigate the crystal chemistry and coordination preferences of these elements in simple oxides. Room-temperature high-pressure equations of state of uranium dioxide (UO2) and thorium dioxide (ThO2) have been reported [1-5], but no in situ high-pressure, high-temperature (high P-T) data are available for these compounds. We present results from a high P-T synchrotron x-ray diffraction study of the equations of state and observed phase relations of UO2 and ThO2. High-pressure, high-temperature in situ X-ray diffraction data were obtained at beamline 13-ID-D of the Advanced Photon Source, and room temperature compression data were obtained at beamline 12.2.2 of the Advanced Light Source. We observed that UO2 exists only in the cubic fluorite structure (space group Fm3m) up to 32 GPa and 2300 K. By fitting these P-V-T data to a Birch-Murnaghan equation of state (EOS), we obtain the thermodynamic parameters K0 = 222 × 3.9 GPa and the thermal contribution to pressure, αK = 0.00254 × 0.00044 GPa/K (V0 = 24.51 cm3/mol and K0' = 5, fixed). At ~46 GPa and up to 2400 K, the cubic structure was found to coexist with a high-pressure phase, which we indexed as the orthorhombic Pnma space group (Z=8). Above this pressure, only the orthorhombic structure was observed up to 61 GPa and 2400 K. The EOS parameters for this phase are V0 = 23.84 × 0.20 cm3/mol, K0 = 187 × 10 GPa and αK = 0.00308 × 0.00042 GPa/K (K0' = 4, fixed). Similarly, ThO2 has the fluorite (Fm3m) structure up to ~23 GPa. The EOS parameters for this phase are K0 = 199 × 10 GPa, K0' = 7.1 × 2.0 and αK = 0.00656 × 0.00092 GPa/K (V0 = 26.38 cm3/mol, fixed). The cubic phase was observed to coexist with an orthorhombic phase (Pnma, Z=4) between 27 and 31 GPa and up to 1900 K. Above

  19. OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

    Energy Technology Data Exchange (ETDEWEB)

    Stoven, G.; Klann, R.; Zhong, Z.; Nuclear Engineering Division

    2007-08-28

    The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.

  20. Comparison of 252Cf time correlated induced fission with AmLi induced fission on fresh MTR reserach reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jay Prakash [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-01

    The objectives of this project are to calibrate the Advanced Experimental Fuel Counter (AEFC), benchmark MCNP simulations using experimental results, investigate the effects of change in fuel assembly geometry, and finally to show the boost in doubles count rates with 252Cf active soruces due to the time correlated induced fission (TCIF) effect.

  1. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  2. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  3. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  4. Nuclear Rocket Ceramic Metal Fuel Fabrication Using Tungsten Powder Coating and Spark Plasma Sintering

    Science.gov (United States)

    Barnes, M. W.; Tucker, D. S.; Hone, L.; Cook, S.

    2017-01-01

    Nuclear thermal propulsion is an enabling technology for crewed Mars missions. An investigation was conducted to evaluate spark plasma sintering (SPS) as a method to produce tungsten-depleted uranium dioxide (W-dUO2) fuel material when employing fuel particles that were tungsten powder coated. Ceramic metal fuel wafers were produced from a blend of W-60vol% dUO2 powder that was sintered via SPS. The maximum sintering temperatures were varied from 1,600 to 1,850 C while applying a 50-MPa axial load. Wafers exhibited high density (>95% of theoretical) and a uniform microstructure (fuel particles uniformly dispersed throughout tungsten matrix).

  5. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    was observed for hypostoichiometric fuels, that correspond to the condition used for irradiation. However, if these two formulas are evaluated for O/M = 2.000, the difference between the predictions is negligible (Fig. 1). The difference becomes significant for non-stoichiometric fuels, as shown for O/M = 1.975 in Fig. 1. The microstructure of the FBR fuel with 21.4 wt.% Pu was not described in the paper of Duriez. Taking into account the rigorous experimental methodology used by Duriez (characterisation of the stoichiometry), a possible explanation is an interaction between the plutonium distribution and the stoichiometry. Another parameter having a strong impact on the conductivity is the porosity correction used to obtain the values for 95% TD. This correction is small in the work of Duriez as the samples density is very close to 95% TD. This was also the case for the samples selected by Philipponneau in order to obtain his recommendation. An effect due to differences in the pores shape can also be excluded, as the results are identical for stoichiometric fuels (Fig. 1). Usually the apparent stoichiometry is obtained by heat treatments and checked before and after the measurements, either by XRD or thermogravimetry. However, for non-perfectly homogeneous samples, the gradients in the plutonium distribution induce a non-uniform oxygen distribution, which is difficult to characterise experimentally. It has been proposed by Baron that the deviation from stoichiometry is the main cause for the differences observed between fresh UO2 and MOX [14,15], this effect is quantified in the next section. In the first model ("Model 1"), the effect of Pu is neglected over the entire relevant Pu compositions range (up to 24 wt.% PuO2), and a correlation obtained for non-stoichiometric homogeneous (U,Pu)O2 is used. In the second model ("Model 2", the effect of Pu is supposed to be present at all compositions, with the stoichiometry effect. The thermal conductivity is described by

  6. Development of fuel-model interfaces: Investigations by XPS, TEM, SEM and AFM

    Science.gov (United States)

    Stumpf, S.; Seibert, A.; Gouder, T.; Huber, F.; Wiss, T.; Römer, J.

    2009-03-01

    The presented work aims to reproducibly prepare UO 2-Pd thin film model systems for spent nuclear fuel in order to further investigate surface reactions of these films under relevant redox conditions. The sputter co-deposition of U and Pd (fission product) in the presence of O 2 results in the homogenous distribution of Pd in a crystalline UO 2 matrix. Heating the films causes the diffusion of film components. Hereby, the formation of ɛ-particles has to be clarified. First electrochemical studies show the influence of the nobel metal Pd on the redox behaviour of UO 2. With increasing Pd concentration the matrix dissolution is decreased. However, we could demonstrate that blocked oxidation processes are of temporary nature. The passivation of the Pd reactive sites with increasing number of cycles finally induces the approximation of the mixed system to the redox behaviour of the pure UO 2 system.

  7. Chemical speciation of uranium(VI) in marine environments: complexation of calcium and magnesium ions with [(UO2 )(CO3 )3 ](4-) and the effect on the extraction of uranium from seawater.

    Science.gov (United States)

    Endrizzi, Francesco; Rao, Linfeng

    2014-10-27

    The interactions of Ca(2+) and Mg(2+) with [UO2 (CO3 )3 ](4-) were studied by calcium ion selective electrode potentiometry and spectrophotometry. The stability constants of ternary Ca-UO2 -CO3 and Mg-UO2 -CO3 complexes were determined with calcium ion selective electrode potentiometry and optical absorption spectrophotometry, respectively. The enthalpies of complexation for two successive complexes, [CaUO2 (CO3 )3 ](2-) and [Ca2 UO2 (CO3 )3 ](aq), were determined for the first time by microcalorimetry. The data help to revise the speciation of uranium(VI) species under seawater conditions. In contrast to the previously accepted assumption that the highly negatively charged [UO2 (CO3 )3 ](4-) is the dominant species, the revised speciation indicates that the dominant aqueous uranium(VI) species under seawater conditions is the neutral [Ca2 UO2 (CO3 )3 ](aq). The results have a significant impact on the strategies for developing efficient sorption processes to extract uranium from seawater.

  8. A Highly Stable 3D Luminescent Indium-Polycarboxylic Framework for the Turn-off Detection of UO2(2+), Ru(3+), and Biomolecule Thiamines.

    Science.gov (United States)

    Du, Ning; Song, Jian; Li, Shuang; Chi, Yu-Xian; Bai, Feng-Ying; Xing, Yong-Heng

    2016-10-17

    Hydrothermal reaction of the multidentate organic ligand (H6TTHA) with indium chloride (InCl3) produced a highly stable 3D luminescent indium-organic framework [In2(OH)2(H2TTHA)(H2O)2]n (1). Complex 1 exhibits remarkable luminescent properties, especially the multifunction sensitivity and selectivity for detecting Ru(3+), UO2(2+); as well as small biomolecules thiamines (TPP, TMP, and TCl) based on a "turn-off" manner. In particular, the pyrophosphate groups of TPP and the phosphate groups of TMP could further affect the quenching rate, leading to different luminescent responds. In addition, we also discussed and proved the luminescence quenching mechanism in detail through comparative test and PXRD characterization. Therefore, complex 1 could be used as a kind of excellent luminescence sensor to detect Ru(3+), UO2(2+), and thiamines (TPP, TMP, and TCl).

  9. 外胶凝法制备高温气冷堆UO2核芯的湿法工艺%Wet Process of External Gelation of Uranium for Preparation of Uranium Dioxide Kernel of High Temperature Gas-cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    周湘文; 郝少昌; 赵兴宇; 马景陶; 王阳; 邓长生

    2012-01-01

    为制备高温气冷堆用燃料致密UO2核芯,对传统的溶胶-凝胶法进行优化和改进.主要对改进后的外胶凝工艺的湿法部分进行介绍,包括U3O8粉的溶解即欠酸硝酸铀酰(ADUN)溶液的制备、胶液的制备、胶液的分散和胶凝及凝胶球的陈化、洗涤和干燥等,并对湿法过程的机理进行了探讨.采用这一工艺,所得重铀酸铵微球的球形度好、尺寸分布均匀且具有良好空隙结构,经过后续的干法工艺如焙烧、还原和烧结,可制备出合格的高温气冷堆用燃料致密UO2核芯.%In order to prepare the dense uranium dioxide (UO2) kernel for high temperature gas-cooled reactors (HTGR) fuel, the conventional sol-gel processes are optimized and modified. The wet process of modified external gelation of uranium (EGU) is introduced, which includes the dissolution of U3Og, i.e. the preparation of acid-deficiency uranyl nitrate, preparation of broth solution, casting and gelation of broth solution and aging, washing and drying of the gelled spheres, and etc. The mechanism of wet process of EGU is also discussed. With the optimized wet process, the ammonium diuranate (ADU) microspheres with good sphericity, uniform diameter and perfect porous structure are prepared. After the subsequent treatments of dry processes such as calcination, reduction and sintering, the eligible dense UO2 kernels for HTGR fuel are manufactured.

  10. Electrochemical characterisation of CaCl2 deficient LiCl-KCl-CaCl2 eutectic melt and electro-deoxidation of solid UO2

    Science.gov (United States)

    Sri Maha Vishnu, D.; Sanil, N.; Mohandas, K. S.; Nagarajan, K.

    2016-03-01

    The CaCl2 deficient ternary eutectic melt LiCl-KCl-CaCl2 (50.5: 44.2: 5.3 mol %) was electrochemically characterised by cyclic voltammetry and polarization techniques in the context of its probable use as the electrolyte in the electrochemical reduction of solid UO2 to uranium metal. Tungsten (cathodic polarization) and graphite (anodic polarization) working electrodes were used in these studies carried out in the temperature range 623 K-923 K. The cathodic limit of the melt was observed to be set by the deposition of Ca2+ ions followed by Li+ ions on the tungsten electrode and the anodic limit by oxidation of chloride ions on the graphite electrode (chlorine evolution). The difference between the onset potential of deposition of Ca2+ and Li+ was found to be 0.241 V at a scan rate of 20 mV/s at 623 K and the difference decreased with increase in temperature and vanished at 923 K. Polarization measurements with stainless steel (SS) cathode and graphite anode at 673 K showed the possibility of low-energy reactions occurring on the UO2 electrode in the melt. UO2 pellets were cathodically polarized at 3.9 V for 25 h to test the feasibility of electro-reduction to uranium in the melt. The surface of the pellets was found reduced to U metal.

  11. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  12. Al2O3用于提取及分离硝酸铀酰溶液中钼和碘的研究%Study of Utilization Alumina for 99Mo and 131I Separation from UO2(NO3)2 Solution

    Institute of Scientific and Technical Information of China (English)

    邓启民; 程作用; 李茂良

    2011-01-01

    Alumina (A12O3) can be used for the extraction of MoO2-/4 and IO3/- from UO2(NO3)2 solu-Tion. IO3/-an be reduced to I-by using Sodium sulphite, then MoO4/2- and IO3/- can be separated with A12O3 for the unadsorption of I" on A12O3. The total recovery yield of Mo and I is 78.3%±7.00% and 87.9%±5.48% respectively, after MoO4/2- and IO3/- in UO2(NO3)2 solution were separated with two A12O3 columns, and Mo in iodide products solution and I in molybdenic products solution meet the requirements. The results indicate that A12O3 can be used for 99Mo and 131I separation from the fuel of Medical Isotope Production Reactor.%使用A12O3可以从硝酸铀酰的硝酸溶液[UO2(NO3)2 -HNO3]中同时提取MoO2-和IO3-.亚硫酸钠可以将IO3-还原成Iˉ,利用Al2O3吸附IO3-而不吸附Iˉ的性质,可将MoO2-4与IO3-分离.使用2根Al2O3柱提取并分离UO2(NO3)2中MoO2-4和IO3-后,Mo的平均回收率为78.3%±7.00%,钼产品中I的含量接近药典要求;I的平均回收率为87.9%±5.48%,碘产品中Mo的含量满足药用要求.实验结果表明,Al2O3可以用于医用同位素生产堆( MIPR)燃料溶液中的99Mo和131I的提取及分离.

  13. Experimental Results for SimFuels

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C.; Casella, Andrew M.; Skomurski, Frances N.; MacFarlan, Paul J.; Soderquist, Chuck Z.; Wittman, Richard S.; Mcnamara, Bruce K.

    2012-08-22

    Assessing the performance of Spent (or Used) Nuclear Fuel (UNF) in geological repository requires quantification of time-dependent phenomena that may influence its behavior on a time-scale up to millions of years. A high-level waste repository environment will be a dynamic redox system because of the time-dependent generation of radiolytic oxidants and reductants and the corrosion of Fe-bearing canister materials. One major difference between used fuel and natural analogues, including unirradiated UO2, is the intense radiolytic field. The radiation emitted by used fuel can produce radiolysis products in the presence of water vapor or a thin-film of water that may increase the waste form degradation rate and change radionuclide behavior. To study UNF, we have been working on producing synthetic UO2 ceramics, or SimFuels that can be used in testing and which will contain specific radionuclides or non-radioactive analogs so that we can test the impact of radiolysis on fuel corrosion without using actual spent fuel. Although, testing actual UNF would be ideal for understanding the long term behavior of UNF, it requires the use of hot cells and is extremely expensive. In this report, we discuss, factors influencing the preparation of SimFuels and the requirements for dopants to mimic the behavior of UNF. We have developed a reliable procedure for producing large grain UO2 at moderate temperatures. This process will be applied to a series of different formulations.

  14. On the mechanical stability of uranyl peroxide hydrates: Implications for nuclear fuel degradation

    Energy Technology Data Exchange (ETDEWEB)

    Weck, Philippe F.; Kim, Eunja; Buck, Edgar C.

    2015-09-11

    The mechanical properties and stability of studtite, (UO2)(O2)(H2O)2·2H2O, and metastudtite, (UO2)(O2)(H2O)2, two important corrosion phases observed on spent nuclear fuel exposed to water, have been investigated using density functional perturbation theory. While (UO2)(O2)(H2O)2 satisfies the necessary and sufficient Born criteria for mechanical stability, (UO2)(O2)(H2O)2·2H2O is found to be mechanically metastable, which might be the underlying cause of the irreversibility of the studtite to metastudtite transformation. According to Pugh’s and Poisson’s ratios and the Cauchy pressure, both phases are considered ductile and shear modulus is the parameter limiting their mechanical stability. Debye temperatures of 294 and 271 K are predicted for polycrystalline (UO2)(O2)(H2O)2·2H2O and (UO2)(O2)(H2O)2, suggesting a lower micro-hardness of metastudtite.

  15. The Welding Process of the Small In-pile Testing Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The small in-pile testing fuel assembly is designed for high performance fuel assembly study. It has two parts of which are four fuel element with double layer cladding and a detect system for measurement of testing pressure and temperature. The fuel element is composed of UO2 pellets, the stainless steel cladding and end caps. The detect system is direct contact with the fuel element by electron beam welding. In the fabrication of the assembly, some special welding technologies are

  16. Pulse irradiation tests of rock-like oxide fuel

    Science.gov (United States)

    Okonogi, K.; Nakamura, T.; Yoshinaga, M.; Ishijima, K.; Akie, H.; Takano, H.

    1999-08-01

    Pulse irradiation tests of special oxide fuel designed for plutonium disposal, called rock-like oxide (ROX), have been conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the transient behavior of ROX fuel under reactivity initiated accident (RIA) conditions. An uranium free ROX, (Zr,Y)O 2-MgAl 2O 4-PuO 2, is proposed for once-through use of Pu in light water reactors. However, because of smaller negative Doppler and void reactivity coefficients in the ROX fuel, higher peak fuel enthalpies are expected under RIAs than for UO 2 fuel. Thus, the tests of simulated ROX, in which Pu was replaced by U for easier realization, were conducted to a peak fuel enthalpy of 0.96 kJ g -1 (230 cal g -1), which is above current Japanese safety limits for UO 2. The transient behavior of the simulated ROX fuel was quite different from that of UO 2, because of its different thermo-physical properties. Fuel failure was associated with fuel melting at peak fuel enthalpies of 1.63 kJ g -1 (390 cal g -1) to 2.22 kJ g -1 (530 cal g -1). Significant mechanical energy generation, the reason for the limit, however, was not observed.

  17. Development of a PVC-membrane ion-selective bulk optode, for UO2(2+) ion, based on tri-n-octylphosphine oxide and dibenzoylmethane.

    Science.gov (United States)

    Shamsipur, Mojtaba; Tashkhourian, Javad; Sharghi, Hashem

    2005-06-01

    A novel uranyl ion-selective bulk optode membrane, incorporating tri-n-octylphosphine oxide for cation recognition and a lipophilic chromoionophore dibenzoylmethane, has been prepared. The PVC membrane composition was optimized to result in the widest working concentration range. The response range of the proposed optode is 4.1 x 10(-6) to 2.0 x 10(-4) mol L(-1) UO2(2+). The probe works at pH 4.0. Ion interference is low and selectivity, reproducibility, and stability are good.

  18. Qualification of pebble fuel for HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich (Germany). IEK-6; Allelein, Hans-Josef [Forschungszentrum Juelich (Germany). IEK-6; RWTH Aachen (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik (LRST)

    2016-05-15

    The German HTGR fuel development program for the HTR-Modul concept has resulted in a reference design based on LEU UO2 TRISO coated particle fuel in a spherical fuel element. The coated particles consist of minute uranium particle kernels coated with layers of carbon and silicon carbide. Analyses on quality of as-manufactured fuel, its behavior under HTR-Modul relevant operating and accident conditions have demonstrated excellent performance. Coated particles can withstand high internal gas pressure without releasing their fission products to the environment. International efforts are on-going for further improvement of coated particle fuel to meet the needs of future generation-IV HTR concepts.

  19. 罗丹明B试法鉴定UO2+2用N-烯丙基-N′-(对苯磺钠)硫脲掩蔽Fe3+的干扰%Eliminating the interference of Fe3+ with the N-allyl-N′-(sodium-p-benzenesulfonate) thiourea in the indentification of UO2+2 using Rodamine B

    Institute of Scientific and Technical Information of China (English)

    马万山; 许春萱; 钟瑞琴; 张秀兰

    2002-01-01

    以实验为基础,提出了用罗丹明B试法鉴定UO2+2时,用N-烯丙基-N′-(对苯磺酸钠)硫脲掩蔽Fe3+离子的干扰.与传统的消除干扰的方法相比,具有条件容易控制,操作简便等优点.

  20. A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions

    Science.gov (United States)

    Le Saux, M.; Besson, J.; Carassou, S.; Poussard, C.; Averty, X.

    2008-08-01

    This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10×1025 nm or burnup up to 64 GWd/tU) material within large temperature (from 20 °C up to 1100 °C) and strain rate ranges (from 3×10-4 s up to 5 s), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program.

  1. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  2. Preparation of UO2, ThO2 and (Th,U)O2 pellets from photochemically-prepared nano-powders

    Science.gov (United States)

    Pavelková, Tereza; Čuba, Václav; de Visser-Týnová, Eva; Ekberg, Christian; Persson, Ingmar

    2016-02-01

    Photochemically-induced preparation of nano-powders of crystalline uranium and/or thorium oxides and their subsequent pelletizing has been investigated. The preparative method was based on the photochemically induced formation of amorphous solid precursors in aqueous solution containing uranyl and/or thorium nitrate and ammonium formate. The EXAFS analyses of the precursors shown that photon irradiation of thorium containing solutions yields a compound with little long-range order but likely "ThO2 like" and the irradiation of uranium containing solutions yields the mixture of U(IV) and U(VI) compounds. The U-containing precursors were carbon free, thus allowing direct heat treatment in reducing atmosphere without pre-treatment in the air. Subsequent heat treatment of amorphous solid precursors at 300-550 °C yielded nano-crystalline UO2, ThO2 or solid (Th,U)O2 solutions with high purity, well-developed crystals with linear crystallite size <15 nm. The prepared nano-powders of crystalline oxides were pelletized without any binder (pressure 500 MPa), the green pellets were subsequently sintered at 1300 °C under an Ar:H2 (20:1) mixture (UO2 and (Th,U)O2 pellets) or at 1600 °C in ambient air (ThO2 pellets). The theoretical density of the sintered pellets varied from 91 to 97%.

  3. Charge Distribution and Local Structure and Speciation in the UO2+x and PuO2+x Binary Oxides for x <= 0.25

    Energy Technology Data Exchange (ETDEWEB)

    Conradson, Steven D.; Begg, Bruce D.; Clark, David L.; Den Auwer, Christophe J.; Ding, Mei; Dorhout, Peter K.; Espinosa-Faller, Francisco J.; Gordon, Pamela L.; Haire, Richard G.; Hess, Nancy J.; Hess, Ryan F.; Keogh, D. Webster; Lander, Gerard H.; Manara, Dario; Morales, Luis A.; Neu, Mary P.; Paviet-Hartmann, Patricia; Rebizant, Jean; Rondinella, Vincenzo V.; Runde, Wolfgang; Tait, C DREW.; Veirs, D. Kirk; Villella, Phillip M.; Wastin, Franck

    2005-02-01

    The local structure and chemical speciation of the mixed valence, fluorite-based oxides UO2+x (0.00pxp0.20) and PuO2+x/PuO2+x*y(OH)2y * zH2O have been determined by U/Pu LIII XAFS spectroscopy. The U spectra indicate (1) that the O atoms are incorporated as oxo groups at short (1.75A ? ) U?O distances consistent with U(VI) concomitant with a large range of U displacements that reduce the apparent number of U neighbors and (2) that the UO2 fraction remains intact implying that these O defects interact to form clusters and give the heterogeneous structure consistent with the diffraction patterns. The PuO2+x system, which does not show a separate phase at its x ? 0:25 endpoint, also displays (1) oxo groups at longer 1.9A ? distances consistent with Pu(V+d), (2) a multisite Pu?O distribution even when x is close to zero indicative of the formation of stable species with H2O and its hydrolysis products with O2*, and (3) a highly disordered, spectroscopically invisible Pu?Pu component. The structure and bonding in AnO2+x are therefore more complicated than have previously been assumed and show both similarities but also distinct differences among the different elements.

  4. Application of a boron doped diamond (BDD) electrode as an anode for the electrolytic reduction of UO2 in Li2O-LiCl-KCl molten salt

    Science.gov (United States)

    Park, Wooshin; Kim, Jong-Kook; Hur, Jin-Mok; Choi, Eun-Young; Im, Hun Suk; Hong, Sun-Seok

    2013-01-01

    A boron doped diamond thin film electrode was employed as an inert anode to replace a platinum electrode in a conventional electrolytic reduction process for UO2 reduction in Li2O-LiCl molten salt at 650 °C. The molten salt was changed into Li2O-LiCl-KCl to decrease the operation temperature to 550 °C at which the boron doped diamond was chemically stable. The potential for oxygen evolution on the boron doped diamond electrode was determined to be approximately 2.2 V vs. a Li-Pb reference electrode whereas that for Li deposition was around -0.58 V. The density of the anodic current was low compared to that of the cathodic current. Thus the potential of the cathode might not reach the potential for Li deposition if the surface area of the cathode is too wide compared to that of the anode. Therefore, the ratio of the surface areas of the cathode and anode should be precisely controlled. Because the reduction of UO2 is dependent on the reaction with Li, the deposition of Li is a prerequisite in the reduction process. In a consecutive reduction run, it was proved that the boron doped diamond could be employed as an inert anode.

  5. 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.

  6. Morphology change of rock-like oxide fuels in reactivity-initiated-accident simulation tests

    Science.gov (United States)

    Nakamura, T.; Sasajima, H.; Yamashita, T.; Uetsuka, H.

    2003-06-01

    Pulse irradiation tests under simulated reactivity-initiated accident (RIA) conditions were performed with three types of rock-like oxide (ROX) fuels. Single phase yttria stabilized zirconia (YSZ), homogeneous mixture of YSZ/spinel and YSZ particle dispersed in spinel type ROX fuels were pulse irradiated in the Nuclear Safety Research Reactor (NSRR). Mode and threshold of the fuel rod failure including its consequences were investigated under the RIA conditions. The cladding failure occurred in a burst type mode in all the three types of ROX fuel tests with considerable fuel melting. Even though the mode was quite different from those of UO 2 fuel, failure threshold enthalpies of the ROX fuels were close to that of UO 2 fuel at about 10 GJ m -3. The consequence of the failure of the ROX fuels rods was different from the one of UO 2 fuel rods, because molten fuel dispersal occurred at lower enthalpies in the ROX fuel tests. Change of the fuel structure and material interaction in the transient heating conditions were examined through optical and secondary electron microscopy, and electron probe micro analysis.

  7. Advances in the Development of a WCl6 CVD System for Coating UO2 Powders with Tungsten

    Science.gov (United States)

    Mireles, Omar R.; Tieman, Alyssa; Broadway, Jeramie; Hickman, Robert

    2013-01-01

    Demonstrated viability and utilization of: a) Fluidized powder bed. b) WCl6 CVD process. c) Coated spherical particles with tungsten. The highly corrosive nature of the WCl6 solid reagent limits material of construction. Indications that identifying optimized process variables with require substantial effort and will likely vary with changes in fuel requirements.

  8. Comparison of optimised germanium gamma spectrometry and multicollector inductively coupled plasma mass spectrometry for the determination of 134Cs, 137Cs and 154Eu single ratios in highly burnt UO 2

    Science.gov (United States)

    Caruso, S.; Günther-Leopold, I.; Murphy, M. F.; Jatuff, F.; Chawla, R.

    2008-05-01

    Non-destructive and destructive methods have been compared to validate their corresponding assessed accuracies in the measurement of 134Cs/137Cs and 154Eu/137Cs isotopic concentration ratios in four spent UO2 fuel samples with very high (52 and 71 GWd/t) and ultra-high (91 and 126 GWd/t) burnup values, and about 10 (in the first three samples) and 4 years (in the latter sample) cooling time. The non-destructive technique tested was high-resolution gamma spectrometry using a high-purity germanium detector (HPGe) and a special tomographic station for the handling of highly radioactive 400 mm spent fuel segments that included a tungsten collimator, lead filter (to enhance the signal to Compton background ratio and reduce the dead time) and paraffin wax (to reduce neutron damage). The non-destructive determination of these isotopic concentration ratios has been particularly challenging for these segments because of the need to properly derive non-Gaussian gamma-peak areas and subtract the background from perturbing capture gammas produced by the intrinsic high-intensity neutron emissions from the spent fuel. Additionally, the activity distribution within each pin was determined tomographically to correct appropriately for self-attenuation and geometrical effects. The ratios obtained non-destructively showed a 1σ statistical error in the range 1.9-2.9%. The destructive technique used was a high-performance liquid chromatographic separation system, combined online to a multicollector inductively coupled plasma mass spectrometer (HPLC-MC-ICP-MS), for the analysis of dissolved fuel solutions. During the mass spectrometric analyses, special care was taken in the optimisation of the chromatographic separation for Eu and the interfering element Gd, as also in the mathematical correction of the 154Gd background from the 154Eu signal. The ratios obtained destructively are considerably more precise (1σ statistical error in the range 0.4-0.8% for most of the samples, but up to

  9. Modeling and Simulation of Nuclear Fuel Materials

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Van Brutzel, Laurent; Chartier, Alan; Gueneau, Christine; Mattsson, Ann E.; Tikare, Veena; Bartel, Timothy; Besmann, T. M.; Stan, Marius; Van Uffelen, Paul

    2010-10-01

    We review the state of modeling and simulation of nuclear fuels with emphasis on the most widely used nuclear fuel, UO2. The hierarchical scheme presented represents a science-based approach to modeling nuclear fuels by progressively passing information in several stages from ab initio to continuum levels. Such an approach is essential to overcome the challenges posed by radioactive materials handling, experimental limitations in modeling extreme conditions and accident scenarios, and the small time and distance scales of fundamental defect processes. When used in conjunction with experimental validation, this multiscale modeling scheme can provide valuable guidance to development of fuel for advanced reactors to meet rising global energy demand.

  10. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  11. Study of Reduced-Enrichment Uranium Fuel Possibility for Research Reactors

    Directory of Open Access Journals (Sweden)

    Ruppel V.A.

    2015-01-01

    Full Text Available Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5% for UO2 fuel and by 7% for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4% for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1% less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

  12. 三丁基氧化膦-离子液体体系萃取UO2(NO3)2的机理和选择性%Extraction Mechanism and Selectivity of UO2(NO3)2 in Tributylphosphine Oxide-Ionic Liquid System

    Institute of Scientific and Technical Information of China (English)

    刘海望; 沈兴海; 陈庆德

    2015-01-01

    研究了三辛基氧化膦(TOPO)和三丁基氧化膦(TBPO)在离子液体(ILs)1-烷基-3-甲基咪唑双三氟甲基磺酰亚胺盐(CnmimNTf2, n=2,4,6,8)中萃取分离UO2(NO3)2. TOPO-C2mimNTf2和TOPO-C4mimNTf2体系萃取UO2(NO3)2时会出现三相,而TBPO萃取UO2(NO3)2的萃合物可以很好地溶解在所有离子液体中.论文也考察了萃取过程中的萃取剂浓度效应、酸效应、盐效应.水相加入HNO3会降低萃取效率.盐效应证明了萃取是一种阳离子交换机理.水相中加入NO3-能够提高U的萃取,这说明NO3-参与萃取.选择性研究表明:除了在高酸度下对Zr的显著萃取, TBPO-C4mimNTf2萃取体系在低酸度下对U呈现较好的选择性;去除U后,在低酸度下该体系对三价Nd仍保持较好的选择性.通过定量比较离子液体中NO3-进入量,电喷雾质谱(ESI-MS)和紫外光谱表征确定了TBPO-CnmimNTf2中萃取机理的差异性.萃取中存在两种萃合物,即UO2(TBPO)3(NO3)+和UO2(TBPO)32+,其中UO2(TBPO)3(NO3)+的比例从C2mimNTf2体系到C8mimNTf2体系逐渐增加.%The extraction of UO2(NO3)2 from aqueous solution was investigated using trioctylphosphine oxide (TOPO) and tributylphosphine oxide (TBPO) in ionic liquids (ILs) (CnmimNTf2, n=2, 4, 6, 8). A third phase was formed in the TOPO-C2mimNTf2 and TOPO-C4mimNTf2 extraction systems, whereas the extracted species of TBPO-CnmimNTf2 (n=2, 4, 6, 8) were wel soluble in al ILs. The influence of the concentrations of the extractant, nitric acid, and salt on the extraction efficiency was also investigated. Adding HNO3 to the aqueous phase decreased the extraction efficiency. The effect of salt indicates the presence of a cation-exchange mechanism in the extraction. The addition of NO3-in the aqueous phase increased the extraction efficiency of U, which indicates that NO3-participates in the extraction. Selective extraction research indicates that TBPO-C4mimNTf2 exhibits good selectivity for U at low acid concentration despite

  13. NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

    2012-01-01

    This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature

  14. Comparative studies of mononuclear Ni(II) and UO2(II) complexes having bifunctional coordinated groups: synthesis, thermal analysis, X-ray diffraction, surface morphology studies and biological evaluation.

    Science.gov (United States)

    Fahem, Abeer A

    2012-03-01

    Two Schiff base ligands derived from condensation of phthalaldehyde and o-phenylenediamine in 1:2 (L(1)) and 2:1 (L(2)) having bifunctional coordinated groups (NH(2) and CHO groups, respectively) and their metal complexes with Ni(II) and UO(2)(II) have been synthesized and characterized by elemental analysis, molar conductance, magnetic susceptibilities and spectral data (IR, (1)H NMR, mass and solid reflectance) as well as thermal, XRPD and SEM analysis. The formula [Ni(L(1))Cl(2)]·2.5H(2)O, [UO(2)(L(1))(NO(3))(2)]·2H(2)O, [Ni(L(2))Cl(2)]·1.5H(2)O and [UO(2)(L(2))(NO(3))(2)] have been suggested for the complexes. The vibrational spectral data show that the ligands behave as neutral ligands and coordinated to the metal ions in a tetradentate manner. The Ni(II) complexes are six coordinate with octahedral geometry and the ligand field parameters: D(q), B, β and LFSE were calculated while, UO(2)(II) complexes are eight coordinate with dodecahedral geometry and the force constant, F(U-O) and bond length, R(U-O) were calculated. The thermal decomposition of complexes ended with metal chloride/nitrate as a final product and the highest thermal stability is displayed by [UO(2)(L(2))(NO(3))(2)] complex. The X-ray powder diffraction data revealed the formation of nano sized crystalline complexes. The SEM analysis provides the morphology of the synthesized compounds and SEM image of [UO(2)(L(2))(NO(3))(2)] complex exhibits nano rod structure. The growth-inhibiting potential of the ligands and their complexes has been assessed against a variety of bacterial and fungal strains.

  15. Determination of in-depth damaged profile by Raman line scan in a pre-cut He2+ irradiated UO2

    Science.gov (United States)

    Guimbretière, G.; Desgranges, L.; Canizarès, A.; Carlot, G.; Caraballo, R.; Jégou, C.; Duval, F.; Raimboux, N.; Ammar, M. R.; Simon, P.

    2012-06-01

    Raman measurements were carried out to probe the spectroscopic signatures of the ion beam irradiation-induced damage and their in-depth profiles on a Uranium dioxide sample previously cut and polished prior to performing a 25 MeV He2+ cyclotron beam irradiation. Raman spectra clearly show the creation of three defects bands (U1 ≈ 530, U2 ≈ 575, and U3 ≈ 635 cm-1) resulting from the ion irradiation and also some changes in the T2g peak of UO2. Their in-depth distribution inside the sample exhibits a clear increase of the damage from the surface up to the position of the implanted He.

  16. Research Progress About Gas-Exhaust-Device for Fuel Element

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wu-ye

    2012-01-01

    <正>UO2-x stack applied in the fuel element has a form of a cylinder with a central hole, where temperature field characterized by high temperature and high gradient is formed due to irradiation. Then nearly all of the gaseous fission products (GFPs) can release into central cavity. However, uranium oxide will evaporate form the fuel stack’s inner surface because of its high temperature (about 1 800-2 000 ℃),

  17. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    Science.gov (United States)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  18. UO2粉末表面活化壳层的制备和性能研究%Preparation and properties of uranium dioxide powders with active surface layer

    Institute of Scientific and Technical Information of China (English)

    高家诚; 吴曙芳; 杨晓东; 李锐; 王勇

    2010-01-01

    UO2粉末表面活化壳层的制备和性能进行了试验研究.用正交试验研究了影响其比表面积的因素,用DSC-TG、XRD、BET等分析了粉末的氧化过程、组织结构和使用性能.结果表明,优化的制备工艺是330℃×8h,18%O2+N2.影响UO2粉末比表面积的主要因素为氧化气氛和氧化温度.分析发现,UO2粉末在240℃氧化8h后有极少量的U3O7生成,382℃和815℃下氧化8h后生成U3O8.因此,实现UO2低温烧结的最佳O与U原子数比的粉末预氧化工艺温度在240~370℃之间.

  19. Energy Frontier Research Center Center for Materials Science of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd Allen

    2014-04-01

    Scientific Successes • The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative, experimental-based anharmonic smoothing technique has enabled quantitative benchmarking of ab initio PDOS simulations. • Direct comparison between anharmonicity-smoothed ab initio PDOS simulations for UO2 and experimental measurements has demonstrated the need for improved understanding of UO2 at the level of phonon dispersion, and, further, that advanced lattice dynamics simulations including finite temperatures approaches will be required for handling this strongly correlated nuclear fuel. • PDOS measurements performed on polycrystalline samples have identified the phonon branches and energy ranges most highly impacted by fission-product and hyper-stoichiometry lattice defects in UO2. These measurements have revealed the broad-spectrum impact of oxygen hyper-stoichiometry on thermal transport. The reduction in thermal conductivity caused by hyper-stoichiometry is many times stronger than that caused by substitutional fission-product impurities. • Laser-based thermo-reflectance measurements on UO2 samples irradiated with light (i.e. He) ions to introduce point defects have been coupled with MD simulations and lattice parameter measurements to determine the role of uranium and oxygen point defects in reducing thermal conductivity. • A rigorous perturbation theory treatment of phonon lifetimes in UO2 based on a 3D discretization of the Brillouin zone coupled with experimentally measured phonon dispersion has been implemented that produces improved predictions of the temperature dependent thermal conductivity. • Atom probe investigations of the influence of grain boundary structure on the segregation behavior of Kr in UO2 have shown that smaller amounts of Kr are present at low angle grain boundaries than at large angle grain

  20. Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Robert T [ORNL; Collins, Jack Lee [ORNL; Hunt, Rodney Dale [ORNL; Ladd-Lively, Jennifer L [ORNL; Patton, Kaara K [ORNL; Hickman, Robert [NASA Marshall Space Flight Center, Huntsville, AL

    2014-01-01

    The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

  1. 照射済みUO2ペレットの加熱による炉外EP放出実験

    OpenAIRE

    石渡 名澄; 永井 斉

    1985-01-01

    LWRの燃料損傷事故条件下での燃料からのFP放出割合については、NUREG-0772において貝体的な数値データが提出された。上出の数値データを評価するため、相対的に小規模の実験装置を用いる測定方法を開発した。1500circC以上の温度範囲において、燃料からのFPのCsの放出割合は相対的に大きいので、高周波誘導加熱炉を含む実験装置を用いて、照射済みUO2ペレットからの137Csの放出割合を測定した。照射済みUO2ペレットはNSRR及びJMTR-RABBITを用いて製作した。加熱実験において、137Csの放出割合は、NSRR照射のペレットでは0.51(Ar、12.2分加熱、1500~2080circC)、RABBIT照射のペレットでは、それぞれに0.63、0.59、0.81及び0.78(Ar、10.7分加熱、1500~1740circC;Ar、32.8分加熱、1500~2255circC;Ar+蒸気、22.0分加熱、1500~2230circC;Ar+蒸気+H2、14.0分加熱、1500~2030circC)であった。...

  2. Numerical design of the Seed-Blanket Unit for the thorium nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,UO2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233U from fertile 232Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233U, 235U and 232Th.

  3. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels are irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.

  4. Screening of advanced cladding materials and UN-U3Si5 fuel

    Science.gov (United States)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  5. 半衰期为数天的惰性气体和卤素裂变产物在UO2颗粒中的扩散与释放%Diffusion and Release of Noble Gas and Halogen Fission Products With Several Days Half-Life in UO2 Particle

    Institute of Scientific and Technical Information of China (English)

    房超

    2013-01-01

    在考虑吸附效应等物理过程的基础上,得到了裂变产物在UO2颗粒中扩散与释放模型的严格解,并导出了不同反应堆运行状态下裂变产物累积释放份额F(t)、释放-产出比R(t)/B(t)的严格表达式.利用上述结果以及相应的近似解、数值解,对半衰期为数天的惰性气体和卤素裂变产物(131I、131Xem、133Xe和133Xem)在不同堆芯历史条件下的F(t)和R(t)/B(t)进行了比较计算.分析表明,F(t)与R(t)/B(t)的结果均有所差别,但当反应堆运行时间达一定长度后,它们的数值相等.此外,严格解去掉了近似解中不必要的保守性,也比数值解更符合物理实际.%The exact solutions of diffusion and release model of noble gas and halogen fission products in UO2 particle of HTGR were built under the conditions of adsorption effect and other physical processes. The corresponding release fractions (F(t)) and the ratio of release and productive amounts (R (t)/B (t)) of fission products were also derived. Furthermore, the F(t) and R(t)/B(t) of 131I, 131Xem, 133Xe and 133Xem whose half-lifes are several days in UO2 particle with the exact solutions, approximate solutions and corresponding numerical solutions under different temperature histories of reactor core were investigated. The results show that the F(t) and R(t)/B(t) are different in numerical values unless the time of release is long enough. The properties of conservation of exact solutions are much more reasonable than the ones of approximate solutions. It is also found that the results of exact solutions approach the actual working conditions more than the approximate and numerical solutions.

  6. On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Nilsson, Sara; Jonsson, Mats

    2008-08-01

    Radiation induced oxidative dissolution of UO 2 is a key process for the safety assessment of future geological repositories for spent nuclear fuel. This process is expected to govern the rate of radionuclide release to the biosphere. In this work, we have studied the catalytic effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel. The experimental studies were performed using UO 2 pellets containing 0%, 0.1%, 1% and 3% Pd as a model for spent nuclear fuel. H 2O 2 was used as a model for radiolytical oxidants (previous studies have shown that H 2O 2 is the most important oxidant in such systems). The pellets were immersed in aqueous solution containing H 2O 2 and HCO3- and the consumption of H 2O 2 and the dissolution of uranium were analyzed as a function of H 2 pressure (0-40 bar). The noble metal inclusions were found to catalyze oxidation of UO 2 as well as reduction of surface bound oxidized UO 2 by H 2. In both cases the rate of the process increases with increasing Pd content. The reduction process was found to be close to diffusion controlled. This process can fully account for the inhibiting effect of H 2 observed in several studies on spent nuclear fuel dissolution.

  7. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  8. Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, S. J. [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2015-10-15

    FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation.

  9. Failure behavior of plutonium-uranium mixed oxide fuel under reactivity-initiated accident condition

    Science.gov (United States)

    Abe, T.; Nakae, N.; Kodato, K.; Matsumoto, M.; Inabe, T.

    1992-06-01

    Two series of in-pile tests on MOX fuels were performed in the NSRR to study failure behavior under RIA (reactivity-initiated accident) conditions in water cooled reactors. PWR type MOX test rods were pulsed in a first series. The test rods were designed to have dimensions identical to standard UO 2 fuel, on which a large number of tests had been conducted previously. The test result was that the failure mechanism and the threshold of MOX fuel was consistent with those of UO 2 fuel. ATR-type MOX test rods with PuO 2 particles as well as reference rods without PuO 2 particles were subjected to pulsing in a second series. PuO 2 particles of 400 and 1100 μm in diameter were artificially embedded at the surface of MOX pellets. No effect of particles appeared on the threshold, and no significant indication of their effect was observed on the cladding.

  10. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  11. Development of the fabrication technology of the simulated fuel-I, 15,000MWd/tU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, D. J.; Kim, H. S.; Lee, J. W.; Yang, M. S

    2001-04-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties, fission gas release, grain growth and et al. of the DUPIC fuel is different from the commercial UO2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, the sintering characterization of wet milled powder for 24 hours to fabricate 15,000MWd/tU equivalent burnup simulated fuel.

  12. Effect of reactor radiation on the thermal conductivity of TREAT fuel

    Science.gov (United States)

    Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

    2017-04-01

    The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

  13. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    OpenAIRE

    Hyun-Gil Kim; Jae-Ho Yang; Weon-Ju Kim; Yang-Hyun Koo

    2016-01-01

    For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density compo...

  14. Characterization of (Th,U)O 2 fuel pellets made by impregnation technique

    Science.gov (United States)

    Kutty, T. R. G.; Nair, M. R.; Sengupta, P.; Basak, U.; Kumar, Arun; Kamath, H. S.

    2008-02-01

    Impregnation technique is an attractive alternative for manufacturing highly radiotoxic 233U bearing thoria based mixed oxide fuel pellets, which are remotely treated in hot cell or shielded glove-box facilities. This technique is being investigated to fabricate the fuel for the forthcoming Indian Advanced Heavy Water Reactor (AHWR). In the impregnation process, porous ThO 2 pellets are prepared in an unshielded facility which are then impregnated with 1.5 molar uranyl nitrate solution in a shielded facility. The resulting composites are dried and denitrated at 500 °C and then sintered in reducing/oxidizing atmosphere to obtain high density (Th,U)O 2 pellets. In this work, the densification behaviour of ThO 2-2% UO 2 and ThO 2-4% UO 2 pellets was studied in reducing and oxidizing atmospheres using a high temperature dilatometer. Densification was found to be larger in air than in Ar-8% H 2. The characterization of the sintered pellets was made by optical microscopy, scanning electron microscopy (SEM) and electron probe microanalysis (EPMA). The grain structure of ThO 2-2% UO 2 and ThO 2-4% UO 2 pellets was uniform. The EPMA data confirmed that the uranium concentration was slightly higher at the periphery of the pellet than that at the centre.

  15. Formula Selection for Dairy Fresh Manure Transformed into Civilian Fuel Materials%牛鲜粪便转化民用燃料原料配方的筛选

    Institute of Scientific and Technical Information of China (English)

    李广有

    2012-01-01

    养牛业已成为我国农业经济发展和畜牧业生产中的一个重要支柱产业,规模集约化发展的同时所产生大量粪便等污物得不到有效环保处理,对环境造成了严重污染。本文采用正交试验,将奶牛鲜粪便、煤粉、除臭吸附粘合剂原料均匀混合处理,对牛鲜粪便转化民用燃料技术环保再利用进行了研究,结果表明,含水量较高的奶牛鲜粪便最低利用率可达40%,除臭吸附粘合剂最适合添加比例为11%,就可及时将牛鲜粪便中大量粪水和挥发性污染大气有害气体吸附固定住,也便于加工成多种形状的民用燃料加以再利用,具有强大的环保功能和鲜粪便循环再利用特点,对推动养牛业健康可持续发展具有重大意义。%Cattle industry has become one of important pillar industry in China's rural economy development. Animal husbandry and scale of the intensive development produce the large amount of waste and sewage, the environment affected by the serious pollution. The orthogonal test was conduct in this study. The cow fresh manure, pulverized coal deodorant adsorption bond raw material processing, mixing of dairy fresh manure into civil fuel technology environmental protection reuse was studied, the result showed that the minimum ratio of fresh manure with higher water content can reach 40%, the most suitable adding ratio for deodorant adsorption adhesive was 11 %, the fresh manure in feces and volatile atmospheric pollution harmful gas adsorption can be fixed and facilitate processed into various shape of civil fuel to recycle, so it has strong environmental protection function and give important significance to dairy in dustry.

  16. 牛鲜粪转化为民用燃料的原料配方筛选%The Screening of for Formula of Fresh Faces of Dairy Cow Transformation into Civil Fuel Materials

    Institute of Scientific and Technical Information of China (English)

    刘明生; 李广有

    2012-01-01

    为开发一种实用、便于普及的鲜牛粪环保处理新技术,明确原料最佳配合比例,实现其资源循环再利用.采用正交试验,把奶牛鲜粪、煤粉、除臭吸附粘合剂为原料均匀混合处理,对牛鲜粪转化民用燃料技术的环保再利用进行了研究.结果表明:含水量较高的奶牛鲜粪最低利用比率可达40%,除臭吸附粘合剂最适合添加比率为11%时,将牛鲜粪中大量粪水和挥发性污染大气有害气体吸附固定,便于加工成多种形状的民用燃料加以再利用,具有强大的环保功能和鲜粪便循环再利用特点,对推动养牛业健康可持续发展具有重大意义.%In order to development a practical and convenient popularization fresh cow dung environmental protection treatment of new technologies, to make clear material optimal proportion, realization of the resource recycling. Blending the fresh cattle feces, pulverized and deodorant adsorption adhesive in the orthogonal test, the mixing environmental protection reuse transformation technology that transformation fresh cattle feces into civil fuel was studied. The studying showed that the using ratio of higher water content fresh cattle feces could reached at least 40%. The most suitable adding ration of deodorant adsorption adhesive was 11%, which could timely adsorbed the water in the feces and volatile atmospheric that would harmfully polluted the environment. The method facilitates the procession and recycle of multi-shape civil fuel, had strong environmental protection function and recycle feature which would be important to the sustain development of the cattle industry.

  17. Updated NGNP Fuel Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    . • Additional funding will be made available beginning in fiscal year (FY) 2012 to support pebble bed fuel fabrication process development and fuel testing while maintaining the prismatic fuel schedule. Options for fuel fabrication for prismatic and pebble bed were evaluated based on the credibility of each option, along with a cost and schedule to implement each strategy. The sole prismatic option is Babcock and Wilcox (B&W) producing uranium oxycarbide (UCO) tristructural-isotropic (TRISO) fuel particles in compacts. This option finishes in the middle of 2022 . Options for the pebble bed are Nuclear Fuel Industries (NFI) in Japan producing uranium dioxide (UO2) TRISO fuel particles, and/or B&W producing UCO or UO2 TRISO fuel particles. All pebble options finish in mid to late 2022.

  18. 奶牛鲜粪便转化民用燃料原料配方的筛选%Formula Screening of Civil Fuel Material Converted from Fresh Feces of Dairy Cow

    Institute of Scientific and Technical Information of China (English)

    李广有

    2011-01-01

    Adopting orthogonal test, the cow's fresh dung, pulverized coal and deodorant absorption bond raw material were mixed together so as to screen the formula for civil fuel material. The results show that the fresh feces of diary cow with higher water content had a lowest utilization rate of 40% , the most proper additive rate of deodorant absorption bond was 11%, at this point it can absorb and fix the toxic gas in feces, and easily be made into various civil fuel material, which will be recycled and enviromental friendly. It also has a significant meaning to the sustainable development of cattle industry.%采用正交试验,将奶牛鲜粪便、煤粉、除臭吸附粘合剂原料均匀混合处理,对牛鲜粪便转化民用燃料的原料配方进行了筛选研究.结果表明:含水量较高的奶牛鲜粪便最低利用比率可达40%,除臭吸附粘合剂最适合添加比率为11%,就可及时将牛鲜粪便中大量粪水和挥发性污染大气有害气体吸附固定住,也便于加工成多种形状的民用燃料加以再利用,具有强大的环保功能和鲜粪便循环再利用特点,对推动养牛业健康可持续发展具有重大意义.

  19. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Yang Zhong; Robert C. O' Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

    2011-11-01

    The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

  20. The high burn-up structure in nuclear fuel

    Directory of Open Access Journals (Sweden)

    Vincenzo V. Rondinella

    2010-12-01

    Full Text Available During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

  1. Laser pulse heating of nuclear fuels for simulation of reactor power transients

    Indian Academy of Sciences (India)

    C S Viswanadham; K C Sahoo; T R G Kutty; K B Khan; V P Jathar; S Anantharaman; Arun Kumar; G K Dey

    2010-12-01

    It is important to study the behaviour of nuclear fuels under transient heating conditions from the point of view of nuclear safety. To simulate the transient heating conditions occurring in the known reactor accidents like loss of coolant accident (LOCA) and reactivity initiated accident (RIA), a laser pulse heating system is under development at BARC, Mumbai. As a prelude to work on irradiated nuclear fuel specimens, pilot studies on unirradiated UO2 fuel specimens were carried out. A laser pulse was used to heat specimens of UO2 held inside a chamber with an optically transparent glass window. Later, these specimens were analysed by metallography and X-ray diffraction. This paper describes the results of these studies.

  2. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Science.gov (United States)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50-60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

  3. ENUSA-TECNATOM collaboration project: improvements to the system of inspection by UT's circular fresh fuel rod welding; Proyecto colaboraci0n ENUSA-TECNATOM: Mejoras en el sistema de inspeccion por UT de la soldadura circular de la barra combustible fresca

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J.; Toral, M.; Moraleda, J.; Quinones, D.

    2014-07-01

    Enusa and Tecnatom have embarked on a road of technological and commercial collaboration that aims to firstly, the continuous improvement of the means of production of fuel from the factory in Juzbado, but uses the joint technological capital to diversify their business global opportunities. This collaboration has emerged a new line for control by UT of welding circular fresh fuel rod and the development of an equipment for sale to the CINF in Yibin fuel factory. The characteristics of these projects are presented in this paper. (Author)

  4. Synthesis, characterization and biological activities of Cu(II), Co(II), Mn(II), Fe(II), and UO2(VI) complexes with a new Schiff Base hydrazone: O-hydroxyacetophenone-7-chloro-4-quinoline hydrazone.

    Science.gov (United States)

    Al-Shaalan, Nora H

    2011-10-13

    The Schiff base hydrazone ligand HL was prepared by the condensation reaction of 7-chloro-4-quinoline with o-hydroxyacetophenone. The ligand behaves either as monobasic bidentate or dibasic tridentate and contain ONN coordination sites. This was accounted for be the presence in the ligand of a phenolic azomethine and imine groups. It reacts with Cu(II), Ni(II), Co(II), Mn(II), UO(2) (VI) and Fe(II) to form either mono- or binuclear complexes. The ligand and its metal complexes were characterized by elemental analyses, IR, NMR, Mass, and UV-Visible spectra. The magnetic moments and electrical conductance of the complexes were also determined. The Co(II), Ni(II) and UO(2) (VI) complexes are mononuclear and coordinated to NO sites of two ligand molecules. The Cu(II) complex has a square-planar geometry distorted towards tetrahedral, the Ni(II) complex is octahedral while the UO(2) (VI) complex has its favoured heptacoordination. The Co(II), Mn(II) complexes and also other Ni(II) and Fe(III) complexes, which were obtained in the presence of Li(OH) as deprotonating agent, are binuclear and coordinated via the NNNO sites of two ligand molecules. All the binuclear complexes have octahedral geometries and their magnetic moments are quite low compared to the calculated value for two metal ions complexes and thus antiferromagnetic interactions between the two adjacent metal ions. The ligand HL and metal complexes were tested against a strain of Gram +ve bacteria (Staphylococcus aureus), Gram -ve bacteria (Escherichia coli), and fungi (Candida albicans). The tested compounds exhibited high antibacterial activities.

  5. Synthesis, Characterization and Biological Activities of Cu(II, Co(II, Mn(II, Fe(II, and UO2(VI Complexes with a New Schiff Base Hydrazone: O-Hydroxyacetophenone-7-chloro-4-quinoline Hydrazone

    Directory of Open Access Journals (Sweden)

    Nora H. Al-Shaalan

    2011-10-01

    Full Text Available The Schiff base hydrazone ligand HL was prepared by the condensation reaction of 7-chloro-4-quinoline with o-hydroxyacetophenone. The ligand behaves either as monobasic bidentate or dibasic tridentate and contain ONN coordination sites. This was accounted for be the presence in the ligand of a phenolic azomethine and imine groups. It reacts with Cu(II, Ni(II, Co(II, Mn(II, UO2 (VI and Fe(II to form either mono- or binuclear complexes. The ligand and its metal complexes were characterized by elemental analyses, IR, NMR, Mass, and UV-Visible spectra. The magnetic moments and electrical conductance of the complexes were also determined. The Co(II, Ni(II and UO2 (VI complexes are mononuclear and coordinated to NO sites of two ligand molecules. The Cu(II complex has a square-planar geometry distorted towards tetrahedral, the Ni(II complex is octahedral while the UO2 (VI complex has its favoured heptacoordination. The Co(II, Mn(II complexes and also other Ni(II and Fe(III complexes, which were obtained in the presence of Li(OH as deprotonating agent, are binuclear and coordinated via the NNNO sites of two ligand molecules. All the binuclear complexes have octahedral geometries and their magnetic moments are quite low compared to the calculated value for two metal ions complexes and thus antiferromagnetic interactions between the two adjacent metal ions. The ligand HL and metal complexes were tested against a strain of Gram +ve bacteria (Staphylococcus aureus, Gram −ve bacteria (Escherichia coli, and fungi (Candida albicans. The tested compounds exhibited high antibacterial activities.

  6. A study of fuel behavior under reactivity initiated accident conditions — review

    Science.gov (United States)

    Ishikawa, Michio; Shiozawa, Shusaku

    1980-11-01

    Results obtained in the 400 tests performed to simulate reactivity initiated accidents since 1975 in the Japanese Nuclear Safety Research Reactor, are described. Tests included the effects of cooling environment, defective fuel elements, fuel design parameters, the behaviour of fuel elements for various reactor types, all done for a wide range of energy deposition. Four types of basic fuel failure mechanisms have been established, and are discussed in detail: cladding melt failure, UO 2 melt failure, high temperature burst failure and low temperature burst failure. Future test plans up to 1990 are out-lined and features requiring particular attention are pointed out.

  7. PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D.; Gamble, K. A.

    2015-09-01

    Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.

  8. A multiphase interfacial model for the dissolution of spent nuclear fuel

    Science.gov (United States)

    Jerden, James L.; Frey, Kurt; Ebert, William

    2015-07-01

    The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H2O2 and O2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary

  9. Thermodynamic treatment of noble metal fission products in nuclear fuel

    Science.gov (United States)

    Kaye, M. H.; Lewis, B. J.; Thompson, W. T.

    2007-06-01

    Based on a critical evaluation of the literature, a comprehensive thermodynamic model has been developed for the complete quinary system involving the noble metal fission products in nuclear fuel: Mo-Pd-Rh-Ru-Tc. This treatment was based on the foundation of ten binary systems and an interpolation scheme. The thermodynamic model has been demonstrated to fit the available experimental data for the ternary sub-systems. This work can be used with other models for potentially non-stoichiometric UO 2+ x containing fission products, as well as data for other phases, to assess the chemical form of fission products in irradiated fuel material.

  10. Dissolution of spent nuclear fuel in carbonate-peroxide solution

    Science.gov (United States)

    Soderquist, Chuck; Hanson, Brady

    2010-01-01

    This study shows that spent UO2 fuel can be completely dissolved in a room temperature carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate-peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.

  11. Two actinide-organic frameworks constructed by a tripodal flexible ligand: Occurrence of infinite {(UO2)O2(OH)3}4n and hexanuclear {Th6O4(OH)4} motifs

    Science.gov (United States)

    Liang, Lingling; Zhang, Ronglan; Zhao, Jianshe; Liu, Chiyang; Weng, Ng Seik

    2016-11-01

    Two new actinide metal-organic frameworks were constructed by using a tripodal flexible ligand tris (2-carboxyethyl) isocyanurate (H3tci) under hydrothermal condition. The combination of H3tci and uranyl nitrate hexahydrate in aqueous solution leads to the isolation of [(UO2)2(H2O)4]0.5(tci)2(UO2)4(OH)4·18H2O (1), which contains two distinct UO22+ coordination environments. Four uranyl cations, linked through μ3-OH respectively, result in the edge-sharing ribbons. Then, the layer structure is constructed by U-O clusters linked through other eight-coordinated uranyl unions, giving rise to a porous structure in the space. Topological analysis reveals that complex 1 belongs to a (4, 8)-connected net with a schläfli symbol of (34.26.3)2(34.46.56.68.73.8). Th3(tci)2O2(OH)2(H2O)3·12H2O (2) generated by the reaction of H3tci and thorium nitrate tetrahydrate, possesses nine-fold coodinated Th(IV) centers with a monocapped square antiprismatic geometry. The hexamers "Th6O4(OH)4" motifs are connected together by the carboxylate groups, showing a three-dimensional structures. Complex 2 takes on an 8-connected architecture and the point symbol is (424.64).

  12. Ligational behavior of thiosemicarbazone, semicarbazone and thiocarbohydrazone ligands towards VO(IV), Ce(III), Th(IV) and UO 2(VI) ions: Synthesis, structural characterization and biological studies

    Science.gov (United States)

    Shebl, M.; Seleem, H. S.; El-Shetary, B. A.

    2010-01-01

    Mono- and binuclear VO(IV), Ce(III), Th(IV) and UO 2(VI) complexes of thiosemicarbazone, semicarbazone and thiocarbohydrazone ligands derived from 4,6-diacetylresorcinol were synthesized. The structures of these complexes were elucidated by elemental analyses, IR, UV-vis, ESR, 1H NMR and mass spectra as well as conductivity and magnetic susceptibility measurements and thermal analyses. The thiosemicarbazone (H 4L 1) and the semicarbazone (H 4L 2) ligands behave as dibasic pentadentate ligands in case of VO(IV) and UO 2(VI) complexes, tribasic pentadentate in case of Ce(III) complexes and monobasic pentadentate in case of Th(IV) complexes. However, the thiocarbohydrazone ligand (H 3L 3) acts as a monobasic tridentate ligand in all complexes except the VO(IV) complex in which it acts as a dibasic tridentate ligand. The antibacterial and antifungal activities were also tested against Rhizobium bacteria and Fusarium-Oxysporium fungus. The metal complexes of H 4L 1 ligand showed a higher antibacterial effect than the free ligand while the other ligands (H 4L 2 and H 3L 3) showed a higher effect than their metal complexes. The antifungal effect of all metal complexes is lower than the free ligands.

  13. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  14. H 2 inhibition of radiation induced dissolution of spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Roth, Olivia; Jonsson, Mats

    2009-01-01

    In order to elucidate the effect of noble metal clusters in spent nuclear fuel on the kinetics of radiation induced spent fuel dissolution we have used Pd particle doped UO 2 pellets. The catalytic effect of Pd particles on the kinetics of radiation induced dissolution of UO 2 during γ-irradiation in HCO3- containing solutions purged with N 2 and H 2 was studied in this work. Four pellets with Pd concentrations of 0%, 0.1%, 1% and 3% were produced to mimic spent nuclear fuel. The pellets were placed in 10 mM HCO3- aqueous solutions and γ-irradiated, and the dissolution of UO22+ was measured spectrophotometrically as a function of time. Under N 2 atmosphere, 3% Pd prevent the dissolution of uranium by reduction with the radiolytically produced H 2, while the other pellets show a rate of dissolution of around 1.6 × 10 -9 mol m -2 s -1. Under H 2 atmosphere already 0.1% Pd effectively prevents the dissolution of uranium, while the rate of dissolution for the pellet without Pd is 1.4 × 10 -9 mol m -2 s -1. It is also shown in experiments without radiation in aqueous solutions containing H 2O 2 and O 2 that ɛ-particles catalyze the oxidation of the UO 2 matrix by these molecular oxidants, and that the kinetics of the catalyzed reactions is close to diffusion controlled.

  15. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  16. Safety and Regulatory Issues of the Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

    2014-02-01

    Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

  17. The dissolution rate of unirratiated uranium dioxide under repository conditions: The influence of fuel and water chemistry, dissolved oxygen, and temperature

    Science.gov (United States)

    Casella, Amanda J.

    The dissolution rate of both unirradiated UO2 and spent fuel has been studied by numerous countries as part of the performance assessments of proposed geologic repositories. 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 dose over geologic time frames. The primary objective of this research was to determine the influence these parameters have on the dissolution rate of unirradiated UO2 under Yucca Mountain repository conditions and compare them to the current Yucca Mountain Model. Fuels containing between 0 and 8 wt% Gd2O3-doped UO2 were tested in a single-pass flow-through setup. These tests have verified that in bicarbonate solutions as temperature increased the dissolution rate increased. However, the presence of silicate in the feedwater altered the system and lowered the dissolution rate at higher temperatures. Pure UO 2 samples exhibited a dependence on the dissolved oxygen concentration, which in the current experiments was varied from 3.0 to 8.7 ppm. The significance of this dissolved oxygen dependence increased with rising temperature. At 75ºC the powder samples had a maximum dependence of 0.7, although the fragment samples had a much larger dependence up to 2.2. For the case of the Gd2O3-doped samples, there was minimal oxygen dependence at any temperature. The Gd2O3-dopant stabilized the fuel matrix, which lowered the dissolution rates by over an order of magnitude at the higher dopant levels. This effect in lowering the dissolution rate was more pronounced at higher temperatures, and additional dopant continued to decrease the dissolution rate up to the 4 wt% Gd2O3-doped UO2 tested. The dissolution rates for pure UO2 compared reasonably well with the Yucca Mountain Model for tests performed at 50ºC and 75ºC, but were found to be approximately half the values predicted by the model at 25ºC. After long

  18. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

    Science.gov (United States)

    Ott, L. J.; Robb, K. R.; Wang, D.

    2014-05-01

    Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

  19. Operational, safety and transmutation behavior of BWR with thorium-based fuel; Betriebs-, Sicherheits- und Transmutationsverhalten von SWR mit thoriumbasiertem Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Winter, D.; Nabbi, R.; Thomauske, B. [RWTH Aachen (Germany). Inst. fuer Nuklearen Brennstoffkreislauf

    2012-11-01

    The contribution on operational, safety and transmutation behavior of BWR with thorium-based fuel is based on high-resolution simulation models for analysis of the neutron physics in case of heterogeneous flow profiles and neutron spectra in the reactor core using thorium-based fuel. It was shown that thorium-based fuel produces less TRU (transuranium elements) than UO2 fuel. Due to the beneficial neutron physical spectral properties it can be expected that this SWR fuel allows a more effective TRU transmutation. In addition more advantageous safety properties are expected.

  20. Distribution Coefficient Model of Main Component in Th(NO3)4-UO2 (NO3)2-HNO3-H2 O/30%TBP-dodecane Extraction System%Th(NO3)4-UO2(NO3)2-HNO3-H2 O/30%TBP-正十二烷萃取体系主要组分分配比模型

    Institute of Scientific and Technical Information of China (English)

    于婷; 李峥; 赵皓贵; 何淑华; 何辉; 李晴暖; 张岚

    2015-01-01

    Based on distribution coefficient data of thorium (Th),uranium (U)and nitric acid in the tributyl phosphate (TBP)extraction system,and by comparing and analyzing the presented distribution coefficient models in the literature,a new mathe-matical distribution coefficient model with relative computer program for simulating the extraction behavior of Th(Ⅳ),U(Ⅵ)and HNO3 in the TBP extraction system.The reliability of the new model with computer program was verified by 34 sets of distribu-tion coefficient data.The calculated results agree well with the experimental data.The results by comparing the actual distribution coefficient data with the computing results indicate that the calculated results by using this new model are more reliable and accu-rate than those of SEPHIS model reported previously.The model presented in this work can be a foundation for simulating the extraction behavior of thorium,uranium and nitric acid in the extraction system of Th (NO3 )4-UO2 (NO3 )2-HNO3-H2 O/30%TBP-dodecane with uranium concentrations from 0 g/L to 100 g/L,thorium from 0 g/L to 232 g/L,and HNO3 from 0 g/L to 4.5 mol/L at 25 ℃.%利用文献报道的Th(NO3)4-UO2(NO3)2-HNO3-H2 O/30%TBP-正十二烷体系各组分的分配比实验数据对现有的分配比模型进行分析和比对,提出了一个计算该体系各组分分配比的新模型。利用34组实验数据对新模型进行了验证,符合情况良好。计算结果表明,本文提出的模型明显优于原模型,可作为Th(NO3)4-UO2(NO3)2-HNO3-H2 O/30%TBP-正十二烷萃取体系中 Th(Ⅳ)、U(Ⅵ)和 HNO3萃取行为计算机模拟的基础。模型建立的条件为:温度,25℃;U(Ⅵ)浓度,0~100 g/L;Th(Ⅳ)浓度,0~232 g/L;硝酸浓度,0~4.5 mol/L。

  1. Experiences in certification of packages for transportation of fresh nuclear fuel in the context of new safety requirements established by IAEA regulations (IAEA-96 regulations, ST-1) for air transportation of nuclear materials (requirements to C-type packages)

    Energy Technology Data Exchange (ETDEWEB)

    Dudai, V.I.; Kovtun, A.D.; Matveev, V.Z.; Morenko, A.I.; Nilulin, V.M.; Shapovalov, V.I.; Yakushev, V.A.; Bobrovsky, V.S.; Rozhkov, V.V.; Agapov, A.M.; Kolesnikov, A.S. [Russian Federal Nuclear Centre - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)]|[JSC ' ' MSZ' ' , Electrostal (Russian Federation)]|[JSC ' ' NPCC' ' , Novosibirsk (Russian Federation)]|[Minatom of Russia, Moscow (Russian Federation)]|[Gosatomnadzor of Russia, Moscow (Russian Federation)

    2004-07-01

    Every year in Russia, a large amount of domestic and international transportation of fresh nuclear fuel (FNF) used in Russian and foreign energy and research atomic reactors and referred to fissile materials based on IAEA Regulations is performed. Here, bulk transportation is performed by air, and it concerns international transportation in particular. According to national ''Main Regulations for Safe Transport and physical Protection of Nuclear Materials (OPBZ- 83)'' and ''Regulations for the Safe Transport of Radioactive Materials'' of the International Atomic Energy Agency (IAEA Regulations), nuclear and radiation security under normal (accident free) and accident conditions of transport must be completely provided by the package design. In this context, high requirements to fissile packages exposed to heat and mechanical loads in transport accidents are imposed. A long-standing experience in accident free transportation of FM has shown that such approach to provide nuclear and radiation security pays for itself completely. Nevertheless, once in 10 years the International Atomic Energy Agency on every revision of the ''Regulations for the Safe Transport of Radioactive Materials'' places more stringent requirements upon the FM and transportation thereof, resulting from the objectively increasing risk associated with constant rise in volume and density of transportation, and also strained social and economical situation in a number of regions in the world. In the new edition of the IAEA Regulations (ST-1), published in 1996 and brought into force in 2001 (IAEA-96 Regulations), the requirements to FM packages conveyed by aircraft were radically changed. These requirements are completely presented in new Russian ''Regulations for the Safe Transport of Radioactive Materials'' (PBTRM- 2004) which will be brought into force in the time ahead.

  2. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou

    2011-09-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  3. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  4. New approaches to reprocessing of oxide nuclear fuel.

    Science.gov (United States)

    Myasoedov, B F; Kulyako, Yu M

    Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9-1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL(-1)). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

  5. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  6. Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

    Energy Technology Data Exchange (ETDEWEB)

    Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2016-05-01

    In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

  7. Baryogensis in fresh inflation

    CERN Document Server

    Bellini, M

    2002-01-01

    I study the possibility of baryogenesis can take place in fresh inflation. I find that it is possible that violation of baryon number conservation can occur during the period out-of-equilibrium in this scenario. Indeed, baryogenesis could be possible before the thermal equilibrium is restored at the end of fresh inflation.

  8. Modeling defect and fission gas properties in U-Si fuels

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark J. [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri Univ. of Science and Technology, Rolla, MO (United States); Grimes, Robin W. [Imperial College, London (United Kingdom)

    2017-04-14

    Uranium silicides, in particular U3Si2, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO2 fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

  9. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    OpenAIRE

    Shengli Chen; Cenxi Yuan

    2017-01-01

    Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into con...

  10. Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

    Directory of Open Access Journals (Sweden)

    Chollet Mélanie

    2017-01-01

    Full Text Available Fission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro- and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped UO2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO2 by positron annihilation spectroscopy (PAS. Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22Na source sandwiched between two 500-μm-thin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO2, 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects. In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO2 produced by Cr and Al oxides

  11. Derivation of effective fission gas diffusivities in UO2 from lower length scale simulations and implementation of fission gas diffusion models in BISON

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, Xiang-Yang [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Perriot, Romain Thibault [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tonks, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-11-07

    This report summarizes the development of new fission gas diffusion models from lower length scale simulations and assessment of these models in terms of annealing experiments and fission gas release simulations using the BISON fuel performance code. Based on the mechanisms established from density functional theory (DFT) and empirical potential calculations, continuum models for diffusion of xenon (Xe) in UO2 were derived for both intrinsic conditions and under irradiation. The importance of the large XeU3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.

  12. Measurements of Fission and Radioactive Capture Reaction Rates Inside the Fuel of the Ipen/MB-01

    Science.gov (United States)

    Mura, Luís Felipe L.; Bitelli, Ulysses d'Utra; Fanaro, Leda C. C. B.

    2011-05-01

    This work presents the measures of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO2 at 4.3% enrichment. From its irradiation, the rate of radioactive capture and fission had been measured as a function of the radius of the pellet disk using a Ortec GMX HPGe detector. Lead collimators had been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO2 disk is used, being inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 h under a neutron flux of 5 ×108 n/cm2 s. In gamma spectrometry, 10 collimators with different diameters have been used; consequently, the nuclear reactions of radioactive capture that occurs in atoms of 238U and the fission that occurs on both 235U and 238U are measured in function of 10 different regions (diameter of collimator) of the UO2 fuel pellet disk. Nuclear fission produces different fission products such as 143Ce with a yield fission of 5.9% which decay is monitored in this work. Corrections in geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement.

  13. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  14. Influences of parameter uncertainties within the ICRP-66 respiratory tract model: regional tissue doses for 239PuO2 and 238UO2/238U3O8.

    Science.gov (United States)

    Farfán, Eduardo B; Huston, Thomas E; Bolch, W Emmett; Vernetson, William G; Bolch, Wesley E

    2003-04-01

    This paper extends an examination of the influence of parameter uncertainties on regional doses to respiratory tract tissues for short-ranged alpha particles using the ICRP-66 respiratory tract model. Previous papers examined uncertainties in the deposition and clearance aspects of the model. The critical parameters examined in this study included target tissue depths, thicknesses, and masses, particularly within the thoracic or lung regions of the respiratory tract. Probability density functions were assigned for the parameters based on published data. The probabilistic computer code LUDUC (Lung Dose Uncertainty Code) was used to assess regional and total lung doses from inhaled aerosols of 239PuO2 and 238UO2/238U3O8. Dose uncertainty was noted to depend on the particle aerodynamic diameter. Additionally, dose distributions were found to follow a lognormal distribution pattern. For 239PuO2 and 238UO2/238U3O8, this study showed that the uncertainty in lung dose increases by factors of approximately 50 and approximately 70 for plutonium and uranium oxides, respectively, over the particle size range from 0.1 to 20 microm. For typical exposure scenarios involving both radionuclides, the ratio of the 95% dose fractile to the 5% dose fractile ranged from approximately 8-10 (corresponding to a geometric standard deviation, or GSD, of about 1.7-2) for particle diameters of 0.1 to 1 microm. This ratio increased to about 370 for plutonium oxide (GSD approximately 4.5) and to about 600 for uranium oxide (GSD approximately 5) as the particle diameter approached 20 microm. However, thoracic tissue doses were quite low at larger particle sizes because most of the deposition occurred in the extrathoracic airways. For 239PuO2, median doses from LUDUC were found be in general agreement with those for Reference Man (via deterministic LUDEP 2.0 calculations) in the particle range of 0.1 to 5 microm. However, median doses to the basal cell nuclei of the bronchial airways (BB

  15. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Science.gov (United States)

    Rose, S. J.; Wilson, J. N.; Capellan, N.; David, S.; Guillemin, P.; Ivanov, E.; Méplan, O.; Nuttin, A.; Siem, S.

    2012-02-01

    The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  16. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Directory of Open Access Journals (Sweden)

    Nuttin A.

    2012-02-01

    Full Text Available The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX and uranium/plutonium mixed oxide (MOX fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  17. Dissolution of fluorite type surfaces as analogues of spent nuclear fuel : Production of suitable analogues and study the effect of surface orientation on dissolution

    OpenAIRE

    Godinho, Jose

    2011-01-01

    It is accepted worldwide that the best final solution for spent nuclear fuel is to bury it in deep geological repositories. Despite the physical and chemical barriers that are supposed to isolate the nuclear waste for at least 100.000 years, some uncertainty factors may cause underground water to get in contact with the nuclear waste. Due to radioactivity and oxidation under air, dissolution experiments using UO2 pellets are difficult and frequently lead to incoherent results. Therefore, to e...

  18. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen, Director

    2011-04-01

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  19. Optimisation for the Blending of MOX Fuel%MOX燃料混料过程的优化

    Institute of Scientific and Technical Information of China (English)

    李怀林; 李文埮

    2001-01-01

    本文采用Householder变换法对MOX燃料混料过程中Pu同位素均一化问题进行优化计算,并用轨迹求解法对球磨中的转速问题进行了初步探讨。%The blending of UO2 and PuO2 powders is the key technology in theMOX fuel manufactor. The Pu isotopic homogeneous, blending of UO2 and PuO2 and ball milling will be done in the blending process. In the paper, the House holder transform is applied to calculate the Pu isotopic homogeneous, and the track method is adopted to calculate speed of ball milling. All of the calculated results are accordance with the those from reference.

  20. Challenges in spent nuclear fuel final disposal:conceptual design models

    Institute of Scientific and Technical Information of China (English)

    Mukhtar Ahmed RANA

    2008-01-01

    The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.

  1. Stability and instability of the isoelectronic UO2(2+) and PaO2+ actinyl oxo-cations in aqueous solution from density functional theory based molecular dynamics.

    Science.gov (United States)

    Spezia, Riccardo; Siboulet, Bertrand; Abadie, Sacha; Vuilleumier, Rodolphe; Vitorge, Pierre

    2011-04-07

    In this work, Pa(V) monocations have been studied in liquid water by means of density functional theory (DFT) based molecular dynamic simulations (CPMD) and compared with their U(VI) isoelectronic counterparts to understand the peculiar chemical behavior of Pa(V) in aqueous solution. Four different Pa(V) monocationic isomers appear to be stable in liquid water from our simulations: [PaO(2)(H(2)O)(5)](+)(aq), [Pa(OH)(4)(H(2)O)(2)](+)(aq), [PaO(OH)(2)(H(2)O)(4)](+)(aq), and [Pa(OH)(4)(H(2)O)(3)](+)(aq). On the other hand, in the case of U(VI) only the uranyl, [UO(2)(H(2)O)(5)](2)(+)(aq), is stable. The other species containing hydroxyl groups replacing one or two oxo bonds are readily converted to uranyl. The Pa-OH bond is stable, while it is suddenly broken in U-OH. This makes possible the formation of a broad variety of Pa(V) species in water and participates to its unique chemical behavior in aqueous solution. Further, the two actinyl oxocations in water are different in the ability of the oxygen atoms to form stable and extended H-bond networks for Pa(V) contrary to U(VI). In particular, protactinyl is found to have between 2 and 3 hydrogen bonds per oxygen atom while uranyl has between zero and one.

  2. Antimicrobial, spectral, magnetic and thermal studies of Cu(II), Ni(II), Co(II), UO(2)(VI) and Fe(III) complexes of the Schiff base derived from oxalylhydrazide.

    Science.gov (United States)

    Melha, Khlood Abou

    2008-04-01

    The Schiff base ligand, oxalyl [( 2 - hydroxybenzylidene) hydrazone] [corrected].H(2)L, and its Cu(II), Ni(II), Co(II), UO(2)(VI) and Fe(III) complexes were prepared and tested as antibacterial agents. The Schiff base acts as a dibasic tetra- or hexadentate ligand with metal cations in molar ratio 1:1 or 2:1 (M:L) to yield either mono- or binuclear complexes, respectively. The ligand and its metal complexes were characterized by elemental analyses, IR, (1)H NMR, Mass, and UV-Visible spectra and the magnetic moments and electrical conductance of the complexes were also determined. For binuclear complexes, the magnetic moments are quite low compared to the calculated value for two metal ions complexes and this shows antiferromagnetic interactions between the two adjacent metal ions. The ligand and its metal complexes were tested against a Gram + ve bacteria (Staphylococcus aureus), a Gram -ve bacteria (Escherichia coli), and a fungi (Candida albicans). The tested compounds exhibited high antibacterial activities.

  3. Quantification of process variables for carbothermic synthesis of UC1-xNx fuel microspheres

    Science.gov (United States)

    Lindemer, T. B.; Silva, C. M.; Henry, J. J.; McMurray, J. W.; Voit, S. L.; Collins, J. L.; Hunt, R. D.

    2017-01-01

    This report details the continued investigation of process variables involved in converting sol-gel-derived, urania-carbon microspheres to ∼820-μm-dia. UC1-xNx fuel kernels in flow-through, vertical Mo and W crucibles at temperatures up to 2123 K. Experiments included calcining of air-dried UO3-H2O-C microspheres in Ar and H2-containing gases, conversion of the resulting UO2-C kernels to dense UO2:2UC in the same gases and vacuum, and its conversion in N2 to UC1-xNx (x = ∼0.85). The thermodynamics of the relevant reactions were applied extensively to interpret and control the process variables. Producing the precursor UO2:2UC kernel of ∼96% theoretical density was required, but its subsequent conversion to UC1-xNx at 2123 K was not accompanied by sintering and resulted in ∼83-86% of theoretical density. Increasing the UC1-xNx kernel nitride component to ∼0.98 in flowing N2-H2 mixtures to evolve HCN was shown to be quantitatively consistent with present and past experiments and the only useful application of H2 in the entire process.

  4. Thermal expansion of simulated thoria-urania fuel by high temperature XRD

    Science.gov (United States)

    Bhagat, R. K.; Krishnan, K.; Kutty, T. R. G.; Kumar, Arun; Kamath, H. S.; Banerjee, S.

    2012-03-01

    The thermal expansion behavior of polycrystalline samples of ThO2-3.45% UO2 and SIMFUEL corresponding to burn-up of 43,000 MWd/Te has been investigated from room temperature to 1473 K, and for SIMFUEL corresponding to burn-up of 28,000 MWd/Te has been investigated from room temperature to 1173 K, using a high temperature X-ray diffraction (HTXRD). Linear and volumetric thermal expansion data like, percentage thermal expansion, average or mean coefficient of thermal expansion (CTE) was generated using the lattice parameters. It is observed that SIMFUEL has a lower lattice parameter compared to ThO2-3.45% UO2 and this is attributed to the dissolution of the rare earths and part of the Zr and Ce in fuel matrix. Also SIMFUEL has slightly higher thermal expansion than ThO2-3.45% UO2 and this is related to the lower melting point of SIMFUEL.

  5. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for

  6. Corrosion of used nuclear fuel in aqueous perchlorate and carbonate solutions

    Science.gov (United States)

    Shoesmith, D. W.; Sunder, S.; Bailey, M. G.; Miller, N. H.

    1996-01-01

    The corrosion of used fuel was investigated using electrodes constructed from fuel pins discharged from the Pickering, Bruce and Darlington CANDU reactors, and compared to the corrosion behaviour observed on unirradiated UO 2 and SIMFUEL. Experiments were carried out in solutions of NaClO 4 (pH˜ 9.5) in the presence and absence of (a) substantial concentrations of sodium carbonate, and (b) additional external gamma fields. Used fuel electrodes reached oxidizing corrosion potentials ( ECORR) rapidly compared with unirradiated UO 2 electrodes. However, optical and SEM examinations showed no evidence for rapid oxidative dissolution. This reaction, expected to be fast since high values of ECORR are observed, appears to be blocked by the accumulation of secondary phases in grain boundaries. The oxidation and dissolution behaviour of used fuel is determined predominantly by (i) the dose rate in solution near the fuel surface, (ii) the extent of burnup (which determines the degree of fission product doping), and (iii) the degree of non-stoichiometry.

  7. New processing methods to produce silicon carbide and beryllium oxide inert matrix and enhanced thermal conductivity oxide fuels

    Science.gov (United States)

    Sarma, K. H.; Fourcade, J.; Lee, S.-G.; Solomon, A. A.

    2006-06-01

    For inert matrix fuels, SiC and BeO represent two possible matrix phase compounds that exhibit very high thermal conductivity, high melting points, low neutron absorption, and reasonably high radiation stability. BeO is chemically compatible with UO2, PuO2 and Zircaloy to very high temperatures, but SiC reacts with all three at somewhat lower temperatures. We have developed the Polymer Impregnation and Pyrolysis or PIP method, making use of a commercial SiC polymeric precursor, to consolidate both particulate fuels like 'TRISO' microsphere fuels, and to impregnate UO2 fuels with pure stoichiometric SiC to improve their thermal conductivity. This method was employed to fabricate Enhanced Conductivity Oxide fuels, or ECO fuels with 5-10 vol.% of the high conductivity phase, and with 50 vol.% for TRISO dispersion fuels. For ECO fuels, a new 'slug/bisque' method of fabricating the UO2 fuel granules was necessary to produce sintered fuel with open pore structures, allowing almost complete impregnation of the continuous SiC phase. The advantages of the PIP process are that it is a non-damaging consolidation process for particulates (TRU, UC or TRISO microspheres), forms a continuous, pure β-SiC phase at temperatures as low as 1573 K, and allows the maximum in fissile atom density. However, several PIP impregnation cycles and high crystallization temperatures are necessary to obtain high thermal conductivity SiC. For producing IMF fuels using the PIP process, the fissile PuC and/or TRU actinides can be added in small concentrations along with SiC 'filler particles' and consolidated with the SiC precursor for either open or closed fuel cycles. For BeO, a second approach was developed for ECO fuels that involves a 'co-sintering' route to produce high density fuels with a continuous BeO phase of 5-10 vol.%. Special granulation and mixing techniques were developed, but only one normal sintering cycle is required. For BeO matrix IMF fuels, PuO2 granules and TRU actinides or

  8. MARMOT update for oxide fuel modeling

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Chao [Idaho National Lab. (INL), Idaho Falls, ID (United States); Aagesen, Larry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ahmed, Karim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Wen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, Bulent [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Tonks, Michael [Pennsylvania State Univ., University Park, PA (United States); Millett, Paul [Univ. of Arkansas, Fayetteville, AR (United States)

    2016-09-01

    This report summarizes the lower-length-scale research and development progresses in FY16 at Idaho National Laboratory in developing mechanistic materials models for oxide fuels, in parallel to the development of the MARMOT code which will be summarized in a separate report. This effort is a critical component of the microstructure based fuel performance modeling approach, supported by the Fuels Product Line in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The progresses can be classified into three categories: 1) development of materials models to be used in engineering scale fuel performance modeling regarding the effect of lattice defects on thermal conductivity, 2) development of modeling capabilities for mesoscale fuel behaviors including stage-3 gas release, grain growth, high burn-up structure, fracture and creep, and 3) improved understanding in material science by calculating the anisotropic grain boundary energies in UO$_2$ and obtaining thermodynamic data for solid fission products. Many of these topics are still under active development. They are updated in the report with proper amount of details. For some topics, separate reports are generated in parallel and so stated in the text. The accomplishments have led to better understanding of fuel behaviors and enhance capability of the MOOSE-BISON-MARMOT toolkit.

  9. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    Science.gov (United States)

    González-Robles, E.; Metz, V.; Wegen, D. H.; Herm, M.; Papaioannou, D.; Bohnert, E.; Gretter, R.; Müller, N.; Nasyrow, R.; de Weerd, W.; Wiss, T.; Kienzler, B.

    2016-10-01

    During reactor operation the fission gases Kr and Xe are formed within the UO2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO2 fuel itself are widely used as indicators for the release properties of 129I, 137Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  10. In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

    Science.gov (United States)

    Diianni, D. C.

    1972-01-01

    The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

  11. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2014-06-01

    2 particle fuel up to about 10% FIMA and 1150°C, UO2 fuel is known to have limitations because of CO formation and kernel migration at the high burnups, power densities, temperatures, and temperature gradients that may be encountered in the prismatic modular HTGRs. With uranium oxycarbide (UCO) fuel, the kernel composition is engineered to prevent CO formation and kernel migration, which are key threats to fuel integrity at higher burnups, temperatures, and temperature gradients. Furthermore, the recent poor fuel performance of UO2 TRISO fuel pebbles measured in Chinese irradiation testing in Russia and in German pebbles irradiated at 1250°C, and historic data on poorer fuel performance in safety testing of German pebbles that experienced burnups in excess of 10% FIMA [1] have each raised concern about the use of UO2 TRISO above 10% FIMA and 1150°C and the degree of margin available in the fuel system. This continues to be an active area of study internationally.

  12. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  13. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  14. Crystal fields in UO2 - revisited

    Energy Technology Data Exchange (ETDEWEB)

    Nakotte, Heinz [Los Alamos National Laboratory; Rajatram, R [NMSU/UNIV OF N.C.; Kern, S [COLORADO STATE UNIV; Mcqueeney, R J [AMES LAB; Lander, G H [EUROPEAN COMMISIONS, JRC; Robinson, R A [BRAGG INSTITUTE

    2009-01-01

    We performed inelastic neutron scattering (INS) in order to re-investigate the crystal-field ground state and the level splitting in UO{sub 2}. Previous INS studies on UO{sub 2} by Amorelli et al. [Physical Review B 15, 1989, 1856] uncovered four excitations at low temperatures in the 150-180 meV range. Considering the dipole-allowed transitions, only three of these transitions could be explained by the published crystal-field model. Our INS results on a different UO{sub 2} sample revealed that the unaccounted peak at about 180 meV is a spurious one, and thus not intrinsic to UO{sub 2}. In good agreement with Amoretti's results, we corroborated that the ground-state of UO{sub 2} is the {Lambda}{sub 5} triplet, and we computed that the fourth- and six-order crystal field parameters are V{sub 4} = -116 meV and V{sub 6} = 26 meV, respectively. We also studied the INS response of the non-magnetic U{sub 0.4}Th{sub 0.6}O{sub 2}. The splitting for this thorium-doped compound is similar to the one of UO{sub 2}, which orders antiferromagnetically at low temperatures. Therefore, we can conclude that magnetic interactions only weakly perturb the energy level splitting, which is dominated by strong crystal fields.

  15. Intelligent Fish Freshness Assessment

    Directory of Open Access Journals (Sweden)

    Hamid Gholam Hosseini

    2008-01-01

    Full Text Available Fish species identification and automated fish freshness assessment play important roles in fishery industry applications. This paper describes a method based on support vector machines (SVMs to improve the performance of fish identification systems. The result is used for the assessment of fish freshness using artificial neural network (ANN. Identification of the fish species involves processing of the images of fish. The most efficient features were extracted and combined with the down-sampled version of the images to create a 1D input vector. Max-Win algorithm applied to the SVM-based classifiers has enhanced the reliability of sorting to 96.46%. The realisation of Cyranose 320 Electronic nose (E-nose, in order to evaluate the fish freshness in real-time, is experimented. Intelligent processing of the sensor patterns involves the use of a dedicated ANN for each species under study. The best estimation of freshness was provided by the most sensitive sensors. Data was collected from four selected species of fishes over a period of ten days. It was concluded that the performance can be increased using individual trained ANN for each specie. The proposed system has been successful in identifying the number of days after catching the fish with an accuracy of up to 91%.

  16. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  17. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    Science.gov (United States)

    Carmack, W. J.; Husser, D. L.; Mohr, T. C.; Richardson, W. C.

    2004-02-01

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  18. Research on PWR Core Performance With MOX Fuel Loading%MOX燃料对压水堆堆芯性能影响研究

    Institute of Scientific and Technical Information of China (English)

    李小生; 靳忠敏

    2013-01-01

    Use of MOX fuel in nuclear reactors is an effective way to dispose of plutonium .A large PWR reactor core with full core loading UO 2 fuel was referenced , the reactor core physics parameters of PWR with whole and part core loading MOX fuel were calculated by using DRAGON and DONJON codes ,and the reactivity worth of control rods and boron acid solution were researched under loading MOX fuel . The results show that PWR core with MOX fuel can achieve the desired cycle length and power distribution ,but loading MOX fuel will significantly decrease the reactivity worth of control rod and boron acid solution ,moreover ,the proportion of loading MOX fuel is positive to the decrease degree of reactivity worth .%在核反应堆中使用MOX燃料是处置钚的有效方式。以大型全UO2燃料压水堆堆芯设计作为参考,使用DRAGON、DONJON程序,计算在大型压水堆中全堆芯及部分堆芯装载MOX燃料后反应堆部分物理性能指标,研究加入MOX燃料后对控制棒与硼酸溶液的反应性价值的影响。结果表明,压水堆堆芯装载各比例MOX燃料均可达到与全UO2燃料堆芯相当的循环长度,功率分布也能满足相应的安全限值要求,但采用MOX燃料会造成控制棒与硼溶液的反应性价值降低,且降低程度与MOX燃料装载比例成正相关。

  19. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  20. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  1. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  2. Analysis on Fuel Thermal Conductivity Model of the Computer Code for Performance Prediction of Fuel Rods%燃料元件性能分析程序中的燃料热导率模型分析

    Institute of Scientific and Technical Information of China (English)

    李海; 黄晨; 杜爱兵; 徐宝玉

    2014-01-01

    The thermal conductivity is one of the most important parameters in the computer code for performance prediction for fuel rods.Several fuel thermal conductivity models used in foreign computer code,including thermal conductivity models for MOX fuel and UO2 fuel were introduced in this paper. Thermal conductivities were calculated by using these models, and the results were compared and analyzed.Finally, the thermal conductivity model for the native computer code for performance prediction for fuel rods in fast reactor was recommended.%热导率是燃料元件性能分析程序最重要的参数之一,本文介绍了各国部分性能分析程序的燃料热导率模型,按照 MOX和 UO2燃料分类,给出了这些性能分析程序热导率模型的计算结果,并进行分析对比,给出了国产快堆性能分析程序的热导率推荐模型。

  3. Reactor physics modelling of accident tolerant fuel for LWRs using answers codes

    OpenAIRE

    Lindley Benjamin A.; Kotlyar Dan; Parks Geoffrey T.; Lillington John N.; Petrovic Bojan

    2016-01-01

    This is the final version of the article. It first appeared from Springer via http://dx.doi.org/10.1051/epjn/2016012 The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/ PuO2 fuel designs which have an excellent performance record for normal operation and most transients. However, the events at Fukushima culminated in significant hydrogen production and hyd...

  4. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    Science.gov (United States)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

  5. Global Shortage of Fresh Water

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>阅读下表。以Global Shortage of Fresh Water为题写一篇短文。词数:100—120学生习作:Global Shortage of Fresh Water Fresh water seems ineverywhere,in rivers,lakes,wells as well as rain,which make some people think that we can’t use up water.

  6. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  7. Radiative capture on $^{242}$Pu for MOX fuel reactors

    CERN Multimedia

    The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

  8. Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

    Science.gov (United States)

    Jégou, C.; Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S.; Vercouter, T.; Roudil, D.

    2010-04-01

    The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2®) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 °C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO 2 matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO 2 grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH) 4(am) phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

  9. Expanding the Chemistry of Actinide Metallocene Bromides. Synthesis, Properties and Molecular Structures of the Tetravalent and Trivalent Uranium Bromide Complexes: (C5Me4R2UBr2, (C5Me4R2U(O-2,6-iPr2C6H3(Br, and [K(THF][(C5Me4R2UBr2] (R = Me, Et

    Directory of Open Access Journals (Sweden)

    Alejandro G. Lichtscheidl

    2016-01-01

    Full Text Available The organometallic uranium species (C5Me4R2UBr2 (R = Me, Et were obtained by treating their chloride analogues (C5Me4R2UCl2 (R = Me, Et with Me3SiBr. Treatment of (C5Me4R2UCl2 and (C5Me4R2UBr2 (R = Me, Et with K(O-2,6-iPr2C6H3 afforded the halide aryloxide mixed-ligand complexes (C5Me4R2U(O-2,6-iPr2C6H3(X (R = Me, Et; X = Cl, Br. Complexes (C5Me4R2U(O-2,6-iPr2C6H3(Br (R = Me, Et can also be synthesized by treating (C5Me4R2U(O-2,6-iPr2C6H3(Cl (R = Me, Et with Me3SiBr, respectively. Reduction of (C5Me4R2UCl2 and (C5Me4R2UBr2 (R = Me, Et with KC8 led to isolation of uranium(III “ate” species [K(THF][(C5Me52UX2] (X = Cl, Br and [K(THF0.5][(C5Me4Et2UX2] (X = Cl, Br, which can be converted to the neutral complexes (C5Me4R2U[N(SiMe32] (R = Me, Et. Analyses by nuclear magnetic resonance spectroscopy, X-ray crystallography, and elemental analysis are also presented.

  10. Irradiation of fresh fish

    Science.gov (United States)

    Yueh-jen, Yen; Jin-lai, Zhou; Shao-chun, Lai

    Occasionally, in China, marine products can not be provided for the markets in good quality, for during the time when they are being transported from the sea port to inland towns or even at the time when they are unloaded from the ship, they are beginning to spoil. Obviously, it is very important that appropiate measures should be taken to prevent them from decay. Our study has proved that the shelf life of fresh Flatfish (Cynoglossue robustus) and Silvery pomfret (stromateoides argenteus), which, packed in sealed containers, are irradiated by 1.5 kGy, 2.2 kGy and 3.0 kGy, can be stored for about 13-26 days at 3° - 5° C.

  11. Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A.; Hales, Jason D.

    2016-12-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of the concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.

  12. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  13. Corrosion of irradiated MOX fuel in presence of dissolved H 2

    Science.gov (United States)

    Carbol, P.; Fors, P.; Van Winckel, S.; Spahiu, K.

    2009-07-01

    The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO 3 solution in presence of dissolved H 2 for 2100 days. The results show that dissolved H 2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10 -10 and 5 × 10 -11 M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO 2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.

  14. Simulations of H 2O 2 concentration profiles in the water surrounding spent nuclear fuel

    Science.gov (United States)

    Nielsen, Fredrik; Lundahl, Karin; Jonsson, Mats

    2008-01-01

    A simple mathematical model describing the hydrogen peroxide concentration profile in water surrounding a spent nuclear fuel pellet as a function of time has been developed. The water volume is divided into smaller elements, and the processes that affect hydrogen peroxide concentration are applied to each volume element. The model includes production of H 2O 2 from α-radiolysis, surface reaction between H 2O 2 and UO 2 and diffusion. Simulations show that the surface concentration of H 2O 2 increases fairly rapidly and approaches the steady-state concentration. The time to reach steady-state is sufficiently short to be neglected compared to the times of interest when simulating spent fuel dissolution under deep repository conditions. Consequently, the steady-state approach can be used to estimate the rate for radiation-induced spent nuclear fuel dissolution.

  15. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  16. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    Directory of Open Access Journals (Sweden)

    Hyun-Gil Kim

    2016-02-01

    Full Text Available For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density composite pellet concepts are being developed as ATF pellets. A microcell UO2 pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts—surface-modified Zr-based alloy and SiC composite material—are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

  17. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  18. Thermal measurements and computational simulations of three-phase (CeO2-MgAl2O4-CeMgAl11O19) and four-phase (3Y-TZP-Al2O3-MgAl2O4-LaPO4) composites as surrogate inert matrix nuclear fuel

    Science.gov (United States)

    Angle, Jesse P.; Nelson, Andrew T.; Men, Danju; Mecartney, Martha L.

    2014-11-01

    This study investigates the temperature dependent thermal conductivity of multiphase ceramic composites for simulated inert matrix nuclear fuel. Fine grained composites were made of CeO2-MgAl2O4-CeMgAl11O19 or 3Y-TZP-Al2O3-MgAl2O4-LaPO4. CeO2 and 3Y-TZP are used as UO2 surrogates due to their similar structures and low thermal conductivities. Laser flash analysis from room temperature to 1273 K (1000 °C) was used to determine the temperature dependent thermal conductivity. A computational approach using Object Oriented Finite Element Analysis Version 2 (OOF2) was employed to simulate the composite thermal conductivity based on the microstructure. Observed discrepancies between experimental and simulated thermal conductivities at low temperature may be due to Kapitza resistance; however, there is less than 3% deviation between models and experiments above 673 K (400 °C) for both compositions. When the surrogate phase was replaced with UO2 in the computational model for the four-phase composite, a 12-16% increase in thermal conductivity resulted compared to single phase UO2, in the range of 673-1273 K (400-1000 °C). This computational approach may be potentially viable for the high-throughput evaluation of composite systems and the strategic selection of inert phases without extensive sample fabrication during the initial development stages of composite nuclear fuel design.

  19. HTGR spent fuel composition and fuel element block flow

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, C.J.; Holder, N.D.; Pierce, V.H.; Robertson, M.W.

    1976-07-01

    The High-Temperature Gas-Cooled Reactor (HTGR) utilizes the thorium-uranium fuel cycle. Fully enriched uranium fissile material and thorium fertile material are used in the initial reactor core and for makeup fuel in the recycle core loadings. Bred /sup 233/U and unburned /sup 235/U fissile materials are recovered from spent fuel elements, refabricated into recycle fuel elements, and used as part of the recycle core loading along with the makeup fuel elements. A typical HTGR employs a 4-yr fuel cycle with approximately one-fourth of the core discharged and reloaded annually. The fuel element composition, including heavy metals, impurity nuclides, fission products, and activation products, has been calculated for discharged spent fuel elements and for reload fresh fuel and recycle fuel elements for each cycle over the life of a typical HTGR. Fuel element compositions are presented for the conditions of equilibrium recycle. Data describing compositions for individual reloads throughout the reactor life are available in a detailed volume upon request. Fuel element block flow data have been compiled based on a forecast HTGR market. Annual block flows are presented for each type of fuel element discharged from the reactors for reprocessing and for refabrication.

  20. Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

    Science.gov (United States)

    Carroll, Spencer

    As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

  1. Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

    Science.gov (United States)

    Hallman, Luther, Jr.

    Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

  2. Development of an accident-tolerant fuel composite from uranium mononitride (UN) and uranium sesquisilicide (U3 Si2) with increased uranium loading

    Science.gov (United States)

    Ortega, Luis H.; Blamer, Brandon J.; Evans, Jordan A.; McDeavitt, Sean M.

    2016-04-01

    The processing steps necessary to prepare a potential accident-tolerant fuel composite consisting of uranium mononitride (UN) combined with uranium sesquisilicide (U3 Si2) are described. Liquid phase sintering was performed with U3 Si2 as the liquid phase combined with UN powder or UN μ-spheres. Various UN to U3 Si2 ratios were tested which resulted in up to 94% dense pellets. Composite UN-U3 Si2 samples had greater than 30% more uranium content than UO2.

  3. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  4. Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

    Directory of Open Access Journals (Sweden)

    Z. Gholamzadeh

    2015-12-01

    Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.

  5. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  6. Conversion of uranium nuclear fuel into U 3O 8 at the head end of HTR reprocessing

    Science.gov (United States)

    Hoogen, N.; Aschhoff, H. G.; Staib, G.

    1984-04-01

    Corresponding to the reference procedure for the head-end treatment of HTR fuel elements, separation of the moderator graphite from the materials uranium and plutonium is envisaged by combustion in the fluidized bed. Due to the defective silicon carbide layers of the uranium fuel particles a chemical conversion of the UO 2 kernel into U 3O 8 takes place in the oxidizing atmosphere of the combustion process. This reaction proceeds spontaneously and quantitatively, and causes a disintegration of the heavy metal kernel. It is observed that the degree of hardness of the kernel fragments is clearly dependent on the heat-up rate. In the commercial design of the head-end process step, attention must be paid to the cross-over of fuel from the stationary fluidized bed into the dust discharge.

  7. Fission gas release behaviour of a 103 GWd/tHM fuel disc during a 1200 °C annealing test

    Science.gov (United States)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.; Tverberg, T.

    2014-03-01

    Within the Nuclear Fuel Industry Research (NFIR) program, several fuel variants, in the form of thin circular discs, were irradiated in the Halden Boiling Water Reactor (HBWR) to a range of burn-ups ˜100 GWd/tHM. The design of the assembly was similar to that used in other HBWR programs: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature gradients within the fuel discs. One such rod contained standard grain UO2 discs (3D grain size = 18 μm) reaching a burn-up of 103 GWd/tHM. After the irradiation, the gas release upon rod puncturing was measured to be 2.9%.

  8. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  9. Reactor Physics Characterization of the HTR Module with UCO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  10. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    Science.gov (United States)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  11. Specific outcomes of the research on the spent fuel long-term evolution in interim dry storage and deep geological disposal

    Science.gov (United States)

    Ferry, C.; Poinssot, C.; Cappelaere, C.; Desgranges, L.; Jegou, C.; Miserque, F.; Piron, J. P.; Roudil, D.; Gras, J. M.

    2006-06-01

    This paper presents an overview of the main results of the French research on the long-term evolution of spent fuel. The behavior of the spent fuel rods in the various conditions likely to be encountered during dry storage and deep geological disposal, i.e., in a closed system, in air and in water were investigated. It appears that in a closed system the effects of helium production on the mechanical stability of grain boundaries remain the major unanswered question. In air, microscopic characterization of the UO2 oxidation leads to introduce a new phase in the classical oxidation scheme. The limiting step assumption on which the oxidation kinetics are based is only partially valid. In water, the effect of the alpha radiolysis which accelerates UO2 dissolution was demonstrated for anoxic conditions. However this effect could be counteracted by the environmental conditions, such as the presence of H2 produced by the container corrosion. The effects of the environmental parameters on the fuel matrix dissolution still need to be assessed.

  12. The IFR modern nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hannum, W.H.

    1991-01-01

    Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs.

  13. Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

    Science.gov (United States)

    Magnin, M.; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

    2015-07-01

    The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5-5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO22+) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10-7 mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10-5 mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3 × 10-6 mol L-1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation

  14. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  15. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

  16. 21 CFR 101.95 - “Fresh,” “freshly frozen,” “fresh frozen,” “frozen fresh.”

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false âFresh,â âfreshly frozen,â âfresh frozen,â âfrozen... fresh,” when used on the label or in labeling of a food, mean that the food was quickly frozen while still fresh (i.e., the food had been recently harvested when frozen). Blanching of the food...

  17. Stress Analysis of Coated Particle Fuel in the Deep-Burn Pebble Bed Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    High fuel temperatures and resulting fuel particle coating stresses can be expected in a Pu and minor actinide fueled Pebble Bed Modular Reactor (400 MWth) design as compared to the ’standard’ UO2 fueled core. The high discharge burnup aimed for in this Deep-Burn design results in increased power and temperature peaking in the pebble bed near the inner and outer reflector. Furthermore, the pebble power in a multi-pass in-core pebble recycling scheme is relatively high for pebbles that make their first core pass. This might result in an increase of the mechanical failure of the coatings, which serve as the containment of radioactive fission products in the PBMR design. To investigate the integrity of the particle fuel coatings as a function of the irradiation time (i.e. burnup), core position and during a Loss Of Forced Cooling (LOFC) incident the PArticle STress Analysis code (PASTA) has been coupled to the PEBBED code for neutronics, thermal-hydraulics and depletion analysis of the core. Two deep burn fuel types (Pu with or without initial MA fuel content) have been investigated with the new code system for normal and transient conditions including the effect of the statistical variation of thickness of the coating layers.

  18. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    . However, it is not that significant compared to the deviation from stoichiometry of oxygen due to the similar material properties of UO2 and PuO2.

  19. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  20. Study on Effect of MOX Fuel Assembly Loaded in Current M310 Reactor Core on Nuclear Design%MOX 燃料组件装入现役 M310堆芯对堆芯核设计的影响研究

    Institute of Scientific and Technical Information of China (English)

    刘晓黎; 宫宇

    2015-01-01

    国际上的 MOX 燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对 MOX 燃料组件的中子学性能进行了分析,对其在我国现役 M310堆芯应用的可行性进行了研究,得到了 M310堆芯由全部使用 UO2燃料组件向使用30%的 MOX 燃料组件过渡的堆芯燃料管理方案,并对使用MOX 燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的 MOX 燃料组件的堆芯可达到与全 UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler 功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为 MOX 燃料在 M310堆芯中应用的进一步研究提供参考。%The MOX fuel technology has been developed and applied in PWR all over the world .In this paper ,the neutronic performance of the MOX fuel assembly was studied , and fuel management scheme of M310 reactor core from all UO2 fuel assemblies to 30%MOX fuel assemblies was given .The results show that the core loaded 30% MOX fuel assemblies can reach the same lifetime as the all UO2 core ,the ability of the control system can meet the requirement of reactivity control ,and the Doppler temperature and power coefficients ,moderator temperature coefficient and the evolutions of Xe and Sm all benefit for the core operation to be more stable .The results of this study prove that the MOX fuel assembly can be used in the M310 reactor core .

  1. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Science.gov (United States)

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  2. Space and Time Distribution of Pu Isotopes inside The First Experimental Fuel Pin Designed for PWR and Manufactured in Indonesia

    Science.gov (United States)

    Suwardi; Setiawan, J.; Susilo, J.

    2017-01-01

    The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared and planned to be tested in power ramp irradiation. An irradiation test should be designed to allow an experiment can be performed safely and giving maximum results of many performance aspects of design and manufacturing. Performance analysis to the fuel specimen shows that the specimen is not match to be used for power ramp testing. Enlargement by 0.20 mm of pellet diameter has been proposed. The present work is evaluation of modified design for important aspect of isotopic Pu distribution during irradiation test, because generated Pu isotopes in natural UO2 fuel, contribute more power relative to the contribution by enriched UO2 fuel. The axial profile of neutrons flux have been chosen from both experimental measurement and model calculation. The parameters of ramp power has been obtained from statistical experiment data. A simplified and typical base-load commercial PHWR profile of LHR history has been chosen, to determine the minimum irradiation time before ramp test can be performed. The data design and Mat pro XI materials properties models have been chosen. The axial profile of neutrons flux has been accommodated by 5 slices of discrete pin. The Pu distribution of slice-4 with highest power rate has been chosen to be evaluated. The radial discretion of pellet and cladding and numerical parameter have been used the default best practice of TU. The results shows that Pu 239 increased rapidly. The maximum burn up of slice 4 at upper the median slice, it reached nearly 90% of maximum value at about 6000 h with peak of 0.8%a Pu/HM at 22000 h, which is higher than initial U 235. Each 240, 241 and 240 Pu grows slower and ends up to 0.4, 0.2 and 0.18 % respectively. This results can be used for verification of other aspect of fuel behavior in the modeling results and also can be used as guide and comparison to the future post irradiation examination for Pu isotopes distribution.

  3. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  4. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of

  5. Preliminary Reactor Physics Assessment of the HTR Module with 14% Enriched UCO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Hans D. Gougar

    2013-03-01

    The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500 µm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7 g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China. In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project. Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.

  6. Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

    Directory of Open Access Journals (Sweden)

    Andrius Slavickas

    2015-01-01

    Full Text Available The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2 and MOX fuels. UO2 fuel enrichment is increased above existing design limitations for LWR fuels (>5%. MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2 fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2 fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.

  7. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

    Full Text Available Neutronic performance is investigated for a potential accident tolerant fuel (ATF, which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod, and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

  8. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  9. A sensitivity study on DUPIC fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong

    1997-01-01

    In DUPIC fuel cycle, the spent pressurized water reactor (PWR) fuel is refabricated as a DUPIC fuel by a dry process. Because the spent PWR fuel composition depends on the initial enrichment and burnup condition of PWR fuel, the composition of a DUPIC fuel is not uniquely defined. Therefore, for the purpose of reducing the effects of such a composition heterogeneity on core performance, a composition adjustment of DUPIC fuel was studies. The composition adjustment was made in two steps: mixing two spent PWR fuel assemblies of higher and lower {sup 239}Pu contents and blending in fresh uranium with the mixed spent PWR fuels. Because the fuel and core performances depend on both the absolute amount of fissile isotopes and the ratio of major fissile isotope contents, a parametric study was performed to determine the reference compositions of {sup 235}U and {sup 239}Pu. The reference enrichments of {sup 235}U and {sup 239}Pu were determined such that the DUPIC core performance is comparable to that of a natural uranium core with high spent PWR fuel utilization and low fuel cycle cost. Under this condition, it is possible to utilize 90% of spent PWR fuels as the DUPIC fuel formula. On average, the amounts of slightly enriched and depleted uranium used for blending correspond to 8.6% and 10.6%, respectively, of the mass of candidate spent PWR fuels. (author). 16 refs., 30 tabs., 9 figs.

  10. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  11. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  12. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    Science.gov (United States)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  13. Radiation protection of MOX-fuel by doping with Pa-23 and U-232

    NARCIS (Netherlands)

    Kryuchkov, EF; Glebov, VB; Apse, VA; Shmelev, AN

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection at the "Spent Fuel Standard" level and more during full cycle of MOX-fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-fuel doped with Pa-231 and U-232. Acco

  14. 溶胶-凝胶法制备二氧化铀核芯的U3O8欠酸溶解工艺%U3O8 Acid-Deficient Dissolution Process for Preparation of UO2 Kernels by Sol-Gel Method

    Institute of Scientific and Technical Information of China (English)

    郝少昌; 周湘文; 赵兴宇; 马景陶; 王阳; 邓长生; 唐亚平

    2013-01-01

    UO2核芯的制备工艺中,为获得高浓度铀的硝酸铀酰溶液,同时降低溶液中硝酸根含量,即获得低于硝酸铀酰标准化学计量比(硝酸铀酰中硝酸根与铀酰离子化学计量比为2)的溶液,必须采用欠酸溶解工艺.本文采用分批加料和阶段性加热方式获得了很好的溶解效果,得到的硝酸铀酰溶液中U含量为2.1~2.5 mol/L,NO3和U的摩尔浓度比为1.6~1.8,溶液pH大于1.4,从而成功制备出合格的欠酸溶解的硝酸铀酰溶液.在此基础上,根据多次溶解试验的结果总结出了溶液密度与U含量的经验公式.%Uranyl nitrate solution with high uranium concentration and hypostoichiomet-ric c(NO3-)/c(U) ratio (the stoichiometric ratio of c (NO3- )/c (U) is 2 in standard uranyl nitrate solution) is crucial for preparation of qualified UO2 kernels for high temperature gas-cooled reactor. An acid-deficient dissolution process was developed for the preparation of uranyl nitrate solution. In the process the raw materials U3 O8 was added stepwise and the heating for the process was controlled in stages. Uranyl nitrate solutions with c(U) = 2. 1-2.5 mol/L, c(NO3- )/c(U)= 1. 6-1. 8 and PH>1. 4 could be prepared. A relationship between solution density and uranium concentration was summarized.

  15. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  16. Helium release from 238PuO2 fuel particles

    Science.gov (United States)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2000-01-01

    Coated plutonia fuel particles have recently been proposed for potential use in future space exploration missions that employ radioisotope power systems and/or radioisotope heater units (RHUs). The design of this fuel form calls for full retention of the helium generated by the natural radioactive decay of 238Pu, with the aid of a strong zirconium carbide coating. This paper reviews the potential release mechanisms of helium in small-grain (7-40 μm) plutonia pellets currently being used in the General Purpose Heat Source (GPHS) modules and RHUs, during both steady-state and transient heating conditions. The applicability of these mechanisms to large-grain and polycrystalline 238PuO2 fuel kernels is examined and estimates of helium release during a re-entry heating pulse up to 1723 K are presented. These estimates are based on the reported data for fission gas release from granular and monocrystal UO2 fuel particles irradiated at isothermal conditions up to 6.4 at.% burnup and 2030 K. It is concluded that the helium release fraction from large-grain (>=300 μm) plutonia fuel kernels heated up to 1723 K could be less than 7%, compared to ~80% from small-grain (7-40 μm) fuel. The helium release fraction from polycrystalline plutonia kernels fabricated using Sol-Gel techniques could be even lower. Sol-Gel fabrication processes are favored over powder metallurgy, because of their high precision and excellent reproducibility and the absence of a radioactive dust waste stream, significantly reducing the fabrication and post-fabrication clean-up costs. .

  17. Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Theodore E. Madey

    2001-10-01

    Characterize the effects of temperature and radiation processes on the interactions of H20 with oxide surfaces. Our experiments focused on the fundamental interaction of H20 molecules with surfaces of U02. We characterized the surface chemistry of adsorbed H2O using thermal desorption methods and radiotracer methods, as well as x-ray photoelectron spectroscopy (XPS) and low energy ion scattering (LEIS). In parallel with these measurements of thermal effects, we examined the effects of secondary electrons and high-energy photons on hydrogen and oxygen generation and, and how this related to corrosion of spent nuclear fuel. These studies concentrated on neutral and ionic (cation and anion) desorption products of low-energy electron irradiation of water-covered UO2.

  18. Fresh inflation with increasing cosmological parameter

    CERN Document Server

    Bellini, M

    2003-01-01

    I study a fresh inflationary model with an increasing F-cosmological parameter. The model provides sufficiently e-folds to solve the flatness/horizon problem and the density fluctuations agree with experimental values. The temperature increases during fresh inflation and reach its maximum value when inflation ends. I find that entropy perturbations always remain below $10^{-4}$ during fresh inflation and become negligible when fresh inflation ends. Hence, the adiabatic fluctuations dominate the primordial spectrum at the end of fresh inflation.

  19. Phytotoxicity of fresh and weathered diesel and gasoline to willow and poplar trees

    DEFF Research Database (Denmark)

    Trapp, Stefan; Köhler, A.; Larsen, L.C.

    2001-01-01

    The toxicity of fresh and weathered gasoline and diesel fuel to willow and poplar trees was studied using a tree transpiration toxicity test. Soils were taken from an abandoned filling station. Concentrations in the samples were measured as the sum of hydrocarbons from C5 to C10 (gasoline) and C1...

  20. Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

    Science.gov (United States)

    Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

    2012-11-30

    High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste.

  1. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    Science.gov (United States)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  2. Effect of Process Variables During the Head-End Treatment of Spent Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    K.J. Bateman; C.D. Morgan; J.F. Berg; D.J. Brough; P.J. Crane; D.G. Cummings; J.J. Giglio; M.W. Huntley; M.J. Rodriquez; J.D. Sommers; R.P. Lind; D.A. Sell

    2006-08-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. The effects of temperature, pressure, oxidative gas, and cladding have been studied with irradiated spent oxide fuel to determine the optimum conditions for process control. Experiments with temperatures ranging from 500oC to 1250oC have been performed on spent fuel using either air or oxygen gas for the oxidative reaction. Various flowrates and applications have been tested with the oxidative gases to discern the effects on the process. Tests have also been performed under vacuum conditions, following the oxidation cycle, at high temperatures to improve the removal of fission products. The effects of cladding on fission product removal have also been investigated with released fuel under vacuum and high temperature conditions. Results from these experiments will be presented as well as operating conditions based on particle size and decladding characteristics.

  3. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  4. Application of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL

    2012-01-01

    Presented is an investigation of the utilization of Tristructural-Isotropic (TRISO) particle-based fuel designs for the recycling of transuranic (TRU) wastes in typical Westinghouse four-loop pressurized water reactors (PWRs). Though numerous studies have evaluated the recycling of TRU in light water reactors (LWRs), this work differentiates itself by employing TRU-loaded TRISO particles embedded within a SiC matrix and formed into pellets that can be loaded into standard 17 x 17 fuel element cladding. This approach provides the capability of TRU recycling and, by virtue of the TRISO particle design, will allow for greater burnup (i.e., removal of the need for UO2 mixing) and improved fuel reliability. In this study, a variety of assembly layouts and core loading patterns were analyzed to demonstrate the feasibility of TRU-loaded TRISO fuel. The assembly and core design herein reported are a work in progress, so they still require some fine-tuning to further flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, TRU-loaded TRISO fuel and core designs that are capable of balancing TRU production and destruction can be designed within the standard constraints for thermal and reactivity performance in PWRs.

  5. Laser-based characterization of nuclear fuel plates

    Science.gov (United States)

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  6. Laser-Based Characterization of Nuclear Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  7. Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

    Science.gov (United States)

    Odorowski, Mélina; Jégou, Christophe; De Windt, Laurent; Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Martin, Christelle

    2016-01-01

    Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 109 Bq.gMOX-1 reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·tHM-1 after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O2] MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO2 reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO2 matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO2 grains were much more sensitive to oxidative dissolution, but the presence of carbonates did not enable observation of an oxidized layer by Raman spectroscopy with the exception of a few areas revealing the presence of U4O9. This data shows the heterogeneous nature of the alteration and the need to combine information from different techniques to determine the origin of releases.

  8. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

  9. Perspectives on the closed fuel cycle Implications for high-level waste matrices

    Science.gov (United States)

    Gras, Jean-Marie; Quang, Richard Do; Masson, Hervé; Lieven, Thierry; Ferry, Cécile; Poinssot, Christophe; Debes, Michel; Delbecq, Jean-Michel

    2007-05-01

    Nuclear energy accounts for 80% of electricity production in France, generating approximately 1150 t of spent fuel for an electrical output of 420 TWh. Based on a reprocessing-conditioning-recycling strategy, the orientations taken by Électricité de France (EDF) for the mid-term and the far-future are to keep the fleet performances at the highest level, and to maintain the nuclear option fully open by the replacement of present pressurized water reactor (PWR) by new light water reactor (LWR), such as the evolutionary pressurized reactor (EPR) and future Generation IV designs. Adaptations of waste materials to new requirements will come with these orientations in order to meet long-term energy sustainability. In particular, waste materials and spent fuels are expected to meet increased requirements in comparison with the present situation. So the treatment of higher burn-up UO2 spent fuel and MOX fuel requires determining the performances of glass and other matrices according to several criteria: chemical 'digestibility' (i.e. capacity of glass to incorporate fission products and minor actinides without loss of quality), resistance to alpha self-irradiation, residual power in view of disposal. Considering the long-term evolution of spent MOX fuel in storage, the helium production, the influence of irradiation damages accumulation and the evolution of the microstructure of the fuel pellet need to be known, as well as for the future fuels. Further, the eventual transmutation of minor actinides in fast neutron reactors (FR) of Generation IV, if its interest in optimising high-level waste management is proven, may also raise new challenges about the materials and fuel design. Some major questions in terms of waste materials and spent fuel are discussed in this paper.

  10. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  11. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Powers, Jeffrey J [ORNL; Maldonado, G Ivan [ORNL

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  12. US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

    2016-03-01

    The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

  13. Systematic Study of the Content of Phytochemicals in Fresh and Fresh-Cut Vegetables

    Directory of Open Access Journals (Sweden)

    María Isabel Alarcón-Flores

    2015-05-01

    Full Text Available Vegetables and fruits have beneficial properties for human health, because of the presence of phytochemicals, but their concentration can fluctuate throughout the year. A systematic study of the phytochemical content in tomato, eggplant, carrot, broccoli and grape (fresh and fresh-cut has been performed at different seasons, using liquid chromatography coupled to triple quadrupole mass spectrometry. It was observed that phenolic acids (the predominant group in carrot, eggplant and tomato were found at higher concentrations in fresh carrot than in fresh-cut carrot. However, in the case of eggplant, they were detected at a higher content in fresh-cut than in fresh samples. Regarding tomato, the differences in the content of phenolic acids between fresh and fresh-cut were lower than in other matrices, except in winter sampling, where this family was detected at the highest concentration in fresh tomato. In grape, the flavonols content (predominant group was higher in fresh grape than in fresh-cut during all samplings. The content of glucosinolates was lower in fresh-cut broccoli than in fresh samples in winter and spring sampling, although this trend changes in summer and autumn. In summary, phytochemical concentration did show significant differences during one-year monitoring, and the families of phytochemicals presented different behaviors depending on the matrix studied.

  14. Sintering of CaF 2 pellets as nuclear fuel analog for surface stability experiments

    Science.gov (United States)

    Godinho, José R. A.; Piazolo, Sandra; Stennett, Martin C.; Hyatt, Neil C.

    2011-12-01

    To enable a detailed study of the influence of microstructure and surface properties on the stability of spent nuclear fuel, it is necessary to produce analogs that closely resemble nuclear fuel in terms of crystallography and microstructure. One such analog can be obtained by sintering CaF 2 powder. This paper reports the microstructures obtained after sintering CaF 2 powders at temperatures up to 1240 °C. Pellets with microstructure, density and pore structure similar to that of UO 2 spent nuclear fuel pellets were obtained in the temperature range between 900 °C and 1000 °C. When CaF 2 was sintered above 1100 °C the formation of CaO at the grain boundaries caused the disintegration of the pellet due to hydration occurring after sintering. First results from a novel set-up of dissolution experiments show that changes in roughness, dissolution rate and etch pit shape of fluorite surfaces are strongly dependent on the crystallographic orientation of the expose surface. Consequently, the differences observed for each orientation will affect the overall dissolution rate and will lead to uncertainties in the estimation of dissolution rates of spent nuclear fuel.

  15. Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

    Science.gov (United States)

    Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

    2012-08-01

    TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

  16. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

  17. Swelling Mechanisms of UO2 Lattices with Defect Ingrowths.

    Directory of Open Access Journals (Sweden)

    Seçkin D Günay

    Full Text Available The swelling that occurs in uranium dioxide as a result of radiation-induced defect ingrowth is not fully understood. Experimental and theoretical groups have attempted to explain this phenomenon with various complex theories. In this study, experimental lattice expansion and lattice super saturation were accurately reproduced using a molecular dynamics simulation method. Based on their resemblance to experimental data, the simulation results presented here show that fission induces only oxygen Frenkel pairs while alpha particle irradiation results in both oxygen and uranium Frenkel pair defects. Moreover, in this work, defects are divided into two sub-groups, obstruction type defects and distortion type defects. It is shown that obstruction type Frenkel pairs are responsible for both fission- and alpha-particle-induced lattice swelling. Relative lattice expansion was found to vary linearly with the number of obstruction type uranium Frenkel defects. Additionally, at high concentrations, some of the obstruction type uranium Frenkel pairs formed diatomic and triatomic structures with oxygen ions in their octahedral cages, increasing the slope of the linear dependence.

  18. Pipe and grain boundary diffusion of He in UO2

    Science.gov (United States)

    Galvin, C. O. T.; Cooper, M. W. D.; Fossati, P. C. M.; Stanek, C. R.; Grimes, R. W.; Andersson, D. A.

    2016-10-01

    Molecular dynamics simulations have been conducted to study the effects of dislocations and grain boundaries on He diffusion in \\text{U}{{\\text{O}}2} . Calculations were carried out for the {1 0 0}, {1 1 0} and {1 1 1} edge dislocations, the screw dislocation and Σ5, Σ13, Σ19 and Σ25 tilt grain boundaries. He diffusivity as a function of distance from the dislocation core and grain boundaries was investigated for the temperature range 2300-3000 K. An enhancement in diffusivity was predicted within 20 Å of the dislocations or grain boundaries. Further investigation showed that He diffusion in the edge dislocations follows anisotropic behaviour along the dislocation core, suggesting that pipe diffusion occurs. An Arrhenius plot of He diffusivity against the inverse of temperature was also presented and the activation energy calculated for each structure, as a function of distance from the dislocation or grain boundary.

  19. Simulations of Xe and U diffusion in UO2

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders D. [Los Alamos National Laboratory; Vyas, Shyam [Los Alamos National Laboratory; Tonks, Michael R. [Idaho National Laboratory; Casillas, Luis [Los Alamos National Laboratory; Uberuaga, Blas P. [Los Alamos National Laboratory; Millett, Paul [Idaho National Laboratory

    2012-09-10

    Diffusion of xenon (Xe) and uranium (U) in UO{sub 2} is controlled by vacancy mechanisms and under irradiation the formation of mobile vacancy clusters is important. Based on the vacancy and cluster diffusion mechanisms established from density functional theory (DFT) calculations, we derive continuum thermodynamic and diffusion models for Xe and U in UO{sub 2}. In order to capture the effects of irradiation, vacancies (Va) are explicitly coupled to the Xe and U dynamics. Segregation of defects to grain boundaries in UO{sub 2} is described by combining the bulk diffusion model with models of the interaction between Xe atoms and vacancies with grain boundaries, which were derived from atomistic calculations. The diffusion and segregation models were implemented in the MOOSE-Bison-Marmot (MBM) finite element (FEM) framework and the Xe/U redistribution was simulated for a few simple microstructures.

  20. Management of spent fuel; Gestion del combustible irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Estrampes Blanch, J.

    2015-07-01

    The management of irradiated fuel has become one of the materials that more time and resources deals within their responsibilities that also cover other areas such as the design of the new cycles, supply of fresh fuel, tracking operation cycles and strategies of power changes. (Author)

  1. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  2. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  3. The Multi - vitamin Nutrient Fresh Vegetable Juice

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    @@ In the Tenth Five- Year Plan period, an important strategy for food industry is to develop fruit vegetable freshening and processing industry. Now,the consumable demand for vegetables turns to fresh,convenient, nutritious, safe and dean ones, while semi-processed vegetables and mixed fresh vegetable juices will meet this market demand exactly.

  4. A model for discrimination freshness of shrimp

    Directory of Open Access Journals (Sweden)

    Linong Du

    2015-12-01

    Full Text Available The shrimp is popular for its nutrition and dainty, however, it is easy to decay, and its freshness degrades, so, it is important to assess its freshness. The shrimp gives off unpleasant odor with its freshness change, detecting its odor difference can evaluate its freshness. The feasibility of using electronic nose for evaluating the freshness of shrimp (Penaeus vanmamei is explored in this paper. The odor of shrimp, stored at 5 °C, was detected by the electronic nose. Combined with the sensory evaluation and TVBN, a model based on the electronic nose was constructed to evaluate the shrimp freshness. In principal components analysis, the first three principal components accounted for 86.97% of total variation, and they are used to establish a model to estimate the shrimp freshness with Fisher Liner Discriminant. The discriminant rates were 98.3% for 120 modeling sample data, and 91.7% for 36 testing sample data. The model could be easily used to evaluate the freshness of shrimp with better accuracy.

  5. Consumers' store choice behavior for fresh food

    NARCIS (Netherlands)

    Meulenberg, M.T.G.; Trijp, van J.C.M.

    1991-01-01

    Consumers' preference for fresh food stores is analyzed. In particular the choice between supermarkets and specialized shops for purchasing fresh food is analyzed. Attention is given to the factors influencing this choice. For this purpose a number of research questions with respect to store choice

  6. Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

    Science.gov (United States)

    Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

    2010-03-01

    For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

  7. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  8. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  9. Hot Fuel Examination Facility/South

    Energy Technology Data Exchange (ETDEWEB)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs.

  10. 9 CFR 319.141 - Fresh pork sausage.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Fresh pork sausage. 319.141 Section... INSPECTION AND CERTIFICATION DEFINITIONS AND STANDARDS OF IDENTITY OR COMPOSITION Sausage Generally: Fresh Sausage § 319.141 Fresh pork sausage. “Fresh Pork Sausage” is sausage prepared with fresh pork or...

  11. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  12. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

    2016-12-01

    Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials of interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.

  13. Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2001-01-11

    Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

  14. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  15. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  16. Fission Product Removal From Spent Oxide Fuel By Head-End Processing

    Energy Technology Data Exchange (ETDEWEB)

    B. R. Westphal; K. J. Bateman; R. P. Lind; K. L. Howden; G. D. Del Cul

    2005-10-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. Experiments have been performed with irradiated oxide fuel to evaluate the removal of fission products. During these experiments, operating parameters such as temperature and pressure have been varied to discern their effects on the behavior of specific fission products. In general, the extent of removal increases with increasing operating temperature and decreasing pressure. Removal efficiencies as high as 98% have been achieved during testing. Given the results of testing, an explanation of the likely fission product species being removed during the test program is also provided. In addition, experiments have been performed with other oxidative gases (steam and ozone) on surrogates to determine their potential benefit for removal of fission products.

  17. Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

  18. Modelling global fresh surface water temperature

    NARCIS (Netherlands)

    Beek, L.P.H. van; Eikelboom, T.; Vliet, M.T.H. van; Bierkens, M.F.P.

    2011-01-01

    Temperature directly determines a range of water physical properties including vapour pressure, surface tension, density and viscosity, and the solubility of oxygen and other gases. Indirectly water temperature acts as a strong control on fresh water biogeochemistry, influencing sediment

  19. Modelling global fresh surface water temperature

    NARCIS (Netherlands)

    Beek, L.P.H. van; Eikelboom, T.; Vliet, M.T.H. van; Bierkens, M.F.P.

    2011-01-01

    Temperature directly determines a range of water physical properties including vapour pressure, surface tension, density and viscosity, and the solubility of oxygen and other gases. Indirectly water temperature acts as a strong control on fresh water biogeochemistry, influencing sediment concentrati

  20. Nitrogen uptake kinetics of freshly isolated zooxanthellae

    Digital Repository Service at National Institute of Oceanography (India)

    Wafar, M.V.M.; Wafar, S.; Rajkumar, R.

    Zooxanthellae freshly isolated from the coral host Pocillopora damicornis exhibited substrate-saturable uptake kinetics for ammonium, nitrate and urea. Maximum uptake velocity for ammonium [10.1 nmol. ( mu chl-a)./1h/1] was greater than...

  1. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

  2. Urban Fresh Water Resources Consumption of China

    Institute of Scientific and Technical Information of China (English)

    ZHU Peng; LU Chunxia; ZHANG Lei; CHENG Xiaoling

    2009-01-01

    From the point of view of urban consumption behavior, urban fresh water consumption could be classified as three types, namely, direct, indirect and induced water consumption. A calculation approach of urban fresh water consumption was presented based on the theory of urban basic material consumption and the input-output method, which was utilized to calculate urban fresh water consumption of China, and to analyze its structural change and causes. The results show that the total urban fresh water consumption increased 561.7×109m3, and the proportion to the total national fresh water resources increased by 20 percentage points from 1952 to 2005. The proportion of direct and induced water consumption had been continuously rising, and it increased by 15 and 35 percentage points separately from 1952 to 2005, while the proportion of indirect water consumption decreased by 50 percentage points. Urban indi-rect water consumption was mainly related to urban grain, beef and mutton consumption, and urban induced water consumption had a close relationship with the amount of carbon emission per capita. Finally, some countermeasures were put forward to realize sustainable utilization of urban fresh water resources in China.

  3. Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; Brian Boer; Gilles Youinou; Abderrafi M. Ougouag

    2011-03-01

    The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity

  4. Potentials for export of fresh raspberries from Serbia to EU fresh markets

    Directory of Open Access Journals (Sweden)

    Nikolić Ivan

    2012-12-01

    Full Text Available The aim of this paper is to present potentials of the most significant EU national markets for imports of fresh raspberries from Serbia. We carried out analysis of three markets with highest trade deficit of fresh raspberries expressed in quantities: Germany, Netherlands and Austria. The paper further analyses production and foreign trade trends in selected countries. According to results of this analysis, we identified monthly periods with highest potential for exports of fresh raspberries from Serbia to target markets. The paper also analyses wholesale prices of fresh raspberries and EU policy of direct support to raspberry producers.

  5. Estimates of helium gas release in 238PuO 2 fuel particles for radioisotope heat sources and heater units

    Science.gov (United States)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2000-06-01

    Release data of noble gases (Xe and Kr) from small-grain (7-40 μm), large-grain (⩾300 μm), and monocrystal UO 2 fuel particles, during isothermal irradiation up to 6.4 at.% and 2030 K are reviewed and their applicability to estimate helium release from 238PuO 2 fuel particles (⩾300 μm in diameter) is examined. Coated 238PuO 2 particles have recently been proposed for use in radioisotope power systems and heater units employed in planetary exploration missions. These fuel particles are intentionally sized and designed to prevent any adverse radiological effect and retain the helium gas generated by the radioactive decay of 238Pu, a desired feature for some planetary missions. Results suggest that helium release from large-grain (⩾300 μm) particles of K could be 80% but less than 7% at 1042 K, which is in general agreement with the experiments conducted at Los Alamos National Laboratory more than two decades ago. In these experiments, the helium gas release from small-grain (7-40 μm) 238PuO 2 fuel pellets has been measured during steady-state heating at temperatures up to 1886 K and ramp heating to 1723 K.

  6. Establishing the Global Fresh Water Sensor Web

    Science.gov (United States)

    Hildebrand, Peter H.

    2005-01-01

    This paper presents an approach to measuring the major components of the water cycle from space using the concept of a sensor-web of satellites that are linked to a data assimilation system. This topic is of increasing importance, due to the need for fresh water to support the growing human population, coupled with climate variability and change. The net effect is that water is an increasingly valuable commodity. The distribution of fresh water is highly uneven over the Earth, with both strong latitudinal distributions due to the atmospheric general circulation, and even larger variability due to landforms and the interaction of land with global weather systems. The annual global fresh water budget is largely a balance between evaporation, atmospheric transport, precipitation and runoff. Although the available volume of fresh water on land is small, the short residence time of water in these fresh water reservoirs causes the flux of fresh water - through evaporation, atmospheric transport, precipitation and runoff - to be large. With a total atmospheric water store of approx. 13 x 10(exp 12)cu m, and an annual flux of approx. 460 x 10(exp 12)cu m/y, the mean atmospheric residence time of water is approx. 10 days. River residence times are similar, biological are approx. 1 week, soil moisture is approx. 2 months, and lakes and aquifers are highly variable, extending from weeks to years. The hypothesized potential for redistribution and acceleration of the global hydrological cycle is therefore of concern. This hypothesized speed-up - thought to be associated with global warming - adds to the pressure placed upon water resources by the burgeoning human population, the variability of weather and climate, and concerns about anthropogenic impacts on global fresh water availability.

  7. Radiation protection potential of MOX-fuel doped with 231Pa and Cs radioisotopes.

    Science.gov (United States)

    Kryuchkov, E F; Glebov, V B; Apse, V A; Shmelev, A N

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection during full cycle of MOX-fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-fuel with 231Pa and Cs feed. As it follows from the paper results, combination of these two admixtures being doped into MOX-fuel is able to enhance the inherent radiation barrier and to weaken shortcomings of both proliferation deterrents.

  8. First Results of Scanning Thermal Diffusivity Microscope (STDM) Measurements on Irradiated Monolithic and Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    T. K. Huber; M. K. Figg; J. R. Kennedy; A. B. Robinson; D. M. Wachs

    2012-07-01

    The thermal conductivity of the fuel material in a reactor before and during irradiation is a sensitive and fundamental parameter for thermal hydraulic calculations that are useds to correctly determine fuel heat fluxes and meat temperatures and to simulate performance of the fuel elements during operation. Several techniques have been developed to measure the thermal properties of fresh fuel to support these calculations, but it is crucial to also investigate the change of thermal properties during irradiation.

  9. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  10. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  11. Fresh fruit: microstructure, texture, and quality

    Science.gov (United States)

    Wood, Delilah F.; Imam, Syed H.; Orts, William J.; Glenn, Gregory M.

    2009-05-01

    Fresh-cut produce has a huge following in today's supermarkets. The trend follows the need to decrease preparation time as well as the desire to follow the current health guidelines for consumption of more whole "heart-healthy" foods. Additionally, consumers are able to enjoy a variety of fresh produce regardless of the local season because produce is now shipped world-wide. However, most fruits decompose rapidly once their natural packaging has been disrupted by cutting. In addition, some intact fruits have limited shelf-life which, in turn, limits shipping and storage. Therefore, a basic understanding of how produce microstructure relates to texture and how microstructure changes as quality deteriorates is needed to ensure the best quality in the both the fresh-cut and the fresh produce markets. Similarities between different types of produce include desiccation intolerance which produces wrinkling of the outer layers, cracking of the cuticle and increased susceptibility to pathogen invasion. Specific examples of fresh produce and their corresponding ripening and storage issues, and degradation are shown in scanning electron micrographs.

  12. Preservation technologies for fresh meat - a review.

    Science.gov (United States)

    Zhou, G H; Xu, X L; Liu, Y

    2010-09-01

    Fresh meat is a highly perishable product due to its biological composition. Many interrelated factors influence the shelf life and freshness of meat such as holding temperature, atmospheric oxygen (O(2)), endogenous enzymes, moisture, light and most importantly, micro-organisms. With the increased demand for high quality, convenience, safety, fresh appearance and an extended shelf life in fresh meat products, alternative non-thermal preservation technologies such as high hydrostatic pressure, superchilling, natural biopreservatives and active packaging have been proposed and investigated. Whilst some of these technologies are efficient at inactivating the micro-organisms most commonly related to food-borne diseases, they are not effective against spores. To increase their efficacy against vegetative cells, a combination of several preservation technologies under the so-called hurdle concept has also been investigated. The objective of this review is to describe current methods and developing technologies for preserving fresh meat. The benefits of some new technologies and their industrial limitations is presented and discussed.

  13. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  14. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  15. Production of bio-jet fuel from microalgae

    Science.gov (United States)

    Elmoraghy, Marian

    The increase in petroleum-based aviation fuel consumption, the decrease in petroleum resources, the fluctuation of the crude oil price, the increase in greenhouse gas emission and the need for energy security are motivating the development of an alternate jet fuel. Bio-jet fuel has to be a drop in fuel, technically and economically feasible, environmentally friendly, greener than jet fuel, produced locally and low gallon per Btu. Bic jet fuel has been produced by blending petro-based jet fuel with microalgae biodiesel (Fatty Acid Methyl Ester, or simply FAME). Indoor microalgae growth, lipids extraction and transetrification to biodiesel are energy and fresh water intensive and time consuming. In addition, the quality of the biodiesel product and the physical properties of the bio-jet fuel blends are unknown. This work addressed these challenges. Minimizing the energy requirements and making microalgae growth process greener were accomplished by replacing fluorescent lights with light emitting diodes (LEDs). Reducing fresh water footprint in algae growth was accomplished by waste water use. Microalgae biodiesel production time was reduced using the one-step (in-situ transestrification) process. Yields up to 56.82 mg FAME/g dry algae were obtained. Predicted physical properties of in-situ FAME satisfied European and American standards confirming its quality. Lipid triggering by nitrogen deprivation was accomplished in order to increase the FAME production. Bio-jet fuel freezing points and heating values were measured for different jet fuel to biodiesel blend ratios.

  16. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  17. Modelling spent fuel and HLW behaviour in repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, A. M.; Esteban, J. A.

    2003-07-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  18. Consumer's Fresh Produce Food Safety Practices: Outcomes of a Fresh Produce Safety Education Program

    Science.gov (United States)

    Scott, Amanda R.; Pope, Paul E.; Thompson, Britta M.

    2009-01-01

    The Centers for Disease Control and Prevention estimate that there are 76 million cases of foodborne disease annually. Foodborne disease is usually associated with beef, poultry, and seafood. However, there is an increasing number of foodborne disease cases related to fresh produce. Consumers may not associate fresh produce with foodborne disease…

  19. Consumer's Fresh Produce Food Safety Practices: Outcomes of a Fresh Produce Safety Education Program

    Science.gov (United States)

    Scott, Amanda R.; Pope, Paul E.; Thompson, Britta M.

    2009-01-01

    The Centers for Disease Control and Prevention estimate that there are 76 million cases of foodborne disease annually. Foodborne disease is usually associated with beef, poultry, and seafood. However, there is an increasing number of foodborne disease cases related to fresh produce. Consumers may not associate fresh produce with foodborne disease…

  20. Microbiological Quality of Fresh Nopal Juice.

    Science.gov (United States)

    Hernández-Anguiano, Ana María; Landa-Salgado, Patricia; Eslava-Campos, Carlos Alberto; Vargas-Hernández, Mateo; Patel, Jitendra

    2016-12-10

    The consumption of fresh nopal cactus juice is widely popular among health-conscious consumers in Mexico. The juice is prepared from fresh cladodes that have only been rinsed with tap water and are not subjected to a pasteurization or terminal bacterial reduction process. The aim of this study was to evaluate the microbial quality of commercially available fresh juices (n = 162) made with nopal in Texcoco, State of Mexico, during the summer and spring season. Standard microbiological methods, the PCR technique and the serological method were used for isolation and identification of bacteria. All samples contained total coliforms and 91% were positive for Escherichia coli. Although total coliforms and E. coli were detected throughout the study, their populations were significantly lower (p nopal juices is unacceptable due to its health significance. The information generated in this study is relevant for human health risk assessment associated with the consumption of unpasteurized nopal juices and potential interventions to minimize pathogen contamination.

  1. Particular applications of food irradiation fresh produce

    Science.gov (United States)

    Prakash, Anuradha

    2016-12-01

    On fresh fruits and vegetables, irradiation at low and medium dose levels can effectively reduce microbial counts which can enhance safety, inhibit sprouting to extend shelf-life, and eliminate or sterilize insect pests which can serve to facilitate trade between countries. At the dose levels used for these purposes, the impact on quality is negligible. Despite the fact that regulations in many countries allow the use of irradiation for fresh produce, the technology remains under-utilized, even in the light of an increase in produce related disease outbreaks and the economic benefits of extended shelf life and reduced food waste. Putative concerns about consumer acceptance particularly for produce that is labeled as irradiated have deterred many companies from using irradiation and retailers to carry irradiated produce. This section highlights the commercial use of irradiation for fresh produce, other than phytosanitary irradiation which is covered in supplementary sections.

  2. SAGD processes with fresh water contact

    Energy Technology Data Exchange (ETDEWEB)

    Thimm, H.F. [Thimm Petroleum Technologies Inc. (Canada)

    2011-07-01

    In the Athabasca region, several bitumen reservoirs are shallow, located less than 400 meters below grade. These deposits are suitable for SAGD exploitation but the steam could come into contact with fresh water, which carries the risk of contaminating this resource. Operators are thus required by regulators to address this issue at the project application stage. The aim of this paper is to examine the potential effect of contact between fresh water and a bitumen bearing zone in a field in Northern Alberta. Investigations were conducted with a steam zone temperature of 200 degree Celsius and measurements were conducted at a plant close to the proposed project. Results showed that the accumulation of hydrogen sulphide would protect the water column and PAH, benzene and toluene were found to be potential concerns but they were not detected during implementation of a similar project. This paper demonstrated that the proposed project does not constitute a threat to fresh water.

  3. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  4. 9 CFR 319.142 - Fresh beef sausage.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Fresh beef sausage. 319.142 Section... Sausage § 319.142 Fresh beef sausage. “Fresh Beef Sausage” is sausage prepared with fresh beef or frozen beef, or both, but not including beef byproducts, and may contain Mechanically Separated (Species)...

  5. Fuel cells

    Directory of Open Access Journals (Sweden)

    D. N. Srivastava

    1962-05-01

    Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

  6. Future Fuels

    Science.gov (United States)

    2006-04-01

    Storage Devices, Fuel Management, Gasification, Fischer-Tropsch, Syngas , Hubberts’s Peak UNCLAS UNCLAS UNCLAS UU 80 Dr. Sujata Millick (703) 696...prices ever higher, and perhaps lead to intermittent fuel shortages as production fluctuates. Clearly, this competition for resources also provides oil...producers multiple options for selling their products, and raises the possibility that the US could face shortages resulting from shifts in

  7. Waste Reduction in Fresh Food Supply Chains

    DEFF Research Database (Denmark)

    Kaipia, Riikka; Loikkanen, Lauri; Dukovska-Popovska, Iskra

    2011-01-01

    The paper studies a well-known phenomenon, information sharing in supply chains, in a new context, fresh foods, with a specific goal, supporting sustainable performance in the supply chain. Fresh foods are important for retail stores, representing around half of retail sales, but form a challenging...... and heterogeneous group of products to manage. The value of the paper lies in its pointing out detailed solutions to how in real-life supply chains data can be used efficiently to improve the performance of the supply chain....

  8. Cultivable microbiome of fresh white button mushrooms.

    Science.gov (United States)

    Rossouw, W; Korsten, L

    2017-02-01

    Microbial dynamics on commercially grown white button mushrooms is of importance in terms of food safety assurance and quality control. The purpose of this study was to establish the microbial profile of fresh white button mushrooms. The total microbial load was determined through standard viable counts. Presence and isolation of Gram-negative bacteria including coagulase-positive Staphylococci were performed using a selective enrichment approach. Dominant and presumptive organisms were confirmed using molecular methods. Total mushroom microbial counts ranged from 5·2 to 12·4 log CFU per g, with the genus Pseudomonas being most frequently isolated (45·37% of all isolations). In total, 91 different microbial species were isolated and identified using Matrix-assisted laser desorption ionization-time of flight mass spectrophotometry, PCR and sequencing. Considering current food safety guidelines in South Africa for ready-to-eat fresh produce, coliform counts exceeded the guidance specifications for fresh fruit and vegetables. Based on our research and similar studies, it is proposed that specifications for microbial loads on fresh, healthy mushrooms reflect a more natural microbiome at the point-of-harvest and point-of-sale. Presence and persistence of micro-organisms within the microbiome of fresh produce is important when identifying a potential niche for foodborne pathogens. Most foodborne outbreaks can be attributed to microbial imbalances or lack of diversity within the associated host surface and residing microbial population. Agaricus bisporus samples analysed during this study showed a higher microbial load (5·2 up to 12·4 log CFU per g) compared to known values for other fresh produce. These mushrooms were considered to carry microbial loads representing a healthy and safe product, fit for consumption, despite showing a high indicator incidence. Although foodborne pathogens may be associated on occasion with fresh mushrooms, it remains a low

  9. Research on Logistics Mode of Fresh Agricultural Products in China

    OpenAIRE

    Xiao Hong; Qi Zhihui

    2013-01-01

    This study has a research on logistics mode of fresh agricultural products in china. Logistics costs are an important part of the price of fresh agricultural products. By researching the characteristics of fresh agricultural products, this study get the main reasons for the high price of fresh agricultural products in China, there are the high logistics cost, large loss and poor preservation in logistics process. Then some measures are proposed to reduce fresh agricultural products logistics ...

  10. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  11. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    Science.gov (United States)

    Li, Bo-Shiuan

    gas release. An optimum design is sought considering both thermal and mechanical models of this ceramic cladding with UO2 and advanced high density fuels.

  12. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  13. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    Science.gov (United States)

    Schanz, G.; Hagen, S.; Hofmann, P.; Schumacher, G.; Sepold, L.

    1992-06-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation ("flowering") and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ("quenching") are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

  14. Biofuels from the Fresh Water Microalgae Chlorella vulgaris (FWM-CV for Diesel Engines

    Directory of Open Access Journals (Sweden)

    Saddam H. Al-lwayzy

    2014-03-01

    Full Text Available This work aims to investigate biofuels for diesel engines produced on a lab-scale from the fresh water microalgae Chlorella vulgaris (FWM-CV. The impact of growing conditions on the properties of biodiesel produced from FWM-CV was evaluated. The properties of FWM-CV biodiesel were found to be within the ASTM standards for biodiesel. Due to the limited amount of biodiesel produced on the lab-scale, the biomass of dry cells of FWM-CV was used to yield emulsified water fuel. The preparation of emulsion fuel with and without FWM-CV cells was conducted using ultrasound to overcome the problems of large size microalgae colonies and to form homogenized emulsions. The emulsified water fuels, prepared using ultrasound, were found to be stable and the size of FWM-CV colonies were effectively reduced to pass through the engine nozzle safely. Engine tests at 3670 rpm were conducted using three fuels: cottonseed biodiesel CS-B100, emulsified cottonseed biodiesel water fuel, water and emulsifier (CS-E20 and emulsified water containing FWM-CV cells CS-ME20. The results showed that the brake specific fuel consumption (BSFC was increased by about 41% when the engine was fueled with emulsified water fuels compared to CS-B100. The engine power, exhaust gas temperature, NOx and CO2 were significantly lower than that produced by CS-B100. The CS-ME20 produced higher power than CS-E20 due to the heating value improvement as a result of adding FWM-CV cells to the fuel.

  15. Report on Evaluation, Verification, and Assessment of Porosity Migration Model in Fast Reactor MOX Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Novascone, Stephen Rhead [Idaho National Laboratory; Peterson, John William [Idaho National Laboratory

    2016-09-01

    Abstract This report documents the progress of simulating pore migration in ceramic (UO2 and mixed oxide or MOX) fuel using BISON. The porosity field is treated as a function of space and time whose evolution is governed by a custom convection-di?usion-reaction equation (described here) which is coupled to the heat transfer equation via the temperature field. The porosity is initialized to a constant value at every point in the domain, and as the temperature (and its gradient) are increased by application of a heat source, the pores move up the thermal gradient and accumulate at the center of the fuel in a time-frame that is consistent with observations from experiments. There is an inverse dependence of the fuel’s thermal conductivity on porosity (increasing porosity decreases thermal conductivity, and vice-versa) which is also accounted for, allowing the porosity equation to couple back into the heat transfer equation. Results from an example simulation are shown to demonstrate the new capability.

  16. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

  17. Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alice Cunha da; Su, Jian, E-mail: alicecs@poli.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot num