Sample records for fuel ceramic uo2

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

    Ristic, M.M.


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

  2. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    Bae, Gonghoon; Hong, Sergi


    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

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

    Novakovic, M.


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

  4. New UO2 fuel studies

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


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

  5. Ceramic UO2 powder production at Cameco Corporation

    Kwong, A.K.; Kuchurean, S.M.


    This presentation covers the various aspects of ceramic grade uranium dioxide (UO 2 ) powder production at Cameco Corporation and its use as fuel and blanket fuel for heavy-water and light-water reactors, respectively. In addition, it discusses the significant production variables that affect production and product quality. It also provides an insight into how various support groups such as Quality Assurance, Analytical Services, and Technology Development fit into the quality cycle and contribute to a successful operation. The ability of Cameco to identify, measure and control the physical and chemical properties of ceramic grade UO 2 has resulted in the production of uniform quality powder. This has meant that 100% of Cameco's ceramic grade UO 2 powder produced since mid-1989 has been accepted by the fuel manufacturers. (author)

  6. Fabrication of nano-structured UO2 fuel pellets

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


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

  7. Thoria-fuel irradiation. Program to irradiate 80% ThO2/20% UO2 ceramic pellets at the Savannah River Plant

    Pickett, J.B.


    This report describes the fabrication of proliferation-resistant thorium oxide/uranium oxide ceramic fuel pellets and preparations at the Savannah River Laboratory (SRL) to irradiate those materials. The materials were fabricated in order to study head end process steps (decladding, tritium removal, and dissolution) which would be required for an irradiated proliferation-resistant thorium based fuel. The thorium based materials were also to be studied to determine their ability to withstand average commercial light water reactor (LWR) irradiation conditions. This program was a portion of the Thorium Fuel Cycle Technology (TFCT) Program, and was coordinated by the Oak Ridge National Laboratory (ORNL) under the Consolidated Fuel Reprocessing Program (CFRP). The fuel materials were to be irradiated in a Savannah River Plant (SRP) reactor at conditions simulating the heat ratings and burnup of a commercial LWR. The program was terminated due to a de-emphasis of the TFCT Program, following completion of the fabrication of the fuel and the modified assemblies which were to be used in the SRP reactor. The reactor grade ceramic pellets were fabricated for SRL by Battelle, Pacific Northwest Laboratories. Five fuel types were prepared: 100% UO 2 pellets (control); 80% ThO 2 /20% UO 2 pellets; approximately 80% ThO 2 /20% UO 2 + 0.25 CaO (dissolution aid) pellets; 100% UO 2 hybrid pellets (prepared from sol-gel microspheres); and 100% ThO 2 pellets (control). All of the fuel materials were transferred to SRL from PNL and were stored pending a subsequent reactivation of the TFCT Programs

  8. Leaching of irradiated CANDU UO2 fuel

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


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

  9. Ceramic UO2 powder production at Cameco Corporation

    Mulligan, J.J.


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

  10. High density UO2 powder preparation for HWR fuel

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


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

  11. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    Tait, J.C.


    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129 I, 85 Kr and 14 C. (author). 104 refs., 9 tabs., 5 figs

  12. Fabrication of metallic channel-containing UO2 fuels

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


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

  13. Thermal expansion of UO2 and simulated DUPIC fuel

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


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

  14. Oxidative dissolution of ADOPT compared to standard UO2 fuel

    Nilsson, Kristina; Roth, Olivia; Jonsson, Mats


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

  15. A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

    Bae, Gonghoon; Hong, Ser Gi


    In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

  16. State of the art of UO2 fuel fabrication processes

    Henke, M.; Klemm, U.


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

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

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


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

  18. Correlation between fuel structure and mechanical properties of UO2

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


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

  19. Irradiation of UO2+x fuels in the TANOX device

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


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

  20. Fabrication of Cr-doped UO2 Fuel Pellet using Liquid Phase Sintering

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


    An enhancement of the thermal conductivity of a pellet can be obtained by the addition of a higher thermal conductive material in the pellet. In addition, the resistance to the PCI can be increased through a plasticity increase of the pellet. Thermal conductivity of ceramic materials is generally lower than that of metallic materials. The thermal conductivity of uranium oxide which is a typical ceramic material is low as well. The steep temperature gradient in the fuel pellet results from the low thermal conductivity. Therefore, the thermal conductivity improvement of a nuclear fuel pellet can enhance the fuel performance in various aspects. The lower centerline temperature of a fuel pellet affects the enhancement of fuel safety as well as fuel pellet integrity during nuclear reactor operation. Besides, the nuclear reactor power can be uprated due to the higher safety margin. So, many researches to enhance the thermal conductivity of nuclear fuel pellet have been performed in various ways. To improve the thermal conductivity of UO 2 pellet, an appropriate arrangement of the high thermal conductive material in UO 2 matrix is one of the various methods. We intended to control a placement of chromium as the high thermal conductive material. The metallic chromium and chromium oxide were arranged in a grain boundary of UO 2 using a liquid phase sintering method. The liquid phase sintering of Cr-doped UO 2 pellet could be adjusted using a control of an oxygen potential in sintering atmosphere

  1. Fabrication and testing of the sintered ceramic UO2 fuel - I - III, Part III - testing of sintered uranium dioxide properties dependent on the fabrication procedure

    Novakovic, M.; Ristic, M.M.


    The objective of this task was testing the influence of some parameters on the properties of sintered UO 2 . The influence of parameters tested were as follows: adhesives; pressure in the pressing procedure; temperature of sintering of the UO 2 powder. Other parameters were chosen according to the theoretical study. Sintering was done in argon atmosphere. Characterization of the UO 2 powder was performed meaning determining the needed chemical, physical and physico-chemical properties. Some new methods were developed within this task: SET method for measuring the specific surfaces, DTA, TGA, high-temperature torsion

  2. The preparation of UO2 ceramic microspheres with an advanced process (TGU)

    Xu Zhichang; Tang Yaping; Zhang Fuhong


    The UO 2 ceramic microspheres are the most important materials in the spherical fuel elements for high temperature reactor (HTR). An advanced process for preparation of UO 2 ceramic microspheres has been developed at Institute of Nuclear Energy Technology, Tsinghua University. This process known as total gelation process of uranium (TGU), is based on the traditional sol-gel process, external gelation process and internal gelation process of uranium (EGU and IGU), and has been selected as the production process. The result of batch test is described. Accordance with the requirements of quality control (QC) and quality assurance (QA), the stabilization of operating parameters and product quality is tested., The results on batch test have shown that as well as all of the operated parameters are fixed, then the product quality can be stable as well as the product specification can be met. When the colloidal flow rate and the vibration frequency of nozzle are fixed, the kernel's size is also fixed. When the sintering temperature and time are fixed, the product density is also fixed. When the hydrogen atmosphere is used, the O/U ratio is very near to stoichiometry. The performance and structure of UO 2 ceramic microspheres are also given

  3. Preparation of UO2 fragments for fuel-debris experiments

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


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

  4. Analysis of UO2 fuel structure for low and high burn-up and its impact on fission gas release

    Szuta, M.; El-Koliel, M.S.


    During irradiation, uranium dioxide (UO 2 ) fuel undergo important restructuring mainly represented by densification and swelling, void migration, equiaxed grain growth, grain subdivision, and the formation of columnar grains. The purpose of this study is to obtain a comprehensive picture of the phenomenon of equiaxed grain growth in UO 2 ceramic material. The change of the grain size in high-density uranium dioxide as a function of temperature, initial grain size, time, and burnup is calculated. Algorithm of fission gas release from UO 2 fuel during high temperature irradiation at high burnup taking into account grain growth effect is presented. Theoretical results are compared with experimental data. (author)

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

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


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

  6. Cracking and relocation of UO2 fuel during nuclear operation

    Appelhans, A.D.; Dagbjartsson, S.J.


    Cracking and relocation of light water reactor (LWR) fuel pellets affect the axial gas flow path within nuclear reactor fuel rods and the thermal performance of the fuel. As part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program, the Thermal Fuels Behavior Program of EG and G Idaho, Inc., is conducting fuel rod behavior studies in the Heavy Boiling Water Reactor in Halden, Norway. The Instrumental Fuel Assembly-430 (IFA-430) operated in that facility is a multipurpose assembly designed to provide information on fuel cracking and relocation, the long-term thermal response of LWR fuel rods subjected to various internal pressures and gas compositions, and the release of fission gases. This report presents the results of an analysis of fuel cracking and relocation phenomena as deduced from fuel rod axial gas flow and fuel temperature data from the first 6.5 GWd/tUO 2 burnup of the IFA-430

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

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


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

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

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


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

  9. Interim results from UO2 fuel oxidation tests in air

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.


    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

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

    Dias, Marcio S.


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

  11. Dose rate measurements in the beta-photon radiation field from UO2 pellets and glazed ceramics containing uranium

    Piesch, E.; Burgkhardt, B.


    In the nuclear fuel cycle, the handling of UO 2 pellets results in a significant exposure, mainly due to beta rays. Depth dose distributions have been investigated at source-to-detector distances of 5 to 80 cm using LiF detectors of different thicknesses. Detailed data for the dose equivalent quantities H(0.07), H(3) and H(10) are presented. These data are compared with those found for the use of glazed tiles and ceramics containing natural uranium. (author)

  12. Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

    Garcia, Chad B.; Brito, Ryan A.; Ortega, Luis H.; Malone, James P.; McDeavitt, Sean M.


    The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being ˜ 100 pct, or doubled, at 10 vol pct BeO.

  13. Thermal expansion of UO2-Gd2O3 fuel pellets

    Une, Katsumi


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

  14. The corrosion of spent UO2 fuel in synthetic groundwater

    Forsyth, R.S.; Werme, L.D.; Bruno, J.


    Leaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1-2 mg/l and 1 μg/l respectively. The leaching rate for Sr-90 decreased from about 10 -5 /d to 10 -7 /d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4. When reducing conditions were imposed on the pH 8.1 groundwater by means of H 2 /Ar in the presence of a Pd catalyst, significanly lower leach rates were attained. The hypothesis that alpha radiolytic decomposition of water is a driving force for UO 2 corrosion even under reducing conditions has been examined in leaching tests on low burnup (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected. Thermodynamically the calculated uranium solubilities in the pH range 4-8.2 generally agreed, well with the measured ones, although assumptions made for certain parameters in the calculations limit the validity of the results. (Author)

  15. Defect trap model of gas behaviour in UO2 fuel during irradiation

    Szuta, A.


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

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

    Hansson, L.


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

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

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


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

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

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


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

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

    Lucchini, J.F.


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

  20. Sintering of Kernel UO2 for High Temperature Reactor Fuel

    Sukarsono; Dwi-Heru-Sucahyo; Hidayati; Evi-Hertiviana; Bambang-Sugeng


    Sintering investigation of UO 2 gel has been done. The gel was preparedthrough two ways. The first, gel was produced using PVA as additive agent.The second gel was produced using HMTA and Urea as additive agent. From thepreparation of gel, the PVA method better than the urea - HMTA method,because was not necessary the cold temperature for sol preparation and alsowas not necessary the hot temperature for gelation process. After nextprocessing, the sintered gel of gel through PVA, also better than HMTAprocess. (author)

  1. Modelling the high burnup UO2 structure in LWR fuel

    Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.


    The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

  2. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    Linde, A. van der.


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

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

    Durazzo, Michelangelo; Riella, Humberto Gracher


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

  4. Statistical model for grain boundary and grain volume oxidation kinetics in UO2 spent fuel

    Stout, R.B.; Shaw, H.F.; Einziger, R.E.


    This paper addresses statistical characteristics for the simplest case of grain boundary/grain volume oxidation kinetics of UO 2 to U 3 O 7 for a fragment of a spent fuel pellet. It also presents a limited discussion of future extensions to this simple case to represent the more complex cases of oxidation kinetics in spent fuels. 17 refs., 1 fig

  5. Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

    Silva, A.T.; Marra Neto, A.


    UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

  6. Improving the Thermal Conductivity of UO2 Fuel with the Addition of Graphene

    Cho, Byoung Jin; Kim, Young Jin; Sohn, Dong Seong


    Improvement of fuel performances by increasing the fuel thermal conductivity using the BeO or W were reported elsewhere. In this paper, some major fuel performances of improved thermal conductivity oxide (ICO) nuclear fuel with the addition of 10 v/o graphene have been compared to those of standard UO 2 fuel. The fuel thermal conductivity affects many performance parameters and thus is an important parameter to determine the fuel performance. Furthermore, it also affects the performance of the fuel during reactor accidents. The improved thermal conductivity of the fuel would reduce the fuel temperature at the same power condition and would improve the fission gas release, rod internal pressure and fuel stored energy. Graphene is well known for its excellent electrical conductivity, strength and thermal conductivity. The addition of graphene to the UO 2 fuel could increase the thermal conductivity of the ICO fuel. Although the graphene material is extensively studied recently, the characteristics of the graphene material, especially the thermal properties, are not well-known yet. In this study, we used the Light Water Reactor fuel performance analysis code FRAPCON-3.2 to analyze the performance of standard UO 2 and ICO fuel

  7. Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments

    Merino, J.; Cera, E.; Bruno, J.; Quinones, J.; Casas, I.; Clarens, F.; Gimenez, J.; Pablo, J. de; Rovira, M.; Martinez-Esparza, A.


    Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO 2 oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO 2 has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO 2 , particularly the role of · OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions

  8. Achieving higher productivity of UO2 fuel at NUOFP through improved in-plant quality surveillance

    Meena, R.; Pramanik, D.; Sairam, S.; Rajkumar, J.V.; Rao, R.V.R.L.V.; Sinha, T.K.; Santra, N.; Rao, G.V.S.H.; Jayaraj, R.N.


    At Nuclear Fuel Complex (NFC), in the production of UO 2 fuel for PHWRs, a standard set of process parameters are monitored regularly for every lot of powder and pellet. Quality of intermediate products in the production process like UNP, ADU(dry), U 3 O 8 , UO 2+x , UO 2 granules, green pellets, sintered pellets are also regularly analysed/monitored apart from the final finished pellet and ensured to be within specified range. This range is decided by final product specifications and sometimes also based on the feed requirement in the next process in the downstream of the flow sheet. Vast experience gained over the years, behavior of various equipment under given set of conditions, feed back from the customer plants etc; have been primary criteria hither to, for defining the process conditions and chemical/physical properties of intermediate products

  9. Technical evaluation of the direct denitration process to obtain ceramic-grade UO2 powders using microwaves

    Lorenzo, Viviana J.; Marchi, Daniel E.; Menghini, Jorge E.


    The direct denitration process to obtain ceramic-grade UO 2 powders using microwaves has been studied and developed at laboratory scale. Conditions were given to obtain powders apt for fuel pellets fabrication within the required specifications, where mechanical treatments before pressing are not necessary. This work describes the equipment used in the process, evaluates the necessary supply and waste generation and describes the characteristics of the product obtained, as well as the conditions for its fabrication. Results show that this method allows to reduce the volume of liquid wastes generated due to their partial re-utilization, simplifying their final disposal treatment, which, in addition to their operational advantages, make this method attractive from the economical point of view. (author)

  10. Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

    Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.


    The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

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

    Hofmann, P.


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

  12. Methods of modification and investigations of properties of fuel UO2

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


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

  13. A study of the effectiveness of hand protection when handling UO2 fuel pellets

    Washington, R.R.; Sullivan, D.F.


    Simple tests were performed to estimate the effectiveness of various forms of hand protection in reducing skin doses when handling UO 2 fuel pellets. Household rubber gloves (rubberized cotton) appeared to be the most effective of the varieties tested. Nylon gloves and latex finger cots were least effective. (author)

  14. Interaction between UO2 kernel and pyrocarbon coating in irradiated and unirradiated HTR fuel particles

    Drago, A.; Klersy, R.; Simoni, O.; Schrader, K.H.


    Experimental observations on unidirectional UO 2 kernel migration in TRISO type coated particle fuels are reported. An analysis of the experimental results on the basis of data and models from the literature is reported. The stoichiometric composition of the kernel is considered the main parameter that, associated with a temperature gradient, controls the unidirectional kernel migration

  15. On the correlation between fuel structure and mechanical properties of UO2

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


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

  16. Physics of the fuel cycle. Evaluation of methods, uncertainties and estimation of the material balance for fuels UO2 and UO2-PuO2

    Monier, C.


    The research works of this thesis are aimed to evaluate the methods and the associated uncertainties for the material balances estimation of the burn-up UO 2 and MOX fuels which intervene in the fuel cycle physics. The studies carried out are used to qualify the cycle 'package' DARWIN for the PWRs material balances estimation. The elaboration and optimisation of the calculation routes are carried out following a very specific methodology, aimed at estimating the bias introduced by the modelizations simplification by a comparison with almost exact reference modelizations. Depending on the precision goals and the informations, the permissible approximation will be determined. Two calculation routes have been developed and the qualified by applying them to the used fuels isotopic analysis interpretation: one 'industry-oriented' calculation route which can calculate full UO 2 assemblies material balances with a 2 % precision on the main actinides, respecting the industrial specifications. This route must run with a reasonable calculation time and stay user-friendly; one reference calculation route for the precise interpretation of fuel samples made of pieces of burn-up MOX rods. Aiming to provide material balances with the best possible precision, this route does not have the same specifications concerning its use and its calculation time performance. (author)

  17. Ceramics as nuclear reactor fuels

    Reeve, K.D.


    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  18. Inspection of the UO2 special fuel for the prototype heavy water reactor 'FUGEN'

    Miura, Makoto; Ohmori, Takuro; Yoshino, Hiroyuki; Matsui, Hiromasa; Hirosawa, Naonori


    UO 2 special fuel assemblies are the fuel for material irradiation incorporating irradiation specimens, for the prototype heavy water reactor ''FUGEN''. In order to monitor the behavior of the pressure tube material irradiated with neutrons for long time, monitoring specimens were equipped in the core. This special fuel was fabricated by the Nuclear Fuel Industries, Ltd. (NFI), and the fuel cladding tubes, the capsule guide tubes and the capsule tubes were furnished by PNC. The irradiation specimens were prepared by PNC, and incorporated into the assemblies by NFI. The inspection by PNC on the special fuel assemblies was conducted following the inspection by the maker, which was made on UO 2 pellets, fuel element and assembly parts except cladding tubes, after completing the fabrication. The specifications of the special fuel, especially for the outer and inner layer pellets, the outer and inner layer fuel elements and the fuel assemblies, are presented. The flow sheet for the inspection process and surveillance test of special fuel assemblies is illustrated. The inspection items, the materials and the quantity inspection are tabulated for the fuel elements, the fuel assemblies and the irradiation capsules, respectively. The structure of a special type fuel assembly is shown. For each inspection, the inspection methods and items and the results are explained. As for the results of inspection of the special fuel, the UO 2 pellets, fuel element parts, fuel elements, fuel assembly parts, fuel assemblies, capsules and irradiation specimens were in accordance with the specifications. Regarding the situation of the quality control in the processes, check was made with many documents, and it was recognized that the quality control was performed in the quality assurance program. (Nakai, Y.)

  19. Fission gas release and grain growth in THO2-UO2 fuel irradiated at high temperature

    Goldberg, I.; Waldman, L.A.; Giovengo, J.F.; Campbell, W.R.


    Data are presented on fission gas release and grain growth in ThO 2 -UO 2 fuels irradiated as part of the LWBR fuel element development program. These data for rods that experienced peak linear power outputs ranging from 15 to 22 KW/ft supplement fission gas release data previously reported for 51 rods containing ThO 2 and ThO 2 -UO 2 fuel irradiated at peak linear powers predominantly below 14 KW/ft. Fission gas release was relatively high (up to 15.0 percent) for the rods operated at high power in contrast to the relatively low fission gas release (0.1 to 5.2 percent) measured for the rods operated at lower power. Metallographic examination revealed extensive equiaxed grain growth in the fuel at the high power axial locations of the three rods

  20. Sensitivity and uncertainty analysis for UO2 and MOX fueled PWR cells

    Foad, Basma; Takeda, Toshikazu


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

  1. The corrosion of spent UO2-fuel in synthetic groundwater

    Forsyth, R.S.; Svanberg, K.; Werme, L.


    Segments of fuel and clad have been leached in deionized water and in groundwater. The leachants were centrifuged through membrane filters. Both centrifugate and the filters were analysed for U, Sr-90, α- and γ-emitters. The results are discussed in terms of preferential leaching, solubility limitations and adsorption effects. For U an apparent saturation at about 800 ppb was observed. Pu also appeared to attain saturation at a few ppb. For Sr the leach rate was 3x10 -7 /d after ca 400 days. Attempts to impose reducing conditions showed decreased leach rates. (Authors)

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

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


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

  3. A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

    Notley, M.J.F.; Hastings, I.J.


    A model for the release of fission gas from irradiated UO2 fuel is presented. It incorporates fission gas diffusion bubble and grain boundary movement,intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO2 fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW/m, burnups between 10 and 300 MW.h/kg U and power histories including constant, high-to-low and low-to-high power periods. The predictions of the model are shown to be most sensitive to fuel power (temperature), the selection of diffusion coefficient for fission gas in UO2 and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth. (author)

  4. Performance evaluation of UO2-Zr fuel in power ramp tests

    Knudsen, P.; Bagger, C.


    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

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

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


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

  6. Microprobe analysis of PuO2--UO2 nuclear fuel

    Clark, W.I.; Rasmussen, D.E.; Carlson, R.L.; Highley, D.M.


    For the preirradiation characterization of FFTF UO 2 --PuO 2 fuel, a program was developed to determine the preirradiation porosity, grain structure, and microcomposition of the fuel. Two computer programs, MITRAN and MERIT, were developed to evaluate the homogeneity of the fuel. These programs use elemental composition data generated by the electron microprobe. MITRAN determines information on the size and frequency of individual regions, whereas MERIT provides an index of the thermal performance of the fuel and calculated statistical data for comparison to other fuel batches

  7. Influence of radiolysis on UO2 fuel matrix dissolution under disposal conditions. Literature Study

    Ollila, K.


    The objective of this study was to examine the recent published literature on the influence of water radiolysis on UO 2 fuel matrix dissolution under the disposal conditions. The α radiation is considered to be dominating over the other types of radiations at times longer than 1000 years. The presence of the anaerobic corrosion products of iron, especially of hydrogen, has been observed to play an important role under radiolysis conditions. It is not possible to exclude gamma/beta radiolysis effects in the experiments with spent fuel, since there is not available a fuel over 100 years old. More direct measurements of α radiolysis effects have been conducted with α doped UO 2 materials. On the basis of the results of these experiments, a specific activity threshold to observe α radiolysis effects has been presented. The threshold is 1.8 x 10 7 to 3.3 x 10 7 Bq/g in anoxic 10 -3 M carbonate solution. It is dependent on the environmental conditions, such as the reducing buffer capacity of the conditions. The results of dissolution rate measurements at VTT with 233 U-doped UO 2 samples in 0.01 to 0.1 M NaCl solutions under anoxic conditions did not show any effect of α radiolysis with doping levels of 5 and 10% 233 U (3.2 x 10 7 and 6.3 x 10 7 Bq/g). Both Fe 2+ and hydrogen can act as reducing species and could react with oxidizing radiolytic species. Fe 2+ concentrations of the order of 10 -5 M can decrease the rate of H 2 O 2 production. Low dissolution rates, 2 x 10 -8 to 2 x 10 -7 /yr, have been measured in the presence of metallic Fe with 5 and 10% 233 U-doped UO 2 in 0.01 to 1 M NaCl solutions. The tests with isotope dilution method showed precipitation phenomena of U to occur during dissolution process. The concentrations of dissolved U were extremely low (≤ 8.4 x 10 -11 M). No effects of -radiolysis could be seen. It is difficult to distinguish the effects of metallic Fe, Fe 2+ or hydrogen in these tests. Hydrogen could also act as a reducing agent

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

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


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

  9. The fabrication process of ceramic grade UO2 powder via fluorid system AUC and the treatment on AUC precipitation filtrate

    Liu Jinhong; Xu Kui; Li Zhiwan; Yi Wei; Tang Yueming; Li Guangrong; Lei Maolin; Cui Chuanjiang


    It is described about the technology of fabricating AUC powder by Circum-fluence Precipitation Reactor with Gas (CPRG) from UF 6 hydrolyzed liquid, manufacturing nuclear pure ceramic grade UO 2 powder via fluorid system AUC process with fluidized bed method, recovering U(VI) with ion exchange resin, depositing fluorin in an outflow of effusion wastewater from the ion exchange using calces. The primary control parameters on the fabricating AUC powder is study, it is discussed to character difference of AUC powder between fluorid system and nitrate. Result show that the composing the manufacture AUC powder is invariable by CORG, and that the AUC quality is consistent, and that by decomposition and reduction of AUC and stabilization of UO 2 powder with fluidized bed, through optimum technological parameters, the excellent UO 2 powder is obtained on the quality. (authors)

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

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


    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 CERMET

  11. Physical and chemical characterization of the (Th, U)O2 mixed oxide fuel

    Santos, A.M.M. dos; Avelar, M.M.; Palmieri, H.E.L.; Lameiras, F.S.; Ferreira, R.A.N.


    The NUCLEBRAS R and D Center (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN) has been performing, together with german institutions (Kernforschungsanlage Julich GmbH - KFA, Krafwerk Union A.G. - KWU and NUKEM GmbH), a program for utilization of thorium in pressurized water reactors. In this paper are presented the physical and chemical characterizations necessary to quality the (Th, U)O 2 fuel and the respective methods. (Author) [pt

  12. Physical characteristics of Gd2O3-UO2 fuel in LWR

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


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

  13. Effects of hyperstoichiometry and fission products on the electrochemical reactivity of UO2 nuclear fuel

    Betteridge, J.S.; Scott, N.A.M.; Shoesmith, D.W.; Bahen, L.E.; Hocking, W.H.; Lucuta, P.G.


    The effects of hyperstoichiometry and fission products on the electrochemical reactivity Of UO 2 nuclear fuel have been systematically investigated using cyclic voltammetry and the O 2 reduction reaction. Significant constraints are placed on the active-site model for O 2 reduction by the modest impact of bulk hyperstoichiometry. Formation of the U 4 O 9 derivative phase was associated with a marked increase in transient surface oxidation/reduction processes, which probably involve localized attack and might be fostered by tensile stresses induced during oxidation. Electrocatalytic reduction Of O 2 on simulated nuclear fuel (SIMFUEL) has been determined to increase progressively with nominal burnup and pronounced enhancement of H 2 O reduction has been observed as well. Substitution of uranium by lower-valence (simulated) fission products, which was formerly considered the probable cause for this behaviour, has now been shown to merely provide good electrical conductivity. Instead, the enhanced reduction kinetics for O 2 and H 2 O on SIMFUEL can be fully accounted for by noble metals, which segregate to the UO 2 grain boundaries as micron-sized particles, despite their low effective surface area. Apparent convergence of the electrochemical properties Of UO 2 and SIMFUEL through natural corrosion likely reflects evolution toward a common active surface. (author)

  14. A review of the thermophysical properties of MOX and UO2 fuels

    Carbajo, Juan J.; Yoder, Gradyon L.; Popov, Sergey G.; Ivanov, Victor K.


    A critical review of the thermophysical properties of UO 2 and MOX fuels has been completed, and the best correlations for thermophysical properties have been selected. The properties reviewed are solidus and liquidus temperatures of the uranium/plutonium dioxide system (melting and solidification temperatures), thermal expansion and density, enthalpy and specific heat, enthalpy (or heat) of fusion, and thermal conductivity. Only fuel properties have been reviewed. The selected set of property correlations was compiled to be used in thermal-hydraulic codes to perform safety calculations

  15. Fission gas release from UO2 pellet fuel at high burn-up

    Vitanza, C.; Kolstad, E.; Graziani, U.


    Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

  16. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    Lawrence, L.A.


    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  17. Completion of UO2 pellets production and fuel rods load for the RA-8 critical facility

    Marajofsky, Adolfo; Perez, Lidia E.; Thern, Gerardo G.; Altamirano, Jorge S.; Benitez, Ana M.; Cardenas, Hugo R.; Becerra, Fabian A.; Perez, Aldo E.; Fuente, Mariano de la


    The Advanced Fuels Division produced fuel pellets of 235 U with 1.8% and 3.6% enrichment and Zry-4 cladding loads for the RA-8 reactor at Pilcaniyeu Technological Unit. For economical and availability reasons, the powder acquired was initially UO 2 with 3.4% enrichment in 235 U, therefore the 235 U powder with 1.8% enrichment was produced by mechanical mixture. The production of fuel pellets for both enrichments was carried out by cold pressing and sintering processes in reducing atmosphere. The load of Zry-4 claddings was performed manually. The production stages can be divided into setup, qualification and production. This production allows not only to fulfill satisfactorily the new fuel rods supply for the RA-8 reactor but also to count with a new equipment and skilled personnel as well as to meet quality and assurance control methods for future pilot-scale production and even new fuel elements production. (author)

  18. Behaviour in air at 175-400 degrees C of irradiated UO2 fuel

    Hastings, I.J.; McCracken, D.


    The authors extended their study of irradiated, defected UO 2 fuel elements to 200 and 400 degrees C. At 200 degrees C there was no diametral change, but at 400 degrees C we observed swelling and severe sheath splitting. Neither short-lived fission products, nor Cs-134, Cs-137 or Ru-106 above background, were detected. Maximum Kr-85 release was 4 Bq ( -6 Ci). Discharge time was 2.5 years. UO 2 fragment studies were extended to 400 degrees C. The oxidation process for unirradiated and irradiated fuel up to 300 degrees C was characterized by activation energies of 140 +- 10 and 120 +- 10 kJ/mol, respectively; enhancement of oxidation rate was confirmed in the irradiated samples. There is an apparent reduction of activation energy above about 300 degrees C. Fuel elements with artificial and natural defects showed similar oxidation and dimensional response at 250 degrees C. Behaviour of fuel fragments from the defect area of a naturally-defected element is consistent with that for fragments from intact elements when prior oxidation during the defect period is considered

  19. Irradiation of defected SAP clad UO2 fuel in the X-7 organic loop

    Robertson, R.F.S.; Cracknell, A.G.; MacDonald, R.D.


    This report describes an experiment designed to test the behaviour under irradiation of a UO 2 fuel specimen clad in a defected SAP sheath and cooled by recirculating organic liquid. The specimen containing the defect was irradiated in the X-7 loop in the NRX reactor from the 25th of November until the 13th of December 1960. Up to the 13th of December the behaviour was analogous to that seen with defected UO 2 specimens clad in zircaloy which were irradiated in water loops. Reactor power transients resulted in peaking of gamma ray activities in the loop, but on steady operation these activities tended to fall to a steady state level, Over this period the pressure drop across the fuel increased by a factor of two, the increases occurring after reactor shut downs and start ups. On 13th December the pressure drop increased rapidly, after a reactor shut down and start up, to over five times its original value and the activities in the loop rose to a high level. The specimen was removed and examination showed that the sheath was very badly split and that the volume between the fuel and the sheath was filled with a hard black organic substance. This report gives full details of the irradiation and of the post -irradiation examination. Correlation of the observed phenomenon is attempted and a preliminary assessment of the problems which would be associated with defect fuel in an organic reactor is given. (author)

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

    Ollila, K.


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

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

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


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

  2. Release of tellurium and cesium from UO2 in LWR fuel rods during irradiation

    Malen, K.A.


    In this paper the release of tellurium (Te-132) and cesium (Cs-134 and Cs-137) from UO 2 -fuel is analyzed. The basis for the analysis is the experimental results from the S176 series of experiments performed at Studsvik. It seems that the model developed earlier for release of iodine applies also to tellurium and cesium. This model assumes sweeping up of the species in question by moving grain boundaries and subsequent release through grain boundary porosity. An interesting extra feature is deposition of tellurium at temperatures in the range 1500-2000 K believed to be due to condensation. (author)

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

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


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

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

    Sabioni, A.C.S.


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

  5. Gaseous swelling of B4C and UO2 fuel: similarities and differences

    Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.


    A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

  6. Studies on the Sintering Behaviour of UO2-Gd2O3 Nuclear Fuel

    Durazzo, Michelangelo; Gracher Riella, Humberto


    The incorporation of gadolinium directly into nuclear power reactor fuel is important from the point of reactivity compensation and adjustment of power distribution enabling thus longer fuel cycles and optimized fuel utilization. The incorporation of Gd 2 O 3 powder directly into the UO 2 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. This is due to blockages during the sintering process. There is little information in published literature about the possible mechanism for this blockage and this is restricted to the hypothesis based on formation of a low diffusivity Gd rich (U,Gd)O 2 phase. Experimental evidences indicated the existence of phases in the (U,Gd)O 2 system with structure different from the fluorite type structure of UO 2 . The apparition of these new phases coincides with the lowering of the density after sintering and with the lowering of the interdiffusion coefficient. However, it has been shown experimentally that the sintering blockage phenomena cannot be explained on the basis of the formation of low diffusivity Gd rich (U,Gd)O 2 phases. The work was continued to investigate other possible blocking mechanism. (authors)

  7. Pressure analysis in the fabrication process of TRISO UO2-coated fuel particle

    Liu Malin; Shao Youlin; Liu Bing


    Highlights: ► The pressure signals during the real TRISO UO2-coated fuel particle fabrication process. ► A new relationship about the pressure drop change and the coated fuel particles properties. ► The proposed relationship is validated by experimental results during successive coating. ► A convenient method for monitoring the fluidized state during coating process. - Abstract: The pressure signals in the coating furnace are obtained experimentally from the TRISO UO 2 -coated fuel particle fabrication process. The pressure signals during the coating process are analyzed and a simplified relationship about the pressure drop change due to the coated layer is proposed based on the spouted bed hydrodynamics. The change of pressure drop is found to be consistent with the change of the combination factor about particle density, bed density, particle diameter and static bed height, during the successive coating process of the buffer PyC, IPyC, SiC and OPyC layer. The newly proposed relationship is validated by the experimental values. Based on this relationship, a convenient method is proposed for real-time monitoring the fluidized state of the particles in a high-temperature coating process in the spouted bed. It can be found that the pressure signals analysis is an effective method to monitor the fluidized state on-line in the coating process at high temperature up to 1600 °C.

  8. A comparative study of fission gas behaviour in UO2 and MOX fuels using the meteor fuel performance code

    Struzik, C.; Garcia, Ph.; Noirot, L.


    The paper reviews some of the fission-gas-related differences observed between MOX MIMAS AUC fuels and homogeneous UO 2 fuels. Under steady-state conditions, the apparently higher fractional release in MOX fuels is interpreted with the METEOR fuel performance code as a consequence of their lower thermal conductivity and the higher linear heat rates to which MOX fuel rods are subjected. Although more fundamental diffusion properties are needed, the apparently greater swelling of MOX fuel rods at higher linear heat rates can be ascribed to enhanced diffusion properties. (authors)

  9. A study of UO2 wafer fuel for very high-power research reactors

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.


    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  10. Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

    Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.


    The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

  11. UO2 - Zr chemical interaction of PHWR fuel pins under high temperature

    Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.


    At high temperature Zircaloy clad interacts with the UO 2 fuel as well as with the steam to produce oxide layer of a-Zr(O) and ZrO 2 . This layer formation significantly reduces the structural strength of the clad. A computer code SFDCPA/MOD1 has been developed to simulate the interaction and predict the oxide layer thickness for any accidental transient condition. It is well validated with published experimental data on the isothermal and transient temperature condition. The program is applied to Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under certain severe transient condition where it experiences temperature above 1000 C. The study gives an idea of the un-oxidized thickness of Zircaloy, which is an important criterion for fuel integrity. (author)

  12. Recent findings on the oxidation of UO2 fuel under nominally dry storage conditions

    Taylor, P.; McEachern, R.J.; Sunder, S.; Wasywich, K.M.; Miller, N.H.; Wood, D.D.


    This paper is an overview of fuel-storage demonstration experiments, supporting research on UO 2 oxidation, and associated model development, in progress at AECL's Whiteshell Laboratories. The work is being performed to determine the time/temperature limits for safe storage of irradiated CANDU fuel in dry air. The most significant recent experimental finding has been the detection of small quantities of U 3 O 8 , formed over periods of one to several years in a variety of experiments at 150-170 deg C. Another important trading is the slight suppression of U 3 O 8 formation in SIMFUEL and other doped U0 2 formulations. The development of a nucleation-and-growth model for U 3 O 8 formation is discussed, along with available activation energy data. These provide a basis for predicting U 3 O 8 formation rates under dry-storage conditions, and hence optimizing fuel storage strategies. (author)

  13. Thermal Expansion and Density Data of UO2 and Simulated Fuel for Standard Reference

    Yang, Jae Hwan; Na, S. H.; Lee, J. W.; Kang, K. H.


    Standard Reference Data (SRD) is the scientific, technical data whose reliability and accuracy are evaluated by scientist group. Since SRD has a great impact on the improvement of national competitiveness by stirring up technological innovation in every sector of industries, many countries are making great efforts on establishing SRD in various areas. Data center for nuclear fuel material in Korea Atomic Energy Research Institute plays a role to providing property data of nuclear fuel material at high temperature, pressure, and radiation which are essential for the safety evaluation of nuclear power. In this study, standardization of data on thermal expansion and density of UO 2 were carried out in the temperature range from 300 K to 3100 K via uncertainty evaluation of indirectly produced data. Besides, standardization of data on thermal expansion and density of simulated fuel were also done in the temperature range from 350 K to 1750 K via uncertainty evaluation of directly produced data

  14. Behaviour of short-lived iodines in operating UO2 fuel elements

    Lipsett, J.J.; Hastings, I.J.; Hunt, C.E.L.


    Sweep gas experiments have been done to determine the behaviour of short-lived fission products within operating UO 2 fuel elements at linear powers of 45, 54, and 60 KW/m, and to burnups of 70, 80, and 50 MWh/kgU respectively. Although radioiodine transport was not observed directly during normal operation, equilibrium gap inventories for I-131 were deduced from the shutdown decay behaviour of the fission gases. These inventories were a strong function of fuel power and ranged from 10 GBq (0.27 Ci) to 100 GBq (2.7 Ci) over the range tested. We conclude that the iodine inventory was adsorbed onto the fuel and/or sheath surfaces with a volatile fraction of less than 10 -2 and a charcoal-filter-penetrating fraction of less than 2x10 -4

  15. Burn-up credit applications for UO2 and MOX fuel assemblies in AREVA/COGEMA

    Toubon, H.; Riffard, C.; Batifol, M.; Pelletier, S.


    For the last seven years, AREVA/COGEMA has been implementing the second phase of its burn-up credit program (the incorporation of fission products). Since the early nineties, major actinides have been taken into account in criticality analyses first for reprocessing applications, then for transport and storage of fuel assemblies Next year (2004) COGEMA will take into account the six main fission products (Rh103, Cs133, Nd143, Sm149, Sm152 and Gd155) that make up 50% of the anti-reactivity of all fission products. The experimental program will soon be finished. The new burn-up credit methodology is in progress. After a brief overview of BUC R and D program and COGEMA's application of the BUC, this paper will focus on the new burn-up measurement for UO2 and MOX fuel assemblies. It details the measurement instrumentation and the measurement experiments on MOX fuels performed at La Hague in January 2003. (author)

  16. On possible mechanisms of rim-layer formation in the high-burnup UO2 fuel

    Zborovskii, V.; Likhanskii, V.


    Two models determining threshold conditions for onset of UO 2 fuel restructuring are developed. In the first model the conditions for fuel restructuring are related with development of the Kinoshita instability. The second model is based upon attainment of critical values by radius of over pressurised bubbles. Possibility of large bubbles formation on dislocation lines is considered with account of Xe atoms drift in the field of mechanical strain of dislocation and irradiation-induced Xe drift in vacancy concentration gradient. Computer simulations of behaviour of point defects and Xe atoms near dislocation core are carried out, results are compared with experimental data. The computer program is developed which consistently calculates point defects and Xe atoms distributions inside fuel grain with account of their behaviour near dislocation core

  17. Fabrication of ceramic grade UO2 by direct conversion of uranyl nitrate hexahydrate

    Lainetti, P.E.O.; Riella, H.G.


    A method of direct conversion of uranyl nitrate hexahydrate (UNH) solution to ceramic grade uranium dioxide powders by thermal denitration in a furnace that combines atomization nozzle and a gas stirred bed is described. The main purpose of this work is to show that this alternative process is technically viable, specially if the recovery of the scrap generated in the nuclear fuel pellet production is required, without further generation of new liquid wastes. (author)

  18. Thermodynamic and kinetic aspects of UO 2 fuel oxidation in air at 400-2000 K

    Taylor, Peter


    Most nuclear fuel oxidation research has addressed either low-temperature (1500 K) steam oxidation linked to reactor safety. This paper attempts to unify modelling for air oxidation of UO 2 fuel over a wide range of temperature, and thus to assist future improvement of the ASTEC code, co-developed by IRSN and GRS. Phenomenological correlations for different temperature ranges distinguish between oxidation on the scale of individual grains to U 3O 7 and U 3O 8 below ˜700 K and individual fragments to U 3O 8 via UO 2+ x and/or U 4O 9 above ˜1200 K. Between about 700 and 1200 K, empirical oxidation rates slowly decline as the U 3O 8 product becomes coarser-grained and more coherent, and fragment-scale processes become important. A more mechanistic approach to high-temperature oxidation addresses questions of oxygen supply, surface reaction kinetics, thermodynamic properties, and solid-state oxygen diffusion. Experimental data are scarce, however, especially at low oxygen partial pressures and high temperatures.

  19. Irradiation of UO2

    Stevanovic, M.


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

  20. Irradiation experiments of recycled PuO2-UO2 fuels by SAXTON reactor, (1)

    Yumoto, Ryozo; Akutsu, Hideo


    Seventy two mixed oxide fuel rods made by PNC were irradiated in Saxton Core 3. This paper generally describes the fuel specifications, the power history of the fuel and the post-irradiation examination of the PNC fuel. The specifications of the 4.0 w/o and 5.0 w/o enriched PuO 2 fuel rods with zircaloy-4 cladding are presented in a table and a figure. The positions of PNC fuel rods in the Saxton reactor are shown in a figure. Sixty eight 5.0 w/o PuO 2 -UO 2 fuel rods were assembled in a 9 x 9 rod array together with zircaloy-4 bars, a flux thimble, and a Sb-Be source. The power history of the Saxton Core 3 and the irradiation history of the PNC fuel rods are summarized in tables. The peak power and burnup of each fuel rod and the axial power profile are also presented. The maximum linear power rate and burnup attained were 512W/cm and 8700 MWD/T, respectively. As for the post irradiation examination, the items of nondestructive test, destructive test, and cladding test are presented together with the working flow diagram of the examination. It is concluded that the performance of all fuel rods was safe and satisfactory throughout the power history. (Aoki, K.)

  1. Overall models and experimental database for UO2 and MOX fuel increasing performance

    Bernard, L.C.; Blanpain, P.


    COPERNIC is an advanced fuel rod performance code developed by Framatome. It is based on the TRANSURANUS code that contains a clear and flexible architecture, and offers many modeling possibilities. The main objectives of COPERNIC are to accurately predict steady-state and transient fuel operations at high burnups and to incorporate advanced materials such as the Framatome M5-alloy cladding. An extensive development program was undertaken to benchmark the code to very high burnups and to new M5-alloy cladding data. New models were developed for the M5-alloy cladding and the COPERNIC thermal models were upgraded and improved to extend the predictions to burnups over 100 GWd/tM. Since key phenomena, like fission gas release, are strongly temperature dependent, many other models were upgraded also. The COPERNIC qualification range extends to 67, 55, 53 GWd/tM respectively for UO 2 , UO 2 -Gd 2 O 3 , and MOX fuels with Zircaloy-4 claddings. The range extends to 63 GWd/tM with UO 2 fuel and the advanced M5-alloy cladding. The paper focuses on thermal and fission gas release models, and on MOX fuel modeling. The COPERNIC thermal model consists of several submodels: gap conductance, gap closure, fuel thermal conductivity, radial power profile, and fuel rim. The fuel thermal conductivity and the gap closure models, in particular, have been significantly improved. The model was benchmarked with 3400 fuel centerline temperature data from many French and international programs. There are no measured to predicted statistical biases with respect to linear heat generation rate or burnup. The overall quality of the model is state-of-the-art as the model uncertainty is below 10 %. The fission gas release takes into account athermal and thermally activated mechanisms. The model was adapted to MOX and Gadolinia fuels. For the heterogeneous MOX MIMAS fuels, an effective burnup is used for the incubation threshold. For gadolinia fuels, a scaled temperature effect is used. The

  2. Steady state behaviour of gaseous fission products in UO2 nuclear fuel at low temperature

    Rao, C.B.; Raj, Baldev


    Theoretical modelling studies have been performed on steady state fission gas behaviour in UO 2 fuels at temperatures in the range 1073deg K to 1473deg K. The concentrations of gas atoms in the matrix and in the bubbles are determined. Fraction of total generated gas atoms migrating to and forming bubbles at grain boundaries is calculated. Contributions of intragranular and intergranular bubbles to the swelling are also computed. The various assumptions made to simplify computer calculations and their validity are discussed at length. Effects of changes in the fission rate, the resolution parameter, bubble concentration, gas atom diffusivity and grain radius on swelling and gas release are studied. The results of this model are compared to other theoretical models and experimental results available in literature. Possibility of extending the present model to advanced carbide and nitride fuels is discussed. (auth.)

  3. Finite element simulation of fission gas release and swelling in UO2 fuel pellets

    Denis, Alicia C.


    A fission gas release model is presented, which solves the atomic diffusion problem with xenon and krypton elements tramps produced by uranium fission during UO 2 nuclear fuel irradiation. The model considers intra and intergranular precipitation bubbles, its re dissolution owing to highly energetic fission products impact, interconnection of intergranular bubbles and gas sweeping by grain border in movement because of grain growth. In the model, the existence of a thermal gradient in the fuel pellet is considered, as well as temporal variations of fission rate owing to changes in the operation lineal power. The diffusion equation is solved by the finite element method and results of gas release and swelling calculation owing to gas fission are compared with experimental data. (author)

  4. The creep of UO2 fuel doped with Nb2O5

    Sawbridge, P.T.; Reynolds, G.L.; Burton, B.


    The creep of UO 2 containing small additions of Nb 2 O 5 has been investigated in the stress range 0.5-90 MN/m 2 at temperatures between 1422 and 1573 K. The functional dependence of the creep rate of five dopant concentrations up to 0.8 mol% Nb 2 O 5 has been examined and it was established that in all the materials the secondary creep rate could be represented by the equation epsilonkT = Asigmasup(n) exp(-Q/RT), where epsilon is the steady state creep rate per hour, Q the activation energy and A and n are constants for each material. It was observed that Nb 2 O 5 additions can cause a dramatic increase in the steady state creep rate as long as the niobium ion is maintainde in the Nb 5+ valence state. Material containing 0.4 mol% Nb 2 O 5 creeps three orders of magnitude faster than the pure material. Analysis of the results in terms of grain size compensated viscosity suggest that, like pure UO 2 , the creep rate of Nb 2 O 5 doped fuel is diffusion-controlled and proportional to the reciprocal square of the grain size. A model is developed which suggests that the increase in creep rate results from suppression of the U 5+ ion concentration by the addition of Nb 5+ ions, which modifies the crystal defect structure and hence the uranium ion diffusion coefficient. (orig.)

  5. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    Ga¨rtner, M.; Fischer, G.


    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  6. Influence of environment on the alteration of the UO2 matrix of spent fuel in storage condition

    Gaulard, C.


    Within the framework of the geological disposal of spent nuclear fuel, research on the long term behavior of spent fuel is undertaken and in particular the study of mechanisms of UO 2 oxidation and dissolution in water-saturated host rock. Under the law program on the sustainable management of radioactive materials and waste of June 28, 2006, France was chose as the reference solution the retreatment of spent fuel and disposal in deep geological repository of vitrified final waste. Nevertheless, studies on a direct disposal of spent fuel will continue for safety. The disposal concept provides for conditioning spent fuel in a steel container whose seal is guaranteed for a period specified in the order of 10,000 years. It is also reasonable to assume that the groundwater comes into contact with the fuel after the deterioration of container and lead to the UO 2 matrix degradation and the release of radionuclides. The oxidation/dissolution of UO 2 has been studied by means electrochemical methods coupled to XPS and ICP-MS measurements.A thermodynamic and bibliographic study of U(VI)/UO 2 (s) system allowed to show the effect of the physical and chemical conditions of the solution on the system, and to show the different mechanisms proposed to describe the oxidation and the dissolution of the uranium dioxide in different media (non-complexing, carbonate and clay). The study of the oxidation/dissolution of UO 2 in acidic and non-complexing media (0.1 mol/L NaCF 3 SO 3 , pH = 3), where UO 2 2+ /UO 2 (s) predominates and the formation of precipitates is limited or even avoided, showed a mechanism with two electrochemical steps and a model characteristic of UO 2 oxidation in acidic non-complexing media. Then, the study in neutral non-complexing media (0.05 mol/L NaCl, pH = 7.5) showed a mechanism with two electrochemical steps and one chemical step (EEC) in which both electrochemical steps are similar to those proposed in acidic media. Finally, a first approach of the UO 2

  7. Modeling fission gas release in high burnup ThO2-UO2 fuel

    Long, Y.; Yuan, Y.; Pilat, E.E.; Rim, C.S.; Kazimi, M.S.


    A preliminary fission gas release model to predict the performance of thoria fuel using the FRAPCON-3 computer code package has been formulated. The following modeling changes have been made in the code: - Radial power/burnup distribution; - Thermal conductivity and thermal expansion; - Rim porosity and fuel density; - Diffusion coefficient of fission gas in ThO 2 -UO 2 fuel and low temperature fission gas release model. Due to its lower epithermal resonance absorption, thoria fuel experiences a much flatter distribution of radial fissile products and radial power distribution during operation as compared to uranian fuel. The rim effect and its consequences in thoria fuel, therefore, are expected to occur only at relatively high burnup levels. The enhanced conductivity is evident for ThO 2 , but for a mixture the thermal conductivity enhancement is small. The lower thermal fuel expansion tends to negate these small advantages. With the modifications above, the new version of FRAPCON-3 matched the measured fission gas release data reasonably well using the ANS 5.4 fission gas release model. (authors)

  8. Quality assurance and control in the manufacture of metalclad UO2 reactor fuels


    The International Atomic Energy Agency has carried out a programme since its earliest days that includes the collection and dissemination of information on nuclear fuels. Since the 1960 symposium on Fuel Element Fabrication with Special Emphasis on Cladding Materials there has been an average of one meeting a year reviewing some aspect of fuel fabrication technology. A recent meeting dealing with the fabrication of UO 2 fuels was the Study Group on the Facilities and Technology needed for Nuclear Fuel Manufacture, held in Grenoble in 1972 (Rep. IAEA-158). After that meeting it became apparent that the quality of fuel production was an important aspect that had received inadequate coverage so far, and the Panel on Quality Assurance and Control in Nuclear Fuel Manufacture was convened by the Agency in Vienna in November 1974. In the working papers and discussions at the Panel meeting the viewpoints of different countries and of various interested parties, such as manufacturers, reactor operators and government authorities, were presented

  9. A Study of the Temperature Distribution in UO2 Reactor Fuel Elements

    Devold, I.


    Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO 2 fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP

  10. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    Zhang Zhuohua; Yu Junchong; Peng Shinian


    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  11. PIE and separate effect test of high burnup UO2 fuel

    Yang, Yong Sik; Kim, S.K.; Kim, D.H.


    To investigate the performance of a high burnup UO 2 fuel, the highest burnup fuel assembly in KOREA was transported to the PIE facility in KAERI. It was a 17·17 fuel assembly irradiated at the Ulchin Unit 2 PWR. The peak fuel rod average burnup was about 57MWd/kgU and locally 65MWd/kgU. The general PIE was performed to investigate the fuel rod irradiation performance. Fission gas release, burnup, oxide thickness, hydrogen pickup, CRUD, and density change were measured by destructive of non-destructive test. Microstructure change, bubble and pore size distributions were observed by optical microscopy, SEM and EPMA. All generated and available PIE results were used to verify high burnup fuel performance code INFRA. Several rods were cut for additional separate effect test. For the high burnup fission gas release behaviour analysis, annealing apparatus were developed and installed in hot cell and preliminary test was performed. In addition to current apparatus new induction furnace will be installed in hot cell to investigate the high temperature and transient fission gas release behaviour. Ring tensile test was performed to analyze the material property degradation which caused by the oxidation and hydride, and additional mechanical tests will be performed. (Author)

  12. The compaction and sintering of UO_2-Zr cermet pellets

    Tri Yulianto; Meniek Rachmawati; Etty Mutiara


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

  13. Effect of technological parameters and microstructure on mechanical strength of UO2 fuel pellets

    Radford, K.


    The effect of various peculiarities of tablet microstructure namely, sammury porosity (tablet density), grain size and pore distribution over sizes on technological parameters, is studied. It is shown that density decrease leads to a fast reduction of UO 2 tablet strength. The maximum effect on strength is produced by pore distribution over sizes, characterized by a median size, and not by the grain size, though a combined effect of those two factors is also observed. The important role of the technology of tablet production manifests itself in the fact that all operations bringing about the increase of pore or grain sizes leads to a reduction of strength. Such factors as powder origin, granule sizes, U 3 O 8 content and the amount of additions do not cause any considerable changes in the strength of tablets. Bend tests under conditions of biaxial loading should be considered as an ideal method of determining fuel tablets strength [ru

  14. Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

    Loesoenen, Pekka


    A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

  15. TEM characterization of UO2-Gd2O3 nuclear fuels synthesized by coprecipitation method

    Soldati, A.; Gana Watkins, I.; Menghini, J.; Prado, M.


    We present a micro and nano structural characterization of 4% weight doped Gd 2 O 3 -UO 2 pellet using Transmission Electron Microscopy (TEM). Agglomerate morphology and crystallite sizes were determined using light/dark field and high resolution (HR-TEM) images. Convergent beam Energy Dispersive Spectroscopy (EDS) and Electron Diffraction (ED) were used to evaluate sample composition and homogeneity, even at the nanometer scale. We obtained an average crystallite size of 90±20 nm. Moreover, from TEM-EDS analyses we determined the presence of Gadolinium in all the analyzed crystallites but with 25% variation among their concentrations. These results show the capability of TEM analysis to characterize a nuclear fuel pellet with burnable poisons nano structure and homogeneity.(author)

  16. Development of UO2-30 WT per cent PuO2 fuel for FBTR

    Majumdar, S.; Kumar, Arun; Kamath, H.S.; Ramachandran, R.; Purushotham, D.S.C.; Roy, P.R.


    The specifications on Fast Breeder Reactor (FBTR) fuel pellets have two apparently contradictory requirements viz. (1) formation of homogeneous solid between UO 2 and PuO 2 which can only be achieved by high temperature sintering and (2) density of sintered pellets in the range of 92 ± 1 per cent T.D. which is normally achieved by low temperature sintering. Deactivation of starting powders under CO 2 or addition of volatile pore formers to the powders are the two methods which have been developed for lowering the denity of the pellets without reducing the sintering temperature. Two alternative fabrication routes utilizing these processes for manufacturing of FBTR pellets are described in this report. (author)

  17. Oxidative corrosion of spent UO2 fuel in vapor and dripping groundwater at 900C

    Finch, R. J.


    Corrosion of spent UO 2 fuel has been studied in experiments conducted for nearly six years. Oxidative dissolution in vapor and dripping groundwater at 90 C occurs via general corrosion at fuel-fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 microm), possibly throughout millimeter-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has created dissolution pits that are 50 to 200 nm in diameter. Dissolution pits penetrate 1-2 microm into each grain, producing a ''worm-like'' texture along fuel-grain-boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fine-grained layer of corrosion products adjacent to the fuel (5-15 microm thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 microm to greater than 40 microm). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments

  18. Dissolution rates of unirradiated UO2, UO2 doped with 233U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    Ollila, Kaija; Albinsson, Yngve; Oversby, Virginia; Cowper, Mark


    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO 2 in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO 2 in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits ( 235 U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO 2 pellet materials containing 233 U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the 233 U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the 233 U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing

  19. Application of boron and gadolinium burnable poison particles in UO2 and PUO2 fuels in HTRs

    Kloosterman, J.L.


    Burnup calculations have been performed on a standard HTR fuel pebble (fuel zone with radius of 2.5 cm surrounded with a 0.5 cm thick graphite layer) and burnable poison particles (BPPs) containing B 4 C made of pure 10 B or containing Gd 2 O 3 made of natural Gd. Two types of fuel were considered: UO 2 fuel made of 8% enriched uranium and PuO 2 fuel made of plutonium from LWR spent fuel. The radius of the BPP and the number of particles per fuel pebble were varied to find the flattest reactivity-to-time curve. For the UO 2 fuel, the reactivity swing is lowest (around 2%) for BPPs made of B 4 C with radius of 75 μm. In this case around 1070 BPPs per fuel pebble are needed. For the PuO 2 fuel to get a reactivity swing below 4%, the optimal radius of the BPP is the same, but the number of particles per fuel pebble should be around 1600. The optimal radius of the Gd 2 O 3 particles in the UO 2 fuel is about 10 times that of the B 4 C particles. The reactivity swing is around 3% when each fuel pebble contains only 9 BPPs with radius of 840 μm. The results of the Gd particles illustrate nicely the usage of black burnable poison particles introduced by Van Dam [Ann. Nuclear Energy 27 (2000) 733

  20. Analytical determination of thermal conductivity of W-UO2 and W-UN CERMET nuclear fuels

    Webb, Jonathan A.; Charit, Indrajit


    The thermal conductivity of tungsten based CERMET fuels containing UO2 and UN fuel particles are determined as a function of particle geometry, stabilizer fraction and fuel-volume fraction, by using a combination of an analytical approach and experimental data collected from literature. Thermal conductivity is estimated using the Bruggeman-Fricke model. This study demonstrates that thermal conductivities of various CERMET fuels can be analytically predicted to values that are very close to the experimentally determined ones.

  1. Effect of PCMI restraint on bubble size distribution in the rim structure of UO2 fuel

    Oh, Je-Yong; Koo, Yang-Hyun; Cheon, Jin-Sik; Lee, Byung-Ho; Sohn, Dong-Seong


    Generally, the bubble size in the rim structure of UO 2 is not dependent on the fuel burnup and the bubble pressure is higher than that in the equilibrium condition. However it was also observed that if the fuel pellet is not restrained, the size of the bubbles in the rim structure could be larger than that in the restraint condition. Although the wide variety of rim bubble sizes and porosities possibly result from an external restrain effect, the quantitative method to analyze the effect of PCMI restraint on bubble distribution in the rim is not available at the moment. In this paper, a method is developed which can be used to analyze the effect of PCMI restraint on the bubble distribution in the rim structure of UO 2 fuel based on the data in the literatures. The total number of Xe atoms in the rim bubbles per unit rim volume could be derived by a summation of the number of Xe atoms of each rim bubble in a unit rim volume. The number of Xe atoms of each rim bubble could be calculated by the Van der Waals equation of state and the pressure expressed by p=σ+C/r, where C is an unknown constant to be determined as a function of the temperature and the burnup. On the other hand, the total number of Xe atoms in the rim bubbles per unit rim volume can also be calculated by Xe depression data. If the fuel pellet is not restrained, the uniform hydrostatic stress, σ is zero. Hence if the data of the fuel disk without a restraint is used, a constant C can be obtained at 823K and a local burnup of 90 GWd/t. Although the local burnup of PCMI restraint case is slightly different from that without PCMI restraint, the value derived above is used for the analysis of PCMI restraint case. The calculated bubble distribution with PCMI restraint was similar to the measured one. Because the effect of PCMI restraint on bubble size increased with the bubble size, the development of a large bubble was suppressed. Hence, the PCMI restraint caused a typical bubble size in the rim and

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

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


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

  3. Correlation between UO2 powder and pellet quality in PHWR fuel manufacturing

    Glodeanu, F.; Spinzi, M.; Balan, V.


    Natural uranium dioxide fuel for heavy water reactors has a series of very tightly controlled quality factors: Chemical purity, density and microstructures. Although the fabrication history may consistently affect the fuel quality, the quality factor mentioned above are function mainly of the quality of the powder used as raw material. As regards the fulfilment of the requirements for very high density of the pellets, it was found that in a definite technology the raw material plays the decisive part. Except for the powder sinterability, one found other important subtile parameters, such as the degree of agglomeration and structural homogeneity. The fuel microstructure, very important for in-serive performances of the fuel, is related to a great extent to some powder characteristics (homogeneity, sinterability). This is why much stress was laid on UO 2 power quality evaluation both by standard methods and non-conventional ones (agglomeration, microscopy, X-rays). Some of the characteristics defined by product specification, such as powder sinterability, should be better defined to guarantee the final product quality. (orig.)

  4. Results of the irradiation of mixed UO2 - PuO2 oxide fuel elements

    Mikailoff, H.; Mustelier, J.P.; Bloch, J.; Ezran, L.; Hayet, L.


    In order to study the behaviour of fuel elements used for the first charge of the reactor Rapsodie, a first batch of eleven needles was irradiated in the reactor EL3 and then examined. These needles (having a shape very similar lo that of the actual needles to be used) were made up of a stack of sintered mixed-oxide pellets: UO 2 containing about 10 per cent of PuO 2 . The density was 85 to 97 per cent of the theoretical, value. The diametral gap between the oxide and the stainless steel can was between 0,06 and 0,27 mm. The specific powers varied from 1230 to 2700 W/cm 3 and the can temperature was between 450 and 630 C. The maximum burn-up attained was 22000 MW days/tonne. Examination of the needles (metrology, radiography and γ-spectrography) revealed certain macroscopic changes, and the evolution of the fuel was shown by micrographic studies. These observations were used, together with flux measurements results, to calculate the temperature distribution inside the fuel. The volume of the fission gas produced was measured in some of the samples; the results are interpreted taking into account the temperature distribution in the oxide and the burn-up attained. Finally a study was made both of the behaviour of a fuel element whose central part was molten during irradiation, and of the effect of sodium which had penetrated into some of the samples following can rupture. (author) [fr

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

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


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

  6. Estimate of the instant release fraction for UO2 and MOX fuel at t=0

    Johnson, L.; Poinssot, C; Ferry, C.; Lovera, P.


    values, which results in significant overprediction of average IRF values. Best estimate IRF values are determined for moderate burnup UO 2 fuel for nuclides for which data exist, because the understanding and data is sufficient. Only pessimistic IRF values are estimated for radionuclides for which little data is available and in the case of MOX fuel and higher burnup UO 2 fuel. Special attention is given to several phenomena occurring in the outer region of fuel pellets (rim region) resulting in restructuring of fuel grains. These include: a) high fission density as a result of high yields of 239 Pu arising from capture of epithermal neutrons; b) increased porosity; c) reduction in grain size; d) increased thermal release of fission gas from the grains. From the perspective of assessing the release of fission products from spent fuel under disposal conditions, the restructuring process is important

  7. The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

    White, R.J.


    Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

  8. High 240Pu FTR/EMC experiments and analysis: Carbide fuel and UO2 blanket subassembly worths

    Ombrellaro, P.A.


    Carbide-plutonium fuel and UO 2 blanket subassembly worth measurements performed at ANL in the EMC/LWR were analyzed. Composition exchange worth calculations were performed for: (a) the replacement of high- 240 Pu fuel composition for low- 240 Pu fuel composition and carbide-plutonium fuel composition, successively, in the center subassembly of the core; (b) the replacement of low- 240 Pu fuel composition for carbide--plutonium fuel composition in one outer driver subassembly; and (c) the replacement of the radial reflector composition with UO 2 blanket composition in one subassembly of the radial reflector. The composition exchange worth calculations were performed in two-dimensional x,y geometry, using diffusion theory and perturbation theory. Each method produces about the same calculated-to-experimental bias factors

  9. Quality control and testing UO2 powder and sintering pellets for nuclear fuel for LWR in out of pile condition

    Djuricic, Lj.; Katanic, J.; Stefanovic, M.


    The analysis of chemical and physical characteristics of fuels based on UO2 from the point of view of requested properties in the nuclear application, of the foreign technical methods of characterisation and domestic experience is given as one of the first steps toward standardization in the field in the state

  10. Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

    Sarmiento, G.S.


    Within natural UO 2 fuel elements enriched with plutonium, this last material should form PuO 2 solid solutions inside the UO 2 pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO 2 islands are formed in the UO 2 matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium. A detailed study of the thermal effects produced by PuO 2 particles in the UO 2 matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spatial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a well-known exact solution, has been obtained. (orig.)

  11. Analysis of neutron parameters in light water moderated lattices of ThO2 and UO2 fuel rods

    Onusic Junior, J.; Oosterkamp, W.J.


    A large number of light water moderated lattices of UO 2 and ThO 2 fuel rods were analyzed with the code HAMMER. The purpose of the study was to compare experimental results with computer calculated values. The model employed is described and some modification were introduced in the resonance parameters of Th-232 to increase the agreement with the experimental value [pt

  12. Behaviour of short-lived fission products within operating UO2 fuel elements

    Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.


    We have carried out experiments using a ''sweep gas'' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 500 mm long and contained fuel of density 10.65-10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. In tests at linear powers of 45 and 60 kW/m to maximum burnups of 70 MW.h/kg U, the species measured directly at the spectrometer were generally the short-lived xenons and kryptons. We did not observe iodine or bromine during normal operation. However, we have deduced the behaviour of I-133 and I-135 from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against lambda (decay constant) or effective lambda for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. Our inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5x10 -3 . The ANS 5.4 release correlation gives calculated results in good agreement with our measurements. (author)

  13. Transport and leaching of technetium and uranium from spent UO2 fuel in compacted bentonite clay

    Ramebaeck, H.; Albinsson, Y.; Skaalberg, M.; Eklund, U.B.; Kjellberg, L.; Werme, L.


    The transport properties of Tc and U in compacted bentonite clay and the leaching behaviour of these elements from spent nuclear fuel in the same system were investigated. Pieces of spent UO 2 fuel were embedded in bentonite clay (ρ d =2100 kg/m 3 ). A low saline synthetic groundwater was used as the aqueous phase. After certain experimental times, the bentonite clay was cut into 0.1 mm thick slices, which were analysed for their content of Tc and U. Measurements were made using inductively coupled plasma mass spectrometry. Tc analysis comprised chemical separation. The analysis of U was done by means of detecting 236 U, since the natural content of U in bentonite clay made it impossible to distinguish between U originating from the fuel and the clay. The influence of different additives mixed into the clay was studied. The results showed an influence on both transport and leaching behaviour when metallic Fe was mixed into the clay. This indicates that Tc and U are reduced to their lower oxidation states as a result of this additive

  14. Molecular dynamics simulation of Xe bubble nucleation in nanocrystalline UO2 nuclear fuel

    Moore, Emily; René Corrales, L.; Desai, Tapan; Devanathan, Ram


    Highlights: ► We simulated the interactions of defects and fission gas with grain boundaries in nuclear fuel. ► We observed the formation of Xe bubble nuclei that are difficult to observe experimentally. ► The bubble nuclei form by vacancy-assisted diffusion of Xe atoms. ► We also observed the initial stages of grain boundary motion. ► The study offers insights to the design of nuclear fuel to control fission gas release. - Abstract: We have performed molecular dynamics (MD) simulations to investigate the dynamical interactions between vacancy defects, fission gas atoms (Xe), and grain boundaries in a model of polycrystalline UO 2 nuclear fuel with average grain diameter of about 20 nm. We followed the mobility and aggregation of Xe atoms in the vacancy-saturated model compound for up to 2 ns. During this time we observed the aggregation of Xe atoms into nuclei, which are possible precursors to Xe bubbles. The nucleation was driven by the migration of Xe atoms via vacancy-assisted diffusion. The Xe clusters aggregate faster than grain boundary diffusion rates and are smaller than experimentally observed bubbles. As the system evolves towards equilibrium, the Xe atom cluster growth slows down significantly, and the lattice relaxes around the cluster. These simulations provide insights into fundamental physical processes that are inaccessible to experiment.

  15. Contribution to the identification and the evaluation of a doped UO2 fuel with controlled oxygen potential

    Pennisi, Vanessa


    Temperature and oxygen partial pressure (PO 2 ) of nuclear oxide fuels are the main parameters governing both their thermochemical evolution in reactor and the speciation of volatile fission products such as Cs, I or Te. An innovative way to limit the risk of cladding rupture by corrosion under irradiation consists in buffering the oxygen partial pressure of the fuel under operation in a PO 2 domain where the fission gas are harmless towards Zr clad, by using solid redox buffers as additives. Niobium, with its NbO 2 /NbO and Nb 2 O 5 /NbO 2 redox couples has been found to be a promising candidate to this end. A manufacturing process of a buffered UO 2 fuel, doped with niobium has been optimized, in order to fulfill usual specifications (density, microstructure). The experimental study of the UO 2 -NbO x system has shown the existence of a liquid phase between UO 2 and NbO x at 810 C, which was not reported in the literature. The characterization of Nb containing phases present in UO 2 both in solid solution and as precipitates has lead us to propose a solubility thermodynamic model of niobium in UO 2 at 1700 C. An extensive study of the niobium precipitates shows the co-existence in the fuel of NbO 2 and NbO as major phases, together with small amounts of metallic Nb. The coexistence of niobium under two oxidation states inside the fuel is a key element of demonstration of a possible in-situ buffering effect, which is likely to impact some properties of the material that are dependent upon PO 2 , such as densification. These results confirm the promising potential of oxygen buffered fuels as regard to their performance in reactor. (author) [fr

  16. Study and optimization of the composite nuclear fuel with burnable poison UO2/Gd2O3

    Balestrieri, D.


    The studied composite ceramics is a nuclear fuel constituted of a uranium dioxide matrix UO 2 in which big grains (or 'macro-masses') of gadolinium oxide (Gd 2 O 3 ) of 300 ± 100 μm of diameter (mass fraction of 12%) are dispersed. Used as burnable poison (neutron absorbent whose action disappears progressively during the irradiation), gadolinium oxide is the object of a particular attention because some of its properties as the crystal structure, the aptitude to sintering and the thermomechanical behavior have been studied. The aim of this work is to perfect and optimize the process of manufacture of the composite in order to answer to accurate specifications for the density, the shape and the mass fraction of macro-masses. In this framework, it has been necessary to strengthen the Gd 2 O 3 macro-masses by a thermal treatment in order to avoid their deformation during the uniaxial pressing. The influence of this pre-consolidation on the ended microstructure, the aptitude to sintering and the thermal conductivity of the composite have been studied. (O.M.)

  17. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.


    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  18. Possible effects of UO2 oxidation on light water reactor spent fuel performance in long-term geologic disposal

    Almassy, M.Y.; Woodley, R.E.


    Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO 2 ) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO 2 oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented

  19. On the behaviour of intragranular fission gas in UO2 fuel

    Loesoenen, Pekka


    Data obtained from the literature concerning the behaviour of intragranular gas in sintered LWR UO 2 fuel are reviewed comprehensively. The characteristics of single gas atoms and bubbles, as a function of irradiation time, temperature, fission rate and burn-up are described, based on the reported experimental data. The relevance of various phenomena affecting gas behaviour is evaluated. The current status of modelling of the behaviour of intragranular gas is considered in light of the present findings. Simple calculations showed that the conventional approximation for the effective diffusion coefficient does not adequately describe the gas behaviour under transient conditions, when bubble coarsening plays a key role in the release. The difference in the release fraction, compared with a more mechanistic approach, could be as large as 30%. A number of recommendations regarding possible defects in the mechanistic approach to modelling of intragranular gas are highlighted. The lack of an effective numerical method for solving the set of relevant non-linear differential equations is shown to be a serious obstacle in implementing the mechanistic models for fission gas release (FGR), in integral fuel performance codes

  20. The Width of High Burnup Structure in LWR UO2 Fuel

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong


    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

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

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


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

  2. Methods for assessing homogeneity in ThO2--UO2 fuels (LWBR Development Program)

    Berman, R.M.


    ThO 2 -UO 2 solid solutions fabricated as LWBR fuel pellets are examined for uniform uranium distribution by means of autoradiography. Kodak NTA plates are used. Images of inhomogeneities are 29 +- 10 microns larger in diameter than the high-urania segregations that caused them, due to the range of alpha particles in the emulsion, and an appropriate correction must be made. Photographic density is approximately linear with urania content in the region between underexposure and overexposure, but the slope of the calibration curve varies with aging and growth of alpha activity from the parasitic 232 U and its decomposition products. A calibration must therefore be performed using two known points--the average photographic density (corresponding to the average composition) and an extrapolated background (corresponding to zero urania). As part of production pellet inspection, plates are evaluated by inspectors, who count segregations by size classes. This is supplemented by microdensitometer scans of the autoradiograph and by electron probe studies of the original sample if apparent homogeneity is marginal

  3. The Effect of the UO2/ZrO2 Composition on Fuel/Coolant Interaction

    Song, Jin Ho; Kim, Jong Hwan


    A series of experiments on fuel/coolant interaction (FCI) was performed in the TROI facility, where the composition of the mixture was varied. The compositions of the UO 2 and ZrO 2 mixture in weight percent were 50:50, 70:30, 80:20, and pure ZrO 2 . The responses of the system including the temperature of the pool of water in the test vessel, pressure and temperature of the containment vessel, and dynamic pressures and force were measured. In addition, high-speed movies were taken through the windows. The tests using corium with a 70:30 composition and pure zirconia resulted in a spontaneous energetic steam explosion, while the tests with other compositions did not lead to an energetic FCI. The debris size distribution and pressure and temperature responses clearly indicated the cases with an energetic explosion and the cases without an explosion. The high-speed movie taken during the FCI through the visible window clearly disclosed the outstanding phases of the FCI, which were the melt entry phase, the triggering phase, and the continued melt jet and expansion of the mixing zone phase

  4. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO2 fuel assemblies

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.


    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO 2 fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for 238 Pu, 144 Nd, 145 Nd, 146 Nd, 148 Nd, 134 Cs, 154 Eu, 152 Sm, 154 Gd, and 157 Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  5. Experimental simulation of irradiation effects on thermomechanical behaviour of UO2 fuel: Impact of solid and gaseous fission products

    Balland, J.


    Predictive simulation of thermomechanical behaviour of nuclear fuel has to take into account irradiation effects. Fission Products (FP) can modify the thermomechanical behaviour of UO 2 . During this thesis, differentiation was made between fission products which create a solid solution with UO 2 and gaseous products, generating pressurized bubbles. SIMFUELS containing gadolinium oxide and pressurized argon bubbles were manufactured, respectively by conventional process and by Gas Pressure Sintering. Brittle and ductile behaviour of UO 2 was investigated, under experimental conditions representative of Pellet-Cladding Interaction (PCI), respectively with 3 points bending tests and compressive creep tests. Investigation of brittle behaviour of UO 2 showed that fracture is mainly controlled by natural defects, like porosities, acting like starting points for cracks propagation. Addition of simulates fission products increase the brittle-to-ductile transition temperature of UO 2 , up to 400-500 C regarding FP in solid solution, and up to 200 C for gaseous products. Fission products although reduce fracture stresses, by a factor between 1.5 and 4, respectively for gas bubbles and solid solutions. Decrease of fracture stress is linked to an increase of microstructural defects due the solid solution and to pressurized bubbles located at grain boundaries. Pellets were tested under compressive solicitation at high temperatures. Experimental results of creep tests are well represented by Norton laws. Creep controlling mechanisms are evidenced by microstructural analysis performed on pellets at different strains. On the basis of calculations made for fuels having the same microstructures than the SIMFUELs, a creep factor is determined. It revealed a strong hardening effect of the solid solution, due to the fact that the added elements anchor the dislocations, whereas pressurized bubbles showed a coupling between hardening and softening effects. (author)

  6. UO2-PuO2 fuel pin capsule-irradiations of the test series FR 2-5a

    Dienst, W.; Goetzmann, O.; Schulz, B.


    In the capsule-irradiation test series FR 2-5a, short UO 2 -PuO 2 fuel pins (80 mm fuel length) of 7 mm diameter were irradiated in a thermal neutron flux at mean rod powers of 400 - 450 W/cm and mean cladding surface temperatures of 500 - 550 0 C to burnups of 0.6, 1.8 and 5.0 at% (U + Pu). Void volume redistribution in the fuel pins was examined in micrographs of cross-sections by measuring crack widths, central void diameters, and fuel porosity. The width of the radial cracks at the outer fuel rim was taken as a basis for measuring the irradiation-induced densification of the UO 2 -PuO 2 fuel. The result was that the final fuel density after irradiation-induced densification amounted to 92 - 94% TD and had already been reached after 0.6 at% burnup. The porosity measurement on fuel cross-sections was to show a possible dependence of the radial porosity redistribution on the initial sintered density. Examining the fuel pin diameters after irradiation showed permanent cladding strains after 5 at% burnup, which must be due to mechanical interaction with the fuel. To judge if the chemical compatibility between the fuel and the cladding of Cr-Ni-stainless steel 1.4988, the depths of chemical attack on the cladding inside was measured by micrographs of fuel pin cross-sections. (orig./GSC) [de

  7. The Influences of Uranium Concentration and Polyvinyl Alcohol on the Quality UO2 Microsphere for Fuel of High Temperature Reactor

    Damunir; Sukarsono; Bangun-Wasito; Endang Nawangsih


    The influences of uranium concentration and PVA on the quality of UO 2 microspheres for fuel of high temperature reactor have been investigated. The UO 2 particles were prepared by gel precipitation using internal gelation process. Uranyl nitrate solution containing uranium of 100 g/l was neutralized using NH 4 OH 1 M. The solution was changed into sol by adding 60 g PVA/l solution while stirred and heated up to 80 o C for 20 minutes. In order to find gels in spherical shape, the sol solution was dropped into 5 M NH 4 OH medium. The formed gels were small spheres, was washed, screened and heated up to 120 o C. After that, the gels were calcined at 800 o C for 4 hours, resulting in U 3 O 8 spheres. The U 3 O 8 particles were reduced using H 2 gas in a N 2 media at 800 o C for 4 hours, yielded in UO 2 spheres. Using a similar procedure, the influence of uranium concentration of 150-250 g/l and PVA 40-80 g/l were studied. The qualities of UO 2 particles were obtained by their physical properties, i.e. density, specific surface area, total volume of pores and pore radius using surface area meter and N 2 gas used as absorbent, and the particle size was observed using optical microscope. The result showed that the changing of uranium and PVA concentrations on the internal gelation affected the density, specific surface area, total volume of pores and pore radius of UO 2 particles. (author)

  8. Effects of additives on the sintering of UO2.Gd2O3 nuclear fuel

    Pagano Junior, Luciano


    The addition of 0.5wt% TiO 2 , Nb 2 O 5 , SiO 2 , Fe 2 O 3 and Al(OH) 3 in the UO 2 ·7%Gd 2 O 3 nuclear fuel and the effect on its sintering kinetics under a 99.999% H 2 atmosphere were investigated by stepwise isothermal dilatometry. This fuel, used as burnable poison in nuclear power plants, presents a diffusion barrier around 1573 K that impairs densification. The aid of the sintering additives TiO 2 , Al(OH) 3 , Nb 2 O 5 and Fe 2 O 3 turned out to be effective to obtain the required final density, unlike the effect observed for the SiO 2 -doped composition. The activation energy for the intermediate sintering stage was calculated by stepwise isothermal dilatometry method and a positive correlation with the sintered body density was found. The method was valid for part of the intermediate sintering stage, in the range from 1200 K to 1700 K for the doped compositions and with no additive, except for the SiO 2 -doped one, whose validity range was between 1500 K and 1900 K. The energy-density correlation was not valid for the SiO 2 -doped composition, whose effect was to reduce the final density. This anomalous behavior may be attributed to the intense loss of Si mass, probably due to lower oxides volatilization, during the initial sintering stage at temperatures lower than 1173 K. Similar loss, but no so intense, was observed for the Al(OH) 3 -doped composition in the temperature interval from 1173 K to 1573 K. The Si concentration decrease to residual values of dozens of parts per million may explain its anomalous behavior. The positive correlation between activation energy and sintered body density may be explained by the inhibitor role played by the TiO 2 , Nb 2 O 5 , Fe 2 O 3 and Al(OH) 3 additives on the diffusion mechanisms that enhance the coarsening regime. As a consequence, the densification mechanisms are favored in the competition for the surface free energy. The coarsening-densification transition temperature model, originally suggested for the UO 2

  9. Transient fission gas release from UO2 fuel for high temperature and high burnup

    Szuta, M.


    In the present paper it is assumed that the fission gas release kinetics from an irradiated UO 2 fuel for high temperature is determined by the kinetics of grain growth. A well founded assumption that Vitanza curve describes the change of uranium dioxide re-crystallization temperature and the experimental results referring to the limiting grain size presented in the literature are used to modify the grain growth model. Algorithms of fission gas release due to re-crystallization of uranium dioxide grains are worked out. The defect trap model of fission gas behaviour described in the earlier papers is supplemented with the algorithms. Calculations of fission gas release in function of time, temperature, burn-up and initial grain sizes are obtained. Computation of transient fission gas release in the paper is limited to the case where steady state of irradiation to accumulate a desired burn-up is performed below the temperature of re-crystallization then the subsequent step temperature increase follows. There are considered two kinds of step temperature increase for different burn-up: the final temperature of the step increase is below and above the re-crystallization temperature. Calculations show that bursts of fission gas are predicted in both kinds. The release rate of gas liberated for the final temperature above the re-crystallization temperature is much higher than for final temperature below the re-crystallization temperature. The time required for the burst to subside is longer due to grain growth than due to diffusion of bubbles and knock-out release. The theoretical results explain qualitatively the experimental data but some of them need to be verified since this sort of experimental data are not found in the available literature. (author)

  10. Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

    Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.


    Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

  11. The dissolution of unirradiated UO2 fuel pellets under simulated disposal conditions

    Ollila, K.; Leino-Forsman, H.


    The dissolution behaviour of unirradiated UO 2 pellets was studied as a function of water composition under oxidizing and reducing conditions at 25 deg C. The waters included deionized water as the reference water, sodium bicarbonate solutions with varying bicarbonate content, and two different synthetic groundwaters. The release of uranium was measured during static batch dissolution experiments of long duration (3-4 years)

  12. Niobia-doped UO2 fuel manufacturing experience at British nuclear fuels Ltd

    Marsh, G.; Wood, G.A.; Perkins, C.P.


    BNFL Fuel Division has made niobia doped fuel for over twenty years in its Springfields Research and Development facilities. This paper reviews this experience together with feedback from successful in-reactor and laboratory tests. Recent experience in qualifying and manufacturing niobia doped fuel pellets for a European PWR will be described. (author)

  13. PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

    Anand, A.K.; Anantharaman, K.; Sarda, V.


    The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

  14. Metallographic examination of (uth) O2 and UO2 fuel tested in power ramp conditions in triga reactor

    Ioncescu, M.; Uta, O.


    The purpose of this paper is to determine the behavior of two fuel experimental elements (EC1 and EC2), by destructive post-irradiation examination. The fuel elements were mounted inside a pattern port, one in extension of the other and irradiated in power ramp conditions in order to check their behavior. Fuel element 1 (EC1) contains (UTh)O''2 pellet, and other one (EC2) UO''2 pellet. The results of destructive post-irradiation examination are evidenced by metallographic and ceramographic analyses. The data obtained from the post-irradiation examinations are used, first to confirm the security, reliability and nuclear fuel performance, and second, for the development of CANDU fuel. The results obtained by destructive examinations regarding the integrity, sheath hydrating and oxidation as well as the structural modifications are typical for fuel elements tested in power ramp conditions. (authors)

  15. Memory list for the ordering of nuclear fuel elements with UO2 fuel


    The memory list will help to simplify and speed up the technical procedure of fuel element supply for nuclear reactors. Operators of nuclear power plants take great interest in the latest state of thechnology, if sufficiently tested, being applied with regard to material, manufacturing and testing methods. In order to obtain an unlimited availability of the nuclear plant in the future, this application of technology should be taken care of when designing and producing fuel elements. When ordering fuel elements special attention should be drawn to the interdependence of reactor and fuel element with reqard to design and construction, about which, howevers, no further details are given. When ordering fuel elements the operator give the producer all design data of the reactor core and the fuel elements as well as the planned operation mode. He also hands in the respective graphs and the required conditions for design so that a correct and detailed offer can be supplied. An exemplary extent of supply is shown in the given memory list. The regulations required herefore on passing technical material to the fuel element producers have to be established by agreements made by the customer. The order to be given should be itemized as follows: requirements, quality controland quality assurance, warranties and conditions, limits and extent of supply, terms of delivery. (orig./HP) [de

  16. Simulation of the neutron-physical properties of the classical UO2 fuel and of MOX fuel during the burn-up by Transuranus

    Breza, J. jr.; Necas, V.; Daoeilek, P.


    The classical nuclear fuel UO 2 is well known for VVER reactors. Nevertheless, in the near future it will be possible to replace this fuel by novel, advanced kinds of fuel, for instance MOX, inert matrices fuel, etc., that will allow to increase the level of burn-up and minimize the amount of hazardous waste. The code Transuranus [2], designed at ITU Karlsruhe, is intended for thermal and mechanical analyses of fuel elements in nuclear reactors. We have utilized the code Transuranus to simulate the neutron-physical properties of the classical UO 2 fuel and of MOX fuel during the burn-up to a level of 40 MWd/kgHM. We compare obtained results of uranium and plutonium nuclides concentrations, their changes during burn-up, with results obtained by code HELIOS [3], which is well-validated code for this kind of applications. We performed calculations of fission gasses concentrations, namely xenon and krypton. (author)

  17. Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

    Dyment, I.G.; Noguera Rojas, Francisco


    The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

  18. Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

    Geithoff, D.


    Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

  19. Improvement of fuel-element reliability by insertion of UO2 microspheres in the gap between pellet and clad

    Mehedinteanu, S.; Glodeanu, F.; Dobos, I.


    With the accumulation of power reactor fuel operating experience, the study of the PCI phenomenon and the development of remedies have become important items in fuel research and development everywhere. The 'power-ramp' failure has drawn attention to the problem of obtaining high reliability from high burn-up fuel rods. Considerable attention has been paid to minimizing the cladding stresses imparted by fuel pellets during the power ramp. The paper describes a new concept of pellet-clad bonding by insertion of UO 2 microspheres in the gap. It is pointed out that the main advantages of this concept are: the low friction coefficient between pellet and clad; the accomodation of cracked pellet expansion by local microyielding of irradiation-embrittled clad; the reduced ridge height by use of undished pellets or other pellet shape; that the fine-sized UO 2 microspheres infiltrate around the pellets thus permitting the use of cracked or chipped pellets and also sintered pellets without the previously required grinding step needed for accurate sizing, etc. (author)

  20. An exercise to establish optimum procedures for the characterisation of porosity in UO2 fuel pellets

    Small, G.J.


    A standard metallographic preparation technique for UO 2 is proposed. The criteria for choosing the optimum route are that the specimen should be scratch-free and that the pores inherent to any sintered UO 2 pellet should be neither enlarged nor filled-in during preparation. Having met these criteria one has a specimen suitable for quantitative metallography which can be used to monitor porosity changes due to in-pile sintering. A procedure for analysing the porosity is suggested. This consists of imaging the specimen surface over a range of magnifications using both optical and Scanning Electron Microscopy in order to cover the range of pore sizes of interest (0.1 μm to 10 μm diameter). These images are then analysed to obtain figures for the distribution of pores as a function of diameter. Two methods of pore-size analysis are reviewed, the manual Zeiss Particle Size Analyser and a more sophisticated electronic instrument - the Quantimet. A comparison is made between these two instruments on the basis of accuracy, reproducibility and ease of operation. (author)

  1. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    Ipek, O.


    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  2. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.


    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  3. Development of a kinetic model for the dissolution of the UO2 spent nuclear fuel. Application of the model to the minor radionuclides

    Bruno, J.; Cera, E.; Duro, L.; Pon, J.; Pablo, J. de; Eriksen, Trygve


    A kinetic model has been developed in order to explain the evolution of the spent fuel matrix/groundwater system. Mass balance equations have been used to follow the evolution of the system with time. The model has been calibrated by using experimental dissolution data from spent fuel leaching tests from Studsvik and KTH and from synthetic unirradiated UO 2 dissolution tests from VTT. The results of the testing exercise indicate that the combination of mass balance equations together with the kinetic rate laws constitute a useful tool to model and explain experimental dissolution data available in the literature for UO 2 solid phases, including uraninites, unirradiated UO 2 and spent fuel. Although the key processes are well identified and understood, there are still some remaining uncertainties concerning some of the critical parameters of the model. This is particularly true for the density of UO 2 sites prone to oxidation and the rates and mechanisms of the hydrogen peroxide and the combined oxygen and bicarbonate promoted dissolution of UO 2 for oxidant concentration ranges relevant to the spent fuel disposal system. The mass balance kinetic model developed has been extended to minor radionuclides contained in the matrix, i.e. Pu, Tc and Sr. In the case of Pu, the model presented reproduces the behaviour of this critical radionuclide even at early contact times. As it would be expected, Tc seems to follow a different mechanism for its release with respect to the UO 2 matrix dissolution, which is probably linked to the rate of oxidation of Tc metallic inclusions in the fuel. A co- dissolution process of Sr with the UO 2 matrix reproduces the long term dissolution behaviour of this radionuclide, better than the initial Sr release rates

  4. A new technique to measure fission-product diffusion coefficients in UO2 fuel

    Hocking, W.H.; Verrall, R.A.; Bushby, S.J.


    This paper describes a new out-reactor technique for the measurement of fission-product diffusion rates in UO 2 . The technique accurately simulates in-reactor fission-fragment effects: a thermal diffusion that is due to localized mixing in the fission track, radiation-enhanced diffusion that is due to point-defect creation by fission fragments, and bubble resolution. The technique utilizes heavy-ion accelerators - low energy (40 keV to 1 MeV) for fission-product implantation, high energy (72 MeV) to create fission-fragment damage effects, and secondary ion mass spectrometry (SIMS) for measuring the depth profile of the implanted species. Preliminary results are presented from annealing tests (not in the 72 MeV ion flux) at 1465 deg. C and 1650 deg. C at low and high concentrations of fission products. (author)

  5. Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

    Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar


    Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

  6. Grain growth in UO2

    Hastings, I.J.; Scoberg, J.A.; Walden, W.


    Grain growth studies have been carried out on UO 2 to provide data for the fuel modelling program and to evaluate fuel fabricated in commissioning the Mixed Oxide Fuel Fabrication Laboratory at Chalk River Nuclear Laboratories. Fuel examined includes natural UO 2 commercially fabricated from ADU powder for CANDU reactors; natural UO 2 commercially fabricated from AU powder; natural UO 2 from ADU and AU powder, fabricated in the MOFFL; and commercially fabricated UO 2 enriched 1.7, 4.5, and 9.6 wt. percent U-235 in U. Samples were step-annealed in vacuo at 1870-2070 K for up to 32.5 h. All data fit a (grain size)sup(2.5) versus annealing time relationship. Apparent activation energy for grain growth, Q, depends on fuel type and varies from 150+-10 kJ/mol for early AU powder to 360+-10 kJ/mol for pellets from ADU fabricated in the MOFFL. Grain sizes calculated using the laboratory equation in a fuel performance code tend to be greater than those measured in irradiated natural fuel, suggesting irradiation-induced inhibition of grain growth. However, any inhibition is equivalent to that expected for a systematic 5 percent underpredicition in reactor power. (author)

  7. Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

    Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.


    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

  8. Chemical activity of noble gases Kr and Xe and its impact on fission gas accumulation in the irradiated UO2 fuel

    Szuta, M.


    It is generally accepted that most of the insoluble inert gas atoms Xe and Kr produced during fissioning are retained in the fuel irradiated at a temperature lower than the threshold. Experimental data imply that we can assume that after irradiation exposure in excess of 10 18 fissions/cm 3 the single gas atom diffusion can be disregarded in description of fission gas behaviour. It is assumed that the vicinity of the fission fragment trajectory is the place of intensive irradiation induced chemical interaction of the fission gas products with UO 2 . Significant part of fission gas product is thus expected to be chemically bound in the matrix of UO 2 . Experiments with mixture of noble gases, coupled with theoretical calculations, provide strong evidence for direct bonds between Ar, Kr, or Xe atoms and the U atom of the CUO molecule. Because of its positive charge, the UO 2 2+ ion, which is isoelectronic with CUO, should form even stronger bonds with noble gas atoms, which could lead to a growing number of complexes that contain direct noble gas - to - actinide bonds. Considering the huge amount of gas immobilised in the UO 2 fuel the solution process and in consequence the re-solution process of rare gases is to be replaced by the chemical bonding process. This explains the fission gas accumulation in the irradiated UO 2 fuel. (author)

  9. Study of UO2-10WT%Gd2O3 fuel pellets obtained by seeding method using AUC co-precipitation and mechanical mixing processes

    Lima, M.M.F.; Ferraz, W.B.A.; Santos, M.M. dos; Pinto, L.C.M.; Santos, A.


    The use of gadolinium and uranium mixed oxide as a nuclear fuel aims to obtain a fuel with a performance better than that of UO 2 fuel. In this work, seeding method was used to improve ionic diffusivity during sintering to produce high density pellets containing coarse grains by co-precipitation and mechanical mixing processes. Sintered UO 2 -10 wt% Gd 2 O 3 pellets were obtained using the reference processes with 2 wt% and 5 wt% UO 2 seeds with two granulometries, less than 20 μm and between 20 and 38 μm. Characterisation was carried out by chemical analysis, surface area, X-ray diffraction, SEM, WDS, image analysis, and densitometry. The seeding method using mechanical mixing process was more effective than the co-precipitation method. Furthermore, mechanical mixing process resulted in an increase in density of UO 2 -10wt% Gd 2 O 3 with seeds in relation to that of UO 2 -10wt% Gd 2 O 3 without seeds. (author)

  10. Microstructure and fracture analysis of fully ceramic microencapsulated fuel

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.


    Nuclear fuel enhancing the accident tolerance is satisfied two parts. First, the performance has to be retained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be reduced largely. 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 compared to the current commercial UO 2 fuel system. SiC ceramic has excellent properties for fuel application. SiC ceramic has low neutron absorption cross-section, excellent irradiation resistivity and high thermal conductivity. Additionally, the relative thermal conductivity of the SiC ceramic as compared to UO 2 is quite good, reducing operational release of fission products form the fuel. TRISO coating layer which is deposited on UO 2 kernel is consists of PyC/SiC/PyC trialyer and buffer PyC layer. SiC matrix composite with TRISO particle was fabricated by hot pressing. 3 to 20 wt.% of sintering additives were added to investigate reaction between sintering additives and outer PyC layer of TRISO coating layer. The relative densities of all specimens show above 92%. The reaction between sintering additives and PyC is observed in most TRISO particles, the thickness of reactants shows about ten micrometers. The thermal shock resistance of SiC matrix composite was investigated

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

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J.; Lee, S. J.


    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

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

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


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

  13. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.


    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  14. Method of producing granulated ceramic nuclear fuels

    Wilkinson, W.L.


    For the production of granulated ceramic nuclear fuels with a grain size spectrum as narrow as possible it is proposed to suspend the nuclear fuel powder in a non-aqueous solvent with small content of hydrogen (e.g. chloridized hydrocarbons) while adding a binding agent and then dry it by means of rays. As binding agent polybutyl methane acrylate in dibutyl phthalate is proposed. The method is described by the example of UO 2 -powder in trichloroethylene. The dry granulated material is produced in one working step. (UWI) [de

  15. Basic tendencies of restructured UO2 nuclear fuels fabrication industry for water-moderated reactors

    Makhova, V.A.; Bokshitskij, V.I.; Blinova, I.V.


    Processes of reformation and consolidation of firms and frontier nuclear fuels fabrication industry associated with processes of globalization and deregulation of electric power market are analyzed. Current state of nuclear fuel market and basic factors influenced on the market are presented. The role of nuclear fuel in increasing competition of NPP and fundamental directions of innovation action on the creation of perspective kinds of fuel were considered [ru

  16. Control rod effects on reaction rate distributions in tight pitched PuO2-UO2 fuel assembly

    Gil, Choong-Sup; Okumura, Keisuke; Ishiguro, Yukio


    Investigations were made for the heterogeneity effects caused by insertion or withdrawal of a B 4 C control rod on fine structure of reaction rates distributions in a tight pitched PuO 2 -UO 2 fuel assembly. Analysis was carried out by using the VIM and SRAC codes with the libraries based on JENDL-2 for the hexagonal fuel assembly basically corresponding to the PROTEUS-LWHCR experimental core. The reaction rates are affected more remarkably by the withdrawal of the control rod rather than its insertion. The changes of the reaction rates were decomposed into three terms of spectrum shifts, the changes of effective cross sections with fine groups, and their higher order components. From the analysis, it is concluded that most changes of reaction rates are caused by spectral shifts. The SRAC code with fine group constants can predict the distribution of reaction rates and their ratios with the accuracy of about 5 % except for the values related to Pu-242 capture rate, as compared with the VIM results. To increase the accuracy, it is necessary to generate the effective cross sections of the fuel near control rods with consideration of the heterogeneities in the fuel assembly. (author)

  17. Early-in-life thermal performance of UO2--PuO2 fast reactor fuel

    Baker, R.B.; Leggett, R.D.


    Results from the combined analyses of two thermal performance tests, HEDL P-19 and HEDL P-20 are described. The tests were designed to provide data on the power required to cause incipient fuel melting early in life under conditions prototypic of FFTF driver fuel pins and similar FBR fuel systems

  18. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    Carbol, P.; Spahiu, K.


    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  19. Evaluation of the internal pressure in UO2 and UO2-Gd2O3 rods of fuel assemblies 10 x 10 with the FEMAXI-Vi code

    Hernandez L, H.; Lucatero, M. A.


    Inside the acceptable criterions of fuel licensing are some that should be fulfilled in relation to the internal pressure of the fuel rods. These criterions are related with the loss of mechanical integrity due to the load excess in the pressure inside the jacket, as well as by the pressure that exercises the pellet on the jacket at the time of suffering the swelling by irradiation. This work shows the calculation of the increment of the internal pressure of the fuel rods caused by the swelling contribution of the pellets and by the accumulation of the fission gases inside the hole, pellet-jacket, in function of the burned for values of the lineal heat generation reason (LHGR) mean of fuel rods in arrangements 10 x 10. (author)

  20. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.


    The document reports in detail the fuel pin fabrication data and describes the irradiation conditions and history. All the relevant results of the non-destructive and destructive post-irradiation examinations are reported. They include: visual inspection and chemical analysis of crud; length and diameter measurements; neutron radiography and gamma scanning; juncture tests and fission gas analysis (including residual gas in fuel samples); microscopy and alpha + beta/gamma autoradiography; microprobe investigations; burn-up and isotopic analysis; and hydrogen analysis in clad. The data and observations obtained are discussed in detail and conclusions are given. The irradiation and post-irradiation examinations of the R-109 pins have shown the safe, pre-calculable performance of LWR fuel pins containing mixed-oxide sphere-pac fuel with the fissile material mainly present in the large spheres

  1. Fabrication, irradiation and post-irradiation examinations of MO2 and UO2 sphere-pac and UO2 pellet fuel pins irradiated in a PWR loop

    Linde, A. van der; Lucas Luijckx, H.J.B.; Verheugen, J.H.N.


    Three fuel pin bundles, R-109/1, 2 and 3, were irradiated in a PWR loop in the HFR at Petten during respectively 131, 57 and 57 effective full power days at average powers of approximately 39 kW.m -1 and at peak powers of approximately 60 kW.m -1 . The results of the post-irradiation examinations of these fuel bundles are presented. (Auth.)

  2. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    Crewe, Maria Teresa I.; Lopes, Paula Corain; Moura, Sergio C.; Sampaio, Jessica A.G.; Bustillos, Oscar V.


    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO 2 , U 3 O 8 , U 3 Si 2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO 2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO 2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen

  3. Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

    Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan


    The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

  4. Neutron Flux Depression in the UO2-PuO2 (15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment

    Lopez Jimenez, J.; Fernandez Marron, J.L.


    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 2 -PUO 2 (15 to 30% PUO 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

  5. Neutron flux depression in the UO2-PuO2 (15 to 30%) fuel rods from IVO-FR2-Vg7-Irradiation experiment

    Lopez Jimenez, J.; Fernandez Marron, J.L.


    The thermal-neutron flux depression within a fuel rod has a great influence on the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO 2 -PuO 2 (15 to 30% PuO 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)

  6. Factors affecting the differences in reactivity and dissolution rates between UO2 and spent nuclear fuel

    Shoesmith, D.W.; Tait, J.C.; Sunder, S.; Steward, S.; Russo, R.E.; Rudnicki, J.D.


    Strategies for the permanent disposal of spent nuclear fuel are being investigated by the U.S. Department of Energy at the Yucca Mountain site and by Atomic Energy of Canada Limited (AECL) in plutonic rock formations in the Canadian Shield. Uranium dioxide is the primary constituent of spent nuclear fuel and dissolution of the matrix is regarded as a necessary step for the release of radionuclides to repository groundwaters. In order to develop models to describe the dissolution of the U0 2 fuel matrix and subsequent release of radionuclides, it is necessary to understand both chemical and oxidative dissolution processes and how they can be affected by parameters such as groundwater composition, pH, temperature, surface area, radiolysis and redox potential. This report summarizes both published and on-going dissolution studies of U0 2 and both LWR and CANDU spent fuels being conducted at the Pacific Northwest Laboratory, Lawrence Livermore National Laboratory and Lawrence Berkeley Laboratory in the U.S. and at AECL's Whiteshell Laboratories in Canada. The studies include both dissolution tests and electrochemical experiments to measure uranium dissolution rates. The report focuses on identifying differences in reactivity towards aqueous dissolution between U0 2 and spent fuel samples as well as estimating bounding values for uranium dissolution rates. This review also outlines the basic tenets for the development of a dissolution model that is based on electrochemical principles. (author). 49 refs., 2 tabs., 11 figs

  7. Modelling of high burnup structure in UO2 fuel with the RTOP code

    Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.


    The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation

  8. Oxidation behaviour of noble-metal inclusions in used UO2 nuclear fuel

    McEachern, R.


    The literature on the chemistry of the noble-metal (Mo-Rh-Ru-Pd-Tc) inclusions found in used nuclear fuel has been reviewed. The Mo-Ru-Pd phase diagram is reasonably well understood, and the pseudoternary Mo-(Tc+Ru)-Rh+Pd) system can be used to qualitatively understand the phase chemistry of the noble-metal inclusions. The kinetics of the oxidation reaction are not particularly well understood, but they are of limited applicability to understanding the properties of used fuel. In contrast, it is important to determine the thermodynamic activity of molybdenum in noble-metal inclusions, so that analysis of their molybdenum content can be used as a probe of the local oxygen potential of the used fuel. (author)

  9. Neutronic simulation of a research reactor core of ( Th, U)O2 fuel ...


    Jan 8, 2013 ... going neutron energy for solids with cross-section σel and an angular treatment derived .... 1 MW power to consider the fuel burn-up and waste inventory. .... assembly management is required after 90 days (figure 11). As seen ...

  10. Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal


    After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

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

    Wang, R.


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

  12. Selective alpha autoradiography for monitoring thorium distribution in UO2-ThO2 fuel pellets

    Shriwastwa, B.B.; Raghunath, B.; Ghosh, J.K.


    Although natural uranium and thorium decay with similar alpha energies (4.20 and 3.98 MeV), their daughter products have different alpha characteristics. This has been exploited for selective alpha autoradiography for thoria in urania-thoria mixed nuclear fuel pellets. Difficulties in getting sufficient track density in alpha sensitive films due to the very low specific activity of natural uranium and thorium material were overcome by using a special film with annealing and pre-etching treatment. (orig./HP) [de

  13. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong


    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  14. The radial distribution of plutonium in high burnup UO2 fuels

    Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.


    A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

  15. Characterization of selenium in UO2 spent nuclear fuel by micro X-ray absorption spectroscopy and its thermodynamic stability.

    Curti, E; Puranen, A; Grolimund, D; Jädernas, D; Sheptyakov, D; Mesbah, A


    Direct disposal of spent nuclear fuel (SNF) in deep geological formations is the preferred option for the final storage of nuclear waste in many countries. In order to assess to which extent radionuclides could be released to the environment, it is of great importance to understand how they are chemically bound in the waste matrix. This is particularly important for long-lived radionuclides such as (79)Se, (129)I, (14)C or (36)Cl, which form poorly sorbing anionic species in water and therefore migrate without significant retardation through argillaceous repository materials and host rocks. We present here X-ray absorption spectroscopic data providing evidence that in the investigated SNF samples selenium is directly bound to U atoms as Se(-II) (selenide) ion, probably replacing oxygen in the cubic UO2 lattice. This result is corroborated by a simple thermodynamic analysis, showing that selenide is the stable form of Se under reactor operation conditions. Because selenide is almost insoluble in water, our data indirectly explain the unexpectedly low release of Se in short-term aqueous leaching experiments, compared to iodine or cesium. These results have a direct impact on safety analyses for potential nuclear waste repository sites, as they justify assuming a small fractional release of selenium in performance assessment calculations.

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

    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)


    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. Long-term safety of radioactive waste disposal: Chemical reaction of fabricated and high burnup spent UO2 fuel with saline brines. Final report

    Grambow, B.; Casas, I.; Pablo, J. de; Gimenez, J.; Torrero, M.E.


    This is the final report of a large EU-research project on spent fuel stability in saline repository environments. Static dissolution experiments with high burnup spent fuel samples and unirradiated UO 2 were performed for about two years in anaerobic NaCl solutions and deionized water with and without container material (iron) being present. Experiments performed at 25 and 150 C gave similar results. Dissolution rates were similar to those measured in the Swedish, or Canadian program for granite media. Rates are strongly influenced by the specific sample surface area, probably related to the mass balance of consumption and production of radiolytic oxidants. In the competition between the oxidizing effect of radiolysis and the reducing effect of iron, the metal corrosion process dominates. Processes controlling radionuclide release are matrix dissolution, solubility, coprecipitation sorption phenomena and colloid formation. In the absence of iron release rates of Sr90, Tc99, Np237, Sb125 and at low reaction progress Ru106 were controlled by matrix dissolution whereas concentrations of tetra-, hexa-, and trivalent actinides (U, Pu, Am, Cm) were controlled by solubility or coprecipitation. The presence of iron did effectively reduce the rates of fuel dissolution and the concentration of many, though not all radionuclides. Solubilities of U were similar for uniradiated UO 2 and for spent fuel both in the case of oxidizing and reducing conditions. In contrast, due to the effect of radiolysis, reaction rates of spent fuel were higher than UO 2 dissolution rates. (orig.) [de

  18. Performance of Bruce natural UO2 fuel irradiated to extended burnups

    Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.


    Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

  19. Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

    Isabelle Guénot-Delahaie


    Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

  20. Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

    Sabundjian, G.; Nakata, H.


    The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

  1. Study by electronic structure calculations of the radiation damage in the UO2 nuclear fuel: behaviour of the point defects and fission gases

    Vathonne, Emerson


    Uranium dioxide (UO 2 ) is worldwide the most widely used fuel in nuclear plants in the world and in particular in pressurized water reactors (PWR). In-pile the fission of uranium nuclei creates fission products and point defects in the fuel. The understanding of the evolution of these radiation damages requires a multi-scale modelling approach of the nuclear fuel, from the scale of the pellet to the atomic scale. We used an electronic structure calculation method based on the density functional theory (DFT) to model radiation damage in UO 2 at the atomic scale. A Hubbard-type Coulomb interaction term is added to the standard DFT formalism to take into account the strong correlations of the 5f electrons in UO 2 . This method is used to study point defects with various charge states and the incorporation and diffusion of krypton in uranium dioxide. This study allowed us to obtain essential data for higher scale models but also to interpret experimental results. In parallel of this study, three ways to improve the state of the art of electronic structure calculations of UO 2 have been explored: the consideration of the spin-orbit coupling neglected in current point defect calculations, the application of functionals allowing one to take into account the non-local interactions such as van der Waals interactions important for rare gases and the use of the Dynamical Mean Field Theory combined to the DFT method in order to take into account the dynamical effects in the 5f electron correlations. (author) [fr

  2. UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

    Pajo, L.


    In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

  3. Enhancement of actinide incineration and transmutation rates in Ads EAP-80 reactor core with MOX PuO2 and UO2 fuel

    Kaltcheva-Kouzminava, S.; Kuzminov, V.; Vecchi, M.


    Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 and PuO 2 MOX fuel elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber B4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B4 C arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second scheme H2 Zr zirconium hydride moderated zones were located in fuel elements with low enrichment (∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the reference EAP-80 core and at the same time the power density distribution remains significantly unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the inner fuel region with UO 2 and PuO 2 high enrichment plutonium fuel and the second one is the outer region with fuel elements containing zirconium hydride layer was also considered. Evolution of neutronics parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were performed using the MCNP-4B code and the SCALE 4.3 computational system. (author)

  4. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    Santos, Lauro Roberto dos


    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH)3), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an 7 industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  5. UO2-7%Gd2O3 fuel process development by mechanical blending with reprocessing of waste products and usage of densification additive

    Santos, Lauro Roberto dos


    In the nuclear fuel cycle, reprocessing and storage of 'burned' fuels, either temporary or permanent, demand high investments and, in addition, can potentially generate environmental problems. A strategy to decrease these problems is to adopt measures to reduce the amount of waste generated. The usage of integrated burnable poison based on gadolinium is a measure that contributes to achieve this goal. The reason to use burnable poison is to control the neutron population in the reactor during the early life of the fresh reactor core or the beginning of each recharging fuel cycle, extending its cycle duration. Another advantage of using burnable poison is to be able to operate the reactor with higher burning rate, optimizing the usage of the fuel. The process of manufacturing UO 2 -Gd 2 O 3 integrated burnable fuel poison generates waste that, as much as possible, needs to be recycled. Blending of Gd 2 O 3 in UO 2 powder requires the usage of a special additive to achieve the final fuel pellet specified density. The objective of this work is to develop the process of obtaining UO 2 - 7% Gd 2 O 3 integrated burnable poison using densification additives, aluminum hydroxide (Al(OH) 3 ), and reprocessing manufacturing waste products by mechanical blending. The content of 7%- Gd 2 O 3 is based on commercial PWR reactor fuels - Type Angra 2. The results show that the usage of Al(OH) 3 as an additive is a very effective choice that promotes the densification of fuel pellets with recycle up to 10%. Concentrations of 0,20 % of Al(OH) 3 were found to be the indicated amount on an industrial scale, specially when the recycled products come from U 3 O 8 obtained by calcination of sintered pellets. This is particularly interesting because it is following the steps of sintering and rectifying of the pellets, which is generating the largest amounts of recycled material. (author)

  6. Surface analysis using X-ray photoelectron spectroscopy and X-ray diffraction of UO2 fuel pellets oxidised in air at 2300C and 2700C

    Tempest, P.A.; Tyler, J.W.


    Factors which affect the UO 2 → U 3 O 8 transformation have been investigated by sequentially oxidising UO 2 fuel pellets in air at 230 0 C and 270 0 C and monitoring the growth of U 3 O 7 and U 3 O 8 using X-ray photoelectron spectroscopy, X-ray diffraction and scanning electron microscopy. Initially oxidation proceeded at a linear rate by the inward diffusion of oxygen to form a complete layer of sub-stoichiometric U 3 O 7 . This phase was tetragonal with a c/a ratio of 1.015, significantly less than the value of 1.03 measured on UO 2 powder when oxidised under identical conditions. This difference and the preferred orientation exhibited by surface grains were caused by growth stresses induced in the pellet surface. Both intergranular and transgranular cracking occurred and became nucleation sites for the growth of U 3 O 8 . The linear oxidation period associated with U 3 O 7 growth was much shorter at 270 0 C than 230 0 C and U 3 O 8 nucleated earlier. Spallation and the production of particulate were only observed during the formation of U 3 O 8 when a 30% increase in volume arose from the U 3 O 7 → U 3 O 8 phase change. (author)

  7. Experimental investigations of the meltdown phase of UO2-Zircaloy fuel rods under conditions of failure of emergency cooling

    Hagen, S.; Mack, A.; Malauschek, H.; Wallenfels, K.


    In the monoxidizing helium atmosphere at 1,850 0 C Zircaloy and UO 2 interact violently. The result is a combined meltdown of pellets and can. This phenomenon appears independent of the velocity of temperature rise. In air the oxid skin splits open at 1,890 0 C and the earlier molten material of the interior begins to flow out. When heating up to more than 2,200 0 C the oxid skin remains solid nevertheless. (orig.) [de

  8. 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

    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.

  9. Analysis of the heat and mass transfer processes of a UO2 bubble in sodium for the Fuel Aerosol Simulant Test (FAST)

    Tobias, M.L.


    The anticipated behavior of uranium oxide vapor bubbles produced by the capacitor discharge vaporization (CDV) method in the Fuel Aerosol Simulant Test (FAST) Facility is discussed on the basis of relatively simple physical models. Results of calculations for the rate of bubble rise and for heat and mass transfer rates are presented. Parametric studies indicate that future analysis efforts should emphasize the diffusion condensation process and the loss of heat from the bubble by radiation. Transfer of heat in the surrounding sodium is rapid enough that simplified models should be adequate. No important effects were noted in connection with bubble depth, initial quantity of UO 2 , or initial superheat

  10. Solubility of unirradiated UO2 fuel in aqueous solutions. Comparison between experimental and calculated (EQ3/6) data

    Ollila, K.


    The solubility behaviour of unirradiated UO 2 pellets was studied under oxic (air-saturated) and anoxic (N 2 ) conditions in deionized water, in sodium bicarbonate solutions with varying bicarbonate content (60 - 600 ppm), in Allard groundwater simulating granitic fresh groundwater conditions, and in bentonite water simulating the effects of bentonite on granitic fresh groundwater (25 deg C). The release of uranium was measured during static batch dissolution experiments of long duration (2-6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code EQ3/6 in order to evaluate solubility (steady state) limiting factors. (orig.) (26 refs., 32 figs., 13 tabs.)

  11. Oxidation kinetic changes of UO2 by additive addition and irradiation

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


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

  12. Thermal diffusivity measurements between 0 0C and 2000 0C: application to UO2

    Van Craeynest, J.C.; Weilbacher, J.C.; Lallement, R.


    We have built two types of apparatus to measure the thermal diffusivity of ceramic fuels. The first apparatus, based on Angstrom's method, operates between 0 deg. C and 1000 deg. C. Satisfactory results have been obtained for iron, nickel and molybdenum. The other apparatus, based on Cowan's method, operates between 1000 deg. C and 2000 deg. C on thin slabs. The thermal conductivity of UO 2 has been measured from 0 deg. C to 2000 deg. C. There is a good agreement between our results and the well known values for UO 2 . (authors) [fr

  13. Characterization of UO2 by infrared spectroscopy

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


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

  14. Measurements of thermal disadvantage factors in light-water moderated PuO2-UO2 and UO2 lattices

    Ohno, Akio; Kobayashi, Iwao; Tsuruta, Harumichi; Hashimoto, Masao; Suzaki, Takenori


    The disadvantage factor for thermal neutrons in light-water moderated PuO 2 -UO 2 and UO 2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4 w/o PuO 2 .UO 2 and 2.6 w/o UO 2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO 2 .UO 2 and UO 2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36 +- 0.07, 1.37 +- 0.08, 1.40 +- 0.06 and 1.38 +- 0.06 in the PuO 2 .UO 2 fuel lattices, and 1.30 +- 0.06, 1.31 +- 0.08, 1.30 +- 0.08 and 1.33 +- 0.06 in the UO 2 , for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO 2 .UO 2 and UO 2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones. (author)

  15. The long-term effect of hydrogen on the UO2 spent fuel stability under anoxic conditions: Findings from the Cigar Lake Natural Analogue study

    Bruno, Jordi; Spahiu, Kastriot


    Highlights: • We have reviewed current information on the effect of hydrogen in UO 2 spent fuel. • We explored the radiolytic models generated in the Cigar Lake project. • The Cigar Lake data supports that H 2 reduces alpha radiolysis oxidants. • The results indicate the hydrogen effect is present after 100.000 years deposition. - Abstract: The present paradigm on UO 2 spent fuel stability under anoxic conditions assumes that the potential oxidative alteration of the matrix is suppressed in the presence of the hydrogen generated by the anoxic corrosion of iron by water. The observations from the Cigar Lake Natural Analogue project indicated the long-term stability of the uraninite ore under anoxic conditions and with substantial hydrogen generation. The radiolytic models developed in the analogue project have been used to test some of the hypothesis concerning the activation of hydrogen on the uranium(IV) oxide surface. Suggestions to pathways of radiolytic oxidant consumption by other processes than uranium dioxide or sulphide oxidation are presented. The stability of the ore body for billions of year indicates the presence of processes which neutralise radiolytic oxidants and one major factor may be the presence of dissolved hydrogen in the groundwaters contacting the ore body. The results from this test would indicate that hydrogen is activated on the surface of the Cigar Lake uraninites by alpha radiation consuming the generated radiolytic oxidants

  16. Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

    Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.


    A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

  17. Innovative microstructures in ThO2-UO2 system

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


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

  18. Determination of the UO2-ZrO2-BaO equilibrium diagram

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


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

  19. Dissolution of UO2 in redox conditions

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


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

  20. Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

    Brucklacher, D.


    This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

  1. Sintering of nonstoichiometric UO2

    Susnik, D.; Holc, J.


    Activated sintering of UO 2 pellets at 1100 deg C is described. In CO 2 atmosphere is UO 2 is nonstoichiometric and pellets from active UO 2 powders sinter at 900 deg C to high density. At 1100 deg C the final sintered density is practically achieved at heating on sintering temperature. After reduction and cooling in H 2 atmosphere which is followed sintering in CO 2 the structure is identical to the structured UO 2 pellets sintered at high temperature in H 2 . Density of activated sintered UO 2 pellets is stable, even after additional sintering at 1800 deg C. (author)

  2. Slowing-down calculation for charged particles, application to the calculation of the (alpha, neutron) reaction yield in UO2 - PuO2 fuel

    Dulieu, P.


    There are no complete theory nor experimental data sufficient to predict exactly, in a systemic way, the slowing down power of any medium for any ion with any energy. However, in each case, the energy range can be divided in three areas, the low energiy range where the de/dx is an ascending energy function, the intermediate energy region where de/dx has a maximum, the high energy region where de/dx is a descending energy function. In practice, the code Irma 3 allows to obtain with a good precision de/dx for the protons, neutrons, tritons, alphas in any medium. For particles heavier than alpha it is better to use specific methods. In the case of calculating the yield of the (alpha, neutron) reaction in a UO 2 -PuO 2 fuel cell, the divergences of experimental origin, between the existing data lead to adopt a range a factor 1.7 on the yields [fr

  3. Migration behavior of palladium in UO2, (3)

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


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

  4. Procedure for the fabrication of ceramic fuel pellets with an adjustable structure

    Henke, M.; Klemm, U.; Sobek, D.


    The invention concerns a procedure for the fabrication of ceramic fuel pellets of UO 2 , PuO 2 , ThO 2 and their mixtures with an adjustable structure. Before or during the milling the particle shaped fuel pellets have been added polyethylenglycol in a 20 - 60 % aqueous solution with an amount of 0.5 - 2.0 % in weight. This additive has an effect on a controlled pore formation and grain growth advancement

  5. An Analysis of the Thermal and Structure Behaviour of the UO2-PuO2-Fuel in the Irradiation Experiment of the UO2-PuO2-Fuel in the Irradiation Experiment FR2 Capsule Test Series 5a

    Lopez Jimenez, J.; Helmut, E.


    In the Karlsruhe research reactor FR2 nine fuel pins were irradiated within three irradiation capsules in the course of the test series 5a. The pins contained UO 2 -PuO 2 fuel pellets. They reached bump values of about 6, 17 and 47 Mwd/Kg Me with linear rod powers of 400 to 600 W/cm and clad surface temperature between 500 and 700 degree centigree. A detailed analysis of the fuel structuration data (columnar-grain and equiaxed- -grain growth regions) have allowed to determine, with the help of physic-mathematical models, the radii of these regions and the heat transfer through the contact zone between fuel and clad depending on the bump. The results of the analysis showed that the fuel surface temperature rose with increasing burnup. (Author) 16 refs

  6. A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

    Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de


    In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

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

    Broadway, Jeramie; Hickman, Robert; Mireles, Omar


    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.

  8. Irradiation of mixed UO2-PuO2 oxide samples for fast neutron reactor fuel elements

    Mikailoff, H.; Mustelier, J.; Bloch, J.; Conte, M.; Hayet, L.; Lauthier, J.C.; Leclere, J.


    Thermal flux irradiation testings of small mixed oxide pellets UPuO 2 fuel elements were performed in support of the fuel reference design for the Phenix fast reactor. The effects of different parameters (stoichiometry, pellet density, pellet clad gap). on the behaviour of the oxide (temperature distribution, microstructural changes, fission gas release) were investigated in various irradiation conditions. In particular, the effect of fuel density decrease and power rate increase on thermal performances were determined on short term irradiations of porous fuels. (authors) [fr

  9. In pile programme of first valutation of UO2 + PuO2 fuel produced by a new process (GSP)

    Caracchin, R.; Lanchi, M.; Marinucci, G.; Nobili, A.; Dupont, G.; Galtier, J.


    The main scope of the ENEA-AGN-CEA programme collaboration is a first valutation of fuel elements produced by GSP method. This valuation will be done by in reactor experiment which enable to compare the performance of GSP and 'standard' FBR fuels. The composition is done by means of theree experimental device: P3, Lugel and Digel. The P3 device gives a direct measurement during irradiation of fuel central temperature, power and integral conductivity. The Lugel device measures fuel stack axial variations and Digel device gives the diameter variations of the pin and PCMI

  10. Simulations and Experimental Measurements of UO2 Thermal Conductivity

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


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

  11. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

    Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.


    The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  12. HTGR Fuel Recycle Development Program (189a OHO45). Fuel refabrication, Task 500. Rate-controlling factors in the carbothermic preparation of UO2--UC2--C microspheres

    Stinton, D.P.; Tiegs, S.M.; Lackey, W.J.; Lindemer, T.B.


    Rate controlling factors in the conversion of UO 2 + C microspheres to UC 2 + C were investigated using a 13-cm-dia fluidized bed furnace. X-ray diffraction, ion microprobe, and microstructural examination revealed that the conversion of UO 2 to UC 2 began at the surface of the microsphere and progressed toward the central unreacted core. Kinetic models for solid state reactions in spheres were evaluated by using quantitative mass spectrometric data on the rae of evolution of carbon monoxide during conversion. This analysis revealed that the rate of conversion was controlled by reaction at the outer surface of the microsphere. Also, decreased partial pressures of carbon monoxide were found to accelerate the rate of reaction

  13. Development of a thermo-kinetic diffusion model for UO2 and (U,Pu)O2 oxide fuels using the DICTRA code

    Moore, Emily Elaine


    Uranium dioxide is the most widely used nuclear fuel for light water reactors, while some countries including France make use of the uranium-plutonium (U,Pu)O 2±x mixed oxide (MOX). The MOX is also considered for future use in the Gen IV reactors, of which the sodium cooled fast reactor (SFR) is of current research interest. Both oxides exhibit a large range of non-stoichiometry due to various oxidative states of uranium and plutonium metal. Thermo-physical properties of the fuel strongly depend on deviations in composition and temperature. Extreme temperature gradients (800 K) between the center (2300 K)and periphery of the MOX fuel pellet expose a central void due to the migration and subsequent redistribution of the fuel-elements. To gain insight into the restructuring, which occurs during the fuel lifetime as well as possible accident scenarios the thermodynamic and kinetic behavior, is crucial. A comprehensive evaluation of these properties can be incorporated in computational models to describe fuel behavior over large temperature and compositions ranges, providing a predictive tool that is applicable to other parts of the fuel cycle, such as optimizing the sintering conditions for manufacturing. Atomic transport especially in UO 2 is widely treated in the experimental and computational materials communities. The current understanding of diffusion properties is limited by the stoichiometric deviations inherent to the fuel. The difficulty is apparent in experimental settings as controlling the oxygen content is problematic. Defects (interstitial and vacancy) associated with the stoichiometric deviations of the oxides facilitate the diffusion process and is of interest in regards to the restructuring of the fuel. Experimental data is widely available; however, coherence between the evaluated diffusion coefficients is not always evident. Existing computational models based on the migration of defects are often based on atomistic level simulations. A complete

  14. Behavior of UO2-Zy fuel elements of nuclear power plants up to 40000 MWj/t U

    Atabek, R.; Contenson, G. de; Houdaille, B.; Lestiboudois, G.; Vignesoult, N.


    The two principal types of fuel elements studied are unstable oxide elements in 15x15 geometry and stable oxide elements in 17x17. Semi-statistical processing of the fission gas amounts released was performed on different fuel elements at specific burn-up varying between 2000 and 40,000 MWd/t U and linear powers between 250 and 600 W/cm. This study enabled the following essential points to be stated at this burn-up level: the swelling of the oxide appears to be less than predicted by the linear law (S=0.75 %/10,000 MWd/t U); the migration of volatile fission products is relatively low and without effect on the behavior of the fuel element; strong zircaloy 4 claddings exhibit little creep and their hydriding is insignificant. On a more general level, the analyses of the fission gases performed in the fuel elements after irradiation show an increase of the fraction released with specific burn-up at a given linear power or central temperature [fr

  15. Fission product retention in TRISO coated UO2 particle fuels subjected to HTR simulated core heating tests

    Baldwin, C.A.; Kania, M.J.


    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbounded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 deg. C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 deg. C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 deg. C and above may exist. (author). 6 refs, 6 figs, 4 tabs

  16. Argentina-LLNL-LANL Comparative Sample Analysis on UO2 fuel pellet CRM-125A for Nuclear Forensics

    Kips, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)


    The recent workshop on analytical plan development provided context and background for the next step in this engagement, i.e. a comparative sample analysis on CRM 125-A. This is a commercially available certified low-enriched uranium oxide fuel pellet material from New Brunswick National Laboratory (NBL) (see certificate in Annex 1).

  17. Fission product retention in TRISCO coated UO2 particle fuels subjected to HTR simulated core heating tests

    Baldwin, C.A.; Kania, M.J.


    Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600 degree C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800 degree C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800 degree C and above may exist. 6 refs., 6 figs., 4 tabs

  18. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    Newman, D.F.; Gore, B.F.


    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  19. Critical sizes of light-water moderated UO2 and PuO2-UO2 lattices

    Tsuruta, Harumichi; Kobayashi, Iwao; Suzuki, Takenori; Ohno, Akio; Murakami, Kiyonobu


    Experimental critical sizes are presented for a total of about 250 lattices with 2.6 w/o UO 2 and 3.0 w/o PuO 2 -natural UO 2 fuel rods. The moderator was H 2 O and water-to-fuel volume ratios in the lattice cells ranged from 1.50 to 3.00 in the UO 2 lattices and from 2.42 to 5.55 in the PuO 2 -UO 2 lattices. The critical sizes were determined with the number of the fuel rods and a water level which were required to make the lattice critical in the shape of a rectangular parallelepiped over the temperature range from room temperature to 80 0 C. Reactivity variations of the PuO 2 -UO 2 lattices due to decaying of 241 Pu to 241 Am were traced during 3 years. Some critical sizes of the UO 2 and PuO 2 -UO 2 lattices with a water gap and of the UO 2 lattices with liquid poison in the moderator are also reported. Some physics parameters, such as the temperature coefficient of reactivity, the water-level worth, the reflector saving, the ratio between a migration area and an infinite multiplication factor and the critical buckling, are shown in relation to the critical sizes of the unperturbed lattices without the water gap and liquid poison. (auth.)

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

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


    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.

  1. Inductive Double-Contingency Analysis of UO2 Powder Bulk Blending Operations at a Commercial Fuel Plant (U)

    Skiles, S. K.


    An inductive double-contingency analysis (DCA) method developed by the criticality safety function at the Savannah River Site, was applied in Criticality Safety Evaluations (CSEs) of five major plant process systems at the Westinghouse Electric Corporation's Commercial Nuclear Fuel Manufacturing Plant in Columbia, South Carolina (WEC-Cola.). The method emphasizes a thorough evaluation of the controls intended to provide barriers against criticality for postulated initiating events, and has been demonstrated effective at identifying common mode failure potential and interdependence among multiple controls. A description of the method and an example of its application is provided

  2. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    Gaspar Junior, Joao Carlos Aguiar


    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  3. Tests to determine the release of short-lived fission products from UO2 fuel operating at linear powers of 45 and 60 kW/m

    Hastings, I.J.; Hunt, C.E.L.; Lipsett, J.J.; MacDonald, R.D.


    Experiments have been carried out using a 'sweep gas' technique to determine the behaviour of short-lived fission products within operating, intact UO 2 fuel elements. The Zircaloy-4-clad elements were 600 mm long and contained fuel of density 10.65 - 10.71 Mg/m 3 . A He-2% H 2 carrier gas swept gaseous or volatile fission products out of the operating fuel element past a gamma spectrometer for measurement. We outline our loop model and give full details of calculational procedures. In tests at linear powers of 45 (FIO-122) and 60 kW/m (FIO-124) to a maximum burnup of 80 MW.h/kg U, the species measured directly at the spectrometer during normal operation were generally the short-lived xenons and kryptons. Iodines were not observed during normal operation. The behaviour of I-133 and I-135 was deduced from the decay of Xe-133 and Xe-135 during reactor shutdowns. Plots of R/B (released/born) against λ (decay constant) or effective λ for all isotopes observed at 45 and 60 kW/m show that a line of slope -0.5, corresponding with diffusion kinetics, is a good fit to the measured xenon and krypton data. The inferred release of iodine fits the same line. From this we can extrapolate to an R/B for I-131 of about 5 x 10 -4 at 45 kW/m, and 3 x 10 -3 at 60 kW/m. Both tests were terminated by defects. Under defect conditions, R/B dependence on λ was about 0.6. I-131 release under defect conditions was 5 Ci and 60 mCi for FIO-122 and FI0-124, respectively. 22 refs

  4. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.


    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO 2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO 2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO 2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM) [de

  5. Deformation behavior of UO2 at temperatures above 24000C

    Slagle, O.D.


    An experimental system was developed for measuring the high-temperature creep rates of ceramic nuclear fuels to temperatures near their melting points. The results of a series of experiments carried out on UO 2 at temperatures above 2400 0 C are reported. The strain rate was found to be proportional to the 5.7 power of the stress while activation energies ranged from 250 to 340 Kcal/mole. An expression for describing the primary creep was derived from the initial time dependence of the deformation after stress application. A technique for studying the hot pressing behavior at 2580 0 C was devised but no definitive results were obtained from the first series of experiments. An empirical relationship is proposed for calculating the creep rates at very high temperatures

  6. Separation of UO2 powder

    Ristic, M.M.


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

  7. Kinetics of UO2 sintering

    Ristic, M.M.


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

  8. Operational comparison of TLD albedo dosemeters and etched-track detectors in the PuO2-UO2 mixed oxide fuel fabrication facilities

    Tsujimura, N.; Takada, C.; Yoshida, T.; Momose, T.


    Full text: The authors carried out an operational study that compared the use of TLD albedo dosemeters with etched-track detector in plutonium environments of Japan Nuclear Cycle Development Institute, Tokai Works. A selected group of workers engaged in the fabrication process of MOX (PuO 2 -UO 2 mixed oxide) fuel wore both TLD albedo dosemeters and etched-track detectors over a period from 1991 to 1993. The TLD albedo dosemeter is the Panasonic model UD-809P and the etched-track detector is the NEUTRAK (polyallyl diglycol carbonate + 1mm-t polyethylene radiator) commercially available from Nagase-Landauer Ltd. Both dosemeters were issued and read monthly. It was found that the TL readings were generally proportional to the counted etch-pits, and thus the dose equivalent results obtained from TLD albedo dosemeter agreed with those from etched-track detector within a factor of 1.5. This result indicates that, in the workplaces of the MOX plants, the neutron spectrum remained almost constant in terms of time and space, and the appropriate range of field-specific correction with spectrum variations could be small in albedo dosimetry. In addition, the calibrations of both dosemeters in the workplaces and in a bare and moderated 252 Cf calibration field were performed for quantitative validation for the results from the operational comparison. In the former experiments, locations were selected that were representative of typical neutron measurements according to the prior neutron spectra measurements with the multi-sphere spectrometer. In the latter experiments, the workplace environments were simulated by using a 252 Cf source surrounded with cylindrical steel/PMMA moderators. From both experiments, the relationship between TL readings and counted etch-pits with neutron spectrum variation was determined. As expected, the relationship obtained from the simulated workplace field calibration reproduced that from the operational comparison. (author)

  9. Modelling of UO2 oxidation in steam

    Brito, A.C.; Iglesias, F.C.; Liu, Y.


    A computer model has been developed for calculating oxidation of UO 2 at high temperatures in steam oxidising conditions. Several methods to calculate the partial pressure of oxygen in the fuel and in the environment surrounding the fuel are available. The various methodologies have been compared and the best models have been compiled into a computer model which will be implemented into fuel thermal/mechanical behaviour codes such as FACTAR 2.0 (LOECI) and ELESIM/ELOCA. Calculations from the computer model have been compared to experimental results. The calculated oxidation reaction kinetics are in good agreement with the experimental data. (author)

  10. Densification Behavior of BN-added UO2

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


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

  11. Study of phase equilibria in function of temperature in UO2-PuO2-Pu2O3 system for nuclear ceramics with high plutonium contents

    Truphemus, Thibaut


    In the UO 2 -PuO 2 -Pu 2 O 3 section, a monophasic (U 1-y ,Pu y )O 2-x domain is stable for y≤0,20 at 25 C and up to solid-liquid equilibrium. At higher Pu content, phase equilibria are more unclear with a phase separation process. The main objective of this work consisted in upgrading the representation of this system for 0,15≤y≤0,65 and 25≤T(C)≤1500. At 25 C, a miscibility gap composed by two different (U 1-y ,Pu y )O 2-x phases has been observed for y≤0,45, with one very closed to stoichiometric state (Oxygen/Metal=2) and one other very reduced. For the first time, a triphasic domain has been characterized at higher Pu contents, with two (U 1-y ,Pu y )O 2-x phases near y=0,45 and one (U 1-y ,Pu y ) 2 O 3 phase with a low U content inside. Concerning the study in function of temperature, we have demonstrated that phase separation temperature increase when Pu content grows. Several representations have been established. At 200 C, the representation is closed to that at 25 C. At 400 C, the phase separation have been specified at a lower Pu content than that of literature: y=0,35. At 600 C, our results have clarified the section, until then very unclear, with a phase separation appearing at y=0,60.The microstructural analysis has clearly demonstrated the significant impact of the phase separation on the material. Indeed many cracks have been observed in our samples, and quantity of these defects increases when Pu content grows. (author) [fr

  12. Molybdenum-UO2 cerment irradiation at 1145 K

    Mcdonald, G.


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

  13. Sinterability of mixtures of UO2 of different morphological features

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


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

  14. Compliance characteristics of cracked UO2 pellets

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.


    The thermally induced cracking of UO 2 fuel pellets causes simultaneous reductions of the bulk (extrinsic) fuel thermal conductivity and elastic moduli to values significantly less than those for solid pellets. The magnitude of these bulk properly reductions was found to be primarily dependent on the amount of crack area in the transverse plane of the fuel. The model described herein uses a simple description of the crack geometry to couple the fuel rod thermal and mechanical behaviors by relating in-reactor data to Hooke's Law and a crack compliance model. Data from the NRC/PNL Halden experiment IFA-432 show that for a typical helium-filled BWR-design rod at 30 kW/m, the effective thermal conductivity and elastic moduli of the cracked fuel are 4/5 and 1/40 of that for solid pellets, respectively

  15. Fabrication of fully ceramic microencapsulated fuel by hot pressing

    Lee, H. G.; Kim, D. J; Park, J. Y.; Kim, W. J.; Lee, S. J


    Fully ceramic microencapsulated(FCM) nuclear fuel is one of the recently suggested concept to enhance stability nuclear fuel itself. The requirements to increase the accident tolerance of nuclear fuel are mainly two parts: First, the performance has to be maintained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be decrease largely. FCM nuclear fuel consists of tristructural isotropic(TRISO) fuel particle and SiC matrix. The relative thermal conductivity of the SiC matrix as compared to UO 2 is quite good, yielding as-irradiated fuel centerline temperature compared to high temperature for the existing fuel leading to reduced stored energy in the core and reduced operational release of fission products from the fuel. Generally SiC ceramics are fabricated via liquid phase sintering due to strong covalent bonding property and low self-diffusivity coefficient. Hot pressing is very effective method to conduct sintering of SiC powder including different second phase. In this study, SiC-matrix composite including TRISO particles were sintered by hot pressing with Al 2 O 3 -Y 2 O 3 additive system. Various sintering condition were investigated to obtain high relative density above 95%. The internal distribution of TRISO particles within SiC-matrix composite was observed by x-ray radiograph. From the analysis of the cross-section of SiC-matrix composite, the fracture of TRISO particles was investigated. In order to uniform distribution of TRISO particle embedded in the SiC matrix, SiC powder overcoating is considered. SiC matrix composite including TRISO was fabricated by hot pressing. FCM pallets with full density were obtained with Al 2 O 3 -Y 2 O 3 additive system. From the microstructure image, the effect of the sintering additive contents and sintering mechanism

  16. Structure changes of irradiated UO2

    Komatsu, Junji; Yokouchi, Yoji; Kajiyama, Takashi; Terunuma, Toshihiro; Koizumi, Masumichi


    The structural change of UO 2 irradiated in GETR reactor was analyzed on void distribution, fuel cracking, and gap conductance between fuel and cladding. Metallographic analysis was carried out on 21 sections of irradiated fuel pins. Radial void distribution was measured by the linear analysis technique based on the equivalence between the volume fraction of voids and the intercepted length of lines between void boundaries. Fuel cracks were classified into two types, namely radial cracks and circumferential cracks. The radial position, length, angle and number of each fuel clad were measured on metallographic section and autoradiography. The gap conductance between fuel and cladding was calculated from the equation h = q/(T sub(s) - T sub(i)) where h is gap conductance, T sub(i) is inside clad temperature and T sub(s) is outside clad temperature. In void distribution, as the result of studying the effect of linear heat rating on the radial void fraction of UO 2 fuel irradiated with the similar level of burnup, one specimen showed that the void fraction of columnar grain growth region was comparable to that of fabricated region, and two specimens showed higher void fraction at fabricated region than the calculated one. In fuel cladding, no significant effect of burnup on fuel cracking was observed, and the number of fuel cracking increased with shutdown or scram numbers, but the radial position of circumferential cracks was not much changed. In gap conductance, it was influenced by the estimation of temperature of columnar grain growth. (Iwakiri, K.)

  17. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO2-PuO2 fuel

    Lopez Jimenez, J.; Elbel, H.


    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs

  18. FY2015 ceramic fuels development annual highlights

    Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)


    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  19. FY2016 Ceramic Fuels Development Annual Highlights

    Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)


    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  20. Tracer surface diffusion on UO2

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


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

  1. Microstructure study of AUC and UO2

    Pan Ying; Gao Dihua; Lu Huaichang


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

  2. Interfaces in ceramic nuclear fuels

    Reeve, K.D.

    Internal interfaces in all-ceramic dispersion fuels (such as these for HTGRs) are discussed for two classes: BeO-based dispersions, and coated particles for graphite-based fuels. The following points are made: (1) The strength of a two-phase dispersion is controlled by the weaker dispersed phase bonded to the matrix. (2) Differential expansion between two phases can be controlled by an intermediate buffer zone of low density. (3) A thin ceramic coating should be in compression. (4) Chemical reaction between coating and substrate and mass transfer in service should be minimized. The problems of the nuclear fuel designer are to develop coatings for fission product retention, and to produce radiation-resistant interfaces. 44 references, 18 figures

  3. The conceptual flowsheet of effluent treatment during total gelation of uranium process for preparing ceramic UO2 particles of high temperature gas-cooled reactors

    Quan Ying; Chen Xiaotong; Wang Yang; Liu Bing; Tang Yaping; Tang Chunhe


    Today, more and more people pay attention to the environmental protection and ecological environment. Along with the development of nuclear industry, many radioactive effluents may be discharged into environment, which can lead to the pollutions of water, atmosphere and soil. So radioactive effluents including low-activity and medium-level wastes solution treatments have been becoming one of significant subjects. High temperature gas-cooled reactor (HTR) is one of advanced nuclear reactors owing to its reliability, security and broad application in which the fabrication of spherical fuel element is a key technology. During the production of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was summarized which met the 'Zero Emission' demand. (authors)

  4. The effect of U3O-8 addition on the UO2 pellet

    Indrati, Y.T.; Syarif, D. G.; Handayani, A.


    The purpose of varied U 3 O 8 addition on the UO 2 pellet fabrication is to from 1-3 mu pores. The green pellets, compacted with 3 ton/cm 2 , are a mixture powder of UO 2 , TiO 2 (0.1% weight) and varied U 3 O 8 (0-12.5% weight). The green pellets were presintered by H2 atmosphere. The presintered pellets were put on the ceramic crucibles and than those were put on the SS 316 tube with argon atmosphere. The 1400 o C sintering was hold with the soaking time 3 hr and the same rate of heating and cooling 150 o C/hr. The UO 2 pellet with 5% (weight) U 3 O 8 addition has 95.17% of theoretic density and 548.4 ±6.57 VH. Based on the identification of microstructure of pellet, it is not acceptable for nuclear fuel although pellet has 10.02 mu on grain size and 1.3 mu on closed pore size. By the diffractometer X-ray, crystal structure of pellet is face centered cubic (FCC) with the O/U ratio is 2.08

  5. Study of UO2 radioinduced densification

    Stora, J.P.; Bruet, M.


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

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

    Carbajo, J.J.


    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.

  7. Development of thermocouple re-instrumentation technique for irradiated fuel rod. Techniques for making center hole into UO2 pellets and thermocouple re-instrumentation to fuel rod

    Shimizu, Michio; Saito, Junichi; Oshima, Kunio


    The information on FP gas pressure and centerline temperature of fuel pellets during power transient is important to study the pellet clad interaction (PCI) mechanism of high burnup LWR fuel rods. At the Department of JMTR, a re-instrumentation technique of FP gas pressure gage for an irradiated fuel rod was developed in 1990. Furthermore, a thermocouple re-instrumentation technique was successfully developed in 1994. Two steps were taken to carry out the development program of the thermocouple re-instrumentation technique. In the first step, a drilling technique was developed for making a center hole of the irradiated fuel pellets. Various drilling tests were carried out using dummy of fuel rods consisted of Ba 2 FeO 3 pellets and Zry-2 cladding. On this work it is important to keep the pellets just the state cracked at a power reactor. In these tests, the technique to fix the pellets by frozen CO 2 was used during the drilling work. Also, diamond drills were used to make the center hole. These tests were completed successfully. A center hole, 54mm depth and 2.5mm diameter, was realized by these methods. The second step of this program is the in-pile demonstration test on an irradiated fuel rod instrumented dually a thermocouple and FP gas pressure gage. The demonstration test was carried out at the JMTR in 1995. (author)

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

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


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

  9. Application of fully ceramic microencapsulated fuels in light water reactors

    Gentry, C.; George, N.; Maldonado, I.; Godfrey, A.; Terrani, K.; Gehin, J.


    This study performs 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 (PWR). Using uranium-based fuel and Pu/Np-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 fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO 2 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 Pu/Np-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, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

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

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


    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.

  11. Low Temperature Two-Steps Sintering (LTTSS) - an innovative method for consolidating porous UO2 pellets

    Sanjay Kumar, D.; Ananthasivan, K.; Senapati, Abhiram; Venkata Krishnan, R.


    Metallic uranium and its alloys are an important fuel for fast reactors. Presently, metallic uranium is being prepared using expensive fluoro-metallothermic process. Recent reports suggest that metal oxide could be reduced to metal using a novel electrochemical de-oxidation method and this could serve as attractive alternate for expensive metallothermic process. In view of which, a research program is being pursued in our Centre to develop an optimum process parameter for the scaled up preparation of metallic uranium efficiently. One of the important process parameter is the size, nature and distribution of porosity in the urania pellet. Essentially the ceramic form of the urania should encompass interconnected porosity that would allow percolation of melts into the UO 2 . However, the matrix density of the pellet should be high to ensure that it possesses good handling strength and is electrically conducting. Hence preparation of high dense porous UO 2 pellets was required. In this study, we report the preparation of porous UO 2 pellets possessing a very high matrix density by using the citrate gel-combustion method. The 'as-prepared' powders were consolidated at various compaction pressures as such and these pellets were sintered in 8 mol %Ar+H 2 gas with a flow rate of 250 mL/min at 1073 K for 30 min followed by soaking at 1473 K for 4 h with heating rate of 5 K min -1 in a molybdenum furnace. X-ray diffraction studies revealed that these pellets contained UO 2 . The morphological analysis sintered pellets was carried out by using Scanning Electron Microscope (M/s. Philips model XL 30, Netherlands). All these pellets were gold coated

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

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


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

  13. Microstructure characterization of ceramic nuclear fuel

    Boehmert, J.; Gaessner, W.


    A system of characterizing methods is described based on quantitative ceramographic methods. This system is applicable in quality assurance of UO 2 nuclear fuel in small-scale production and for determining microstructural parameters in scientific investigations. The system is based essentially on the measuring of microstructural parameters by the methods of linear analysis by the VEB Carl Zeiss Jena EPIQUANT mechanical optical microstructural analyzer. It is completed by measuring the pore size using automatic the television analyzer QTM. Before the quantitative microstructural characterization, in each case the morphology of the structure is estimated qualitatively. (author)

  14. Method of manufacturing UO2 pellet

    Harada, Yuhei; Asami, Yasuji.


    The present invention concerns a method of manufacturing UO 2 pellets with less FP gas release and having fine structure for moderating PCMI. At first, oxide nuclear fuel pellets are placed in a sintering furnance and preliminarily sintered in a H 2 gas atmosphere at 1400 - 1600 degC. In this step, sintering is progressed to about 90 % TD, by which closed cells are formed substantially completely. Then, when sintering is further advanced at an identical temperature in a CO 2 gas atmosphere, growth of the crystal grains is advanced at the central portion of the pellets. Then, reductive heat treatment is applied at the identical temperature in a H 2 gas atmosphere. As a result, pellets having a fine double structure with the larger grain size region being in the central portion and smaller grain size region in the outer periphery can be obtained. (I.J.)

  15. Results of REIMEP '89 UO2 pellet

    Mayer, K.; Alonso, A.; Bievre, P. de; Lycke, W.; Bolle, W. de; Gallet, M.; Hendrickx, F.


    The interest in the safeguards of fissile material focuses on a limited number of compounds which play key roles in the nuclear fuel cycle. Amongst these materials Uranium Dioxide pellets are of considerable importance as they enter the reactors in order to generate energy. In LWR's pellets with an initial 235 U content of about 3 mass % are used, whereas natural or depleted material is applied for the breeding zone in FBR's. The 89/90 round o REIMEP covered Uranium materials with 235 U abundances in the range of natural or depleted material. UO 2 pellets were distributed to 21 laboratories for analysis. The participating laboratories were asked to determine the Uranium content and the isotopic composition of the material. The results reported by the participants are presented as graphs thus giving a picture of the state-of-the-practice

  16. Production of Depleted UO2Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

    Collins, J.L.


    The main objective of the Depleted UO 2 Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO 2 kernels with diameters of 500 ± 20 (micro)m and 3.5 kg of UO 2 kernels with diameters of 350 ± 10 (micro)m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO 2 kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO 3 · 2H 2 O microspheres to form dense UO 2 kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 ± 10-(micro)m-diameter kernels, and to obtain very high yields

  17. Sintering diagrams of UO2

    Mohan, A.; Soni, N.C.; Moorthy, V.K.


    Ashby's method (see Acta Met., vol. 22, p. 275, 1974) of constructing sintering diagrams has been modified to obtain contribution diagrams directly from the computer. The interplay of sintering variables and mechanisms are studied and the factors that affect the participation of mechanisms in UO 2 are determined. By studying the physical properties, it emerges that the order of inaccuracies is small in most cases and do not affect the diagrams. On the other hand, even a 10% error in activation energies, which is quite plausible, would make a significant difference to the diagram. The main criticism of Ashby's approach is that the numerous properties and equations used, communicate their inaccuracies to the diagrams and make them unreliable. The present study has considerably reduced the number of factors that need to be refined to make the sintering diagrams more meaningful. (Auth.)

  18. A review and discussion of candidate ceramics for immobilization of high-level fuel reprocessing wastes

    Hayward, P.J.


    This review discusses and attempts to evaluate 11 of the leading ceramic processes for hosting the high-level and high-level plus medium-level wastes which would arise from the reprocessing of used UO 2 , (Th,Pu)O 2 and (Th,U)O 2 fuels. The wasteform materials considered include glass ceramics, supercalcine ceramics, SYNROC ceramics, 'stuffed glass', titanate ceramics, cermets, clay ceramics, cement-based materials and multibarrier wasteforms. Although no attempt has been made to rank these candidates in order of superiority, the conclusion is drawn that, of the materials proposed so far, a glass ceramic appears to be best suited to the Canadian program, taking into account durability in the potential environment of a flooded vault, ability to withstand radiation and transmutation damage without serious loss of durability, ability to accommodate variable waste compositions, and ease of processing and quality control. This conclusion does not necessarily apply to other national waste management programs. However, many of the points raised might be included in any critical assessment of alternative wasteform materials

  19. Industrial ceramics

    Mengelle, Ch.


    After having given the definition of the term 'ceramics', the author describes the different manufacturing processes of these compounds. These materials are particularly used in the fields of 1)petroleum industry (in primary and secondary reforming units, in carbon black reactors and ethylene furnaces). 2)nuclear industry (for instance UO 2 and PuO 2 as fuels; SiC for encapsulation; boron carbides for control systems..)

  20. Use of UO 2 films for electrochemical studies

    Miserque, F.; Gouder, T.; Wegen, D. H.; Bottomley, P. D. W.


    UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water-gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

  1. Photochemical synthesis of UO2 nanoparticles

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


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

  2. Oxidation of UO2 at 150 to 3500C

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


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

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

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


    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.

  4. Factors Affecting the Sintering of UO2 Pellets

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


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

  5. Heat transfer coefficient between UO2 and Zircaloy-2

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


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

  6. On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

    Walker, C.T.; Staicu, D.; Sheindlin, M.; Papaioannou, D.; Goll, W.; Sontheimer, F.


    A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation

  7. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.


    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  8. Measurement of the friction coefficient between UO2 and cladding tube

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


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

  9. Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

    Mogensen, M.; Walker, C.T.


    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. (orig.)

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

    Amblard, M.


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

  11. An improved UO2 thermal conductivity model in the ELESTRES computer code

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


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

  12. Experimental Observation of Densification Behavior of UO2 Annular Pellet

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


    Recently, in the nuclear industry, one of the major issues is the improvement of a fuel economy. And many efforts have been made to develop a nuclear fuel for a high burnup and extended cycle. In the development of a high performance fuel, in-reactor fuel behavior (fission gas release, pellet-clad interaction, stress corrosion cracking, cladding corrosion, etc.) must be seriously reconsidered. Also, fuel fabrication (high enriched UO 2 powder handling, fuel rod and assembly manufacturing, fabricated fuel rod and assembly storage and transport, etc.) and an enrichment process (5 w/o criticality limit, etc.) must be discussed. A modification and an improvement of the nuclear fuel system will be also required. The typical fuel geometry of a PWR (Pressurized Water Reactor) is composed of a cylindrical pellet with a tubular cladding. And the outer surface of the cladding is cooled with water. However, to allow a substantial increase in the power density, an additional cooling is needed. One of the best ways is the application of the new fuel geometry that is of annular shape and has both internal and external cooling. From this point of view, the double cooled fuel is being developed by KAERI (Korea Atomic Energy Research Institute), and as a part of the project, the development of a fabrication process of a UO 2 annular pellet is now in progress. The dimensional behavior of UO 2 fuel is an important parameter in an irradiation performance. Various investigations (resintering test, model calculation, in-pile dimensional change measuring, etc.) had been performed. In designing a double cooled fuel, the importance of the dimensional behavior of a fuel pellet is higher, because the gap distance between a pellet and cladding can considerably affect on the in reactor fuel performance (gap conductance). And the dimensional behavior of an inner/outer gap is different with a cylindrical pellet, when the pellet shrinks (densification), the inner gap distance decreases and the

  13. Technological aspects of UO2 sintering at low temperature

    Thern, Gerardo G.; Dominguez, Carlos A.; Benitez, Ana M.; Marajofsky, Adolfo


    Within the Fuel Cycle Program of CNEA, the knowledge that plant personnel has on sintering at low temperature was evaluated, because this process could decrease costs for UO 2 and (U,Gd)O 2 pellets production, simplify the furnace maintenance and facilitate the automation of the production process, specially convenient for uranium recovery. By applying this technology, some companies have achieved production at pilot-scale and irradiated a significant number of pellets. (author)

  14. UO2 pellet and manufacturing method

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


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

  15. Geometrical dimensioning of PWR UO2 pellets

    Silva, A.T.


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

  16. Creep of UO2 at 25000C

    Slagle, O.D.


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

  17. Porosity influence on UO2 pellet fracture

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


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

  18. Recycling process of Mn-Al doped large grain UO2 pellets

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


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

  19. Oxidation of UO2 at 150 to 3500C

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


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

  20. Three-dimensional single-channel thermal analysis of fully ceramic microencapsulated fuel via two-temperature homogenized model

    Lee, Yoonhee; Cho, Nam Zin


    Highlights: • Two-temperature homogenized model is applied to thermal analysis of fully ceramic microencapsulated (FCM) fuel. • Based on the results of Monte Carlo calculation, homogenized parameters are obtained. • 2-D FEM/1-D FDM hybrid method for the model is used to obtain 3-D temperature profiles. • The model provides the fuel-kernel and SiC matrix temperatures separately. • Compared to UO 2 fuel, the FCM fuel shows ∼560 K lower maximum temperatures at steady- and transient states. - Abstract: The fully ceramic microencapsulated (FCM) fuel, one of the accident tolerant fuel (ATF) concepts, consists of TRISO particles randomly dispersed in SiC matrix. This high heterogeneity in compositions leads to difficulty in explicit thermal calculation of such a fuel. For thermal analysis of a fuel element of very high temperature reactors (VHTRs) which has a similar configuration to FCM fuel, two-temperature homogenized model was recently proposed by the authors. The model was developed using particle transport Monte Carlo method for heat conduction problems. It gives more realistic temperature profiles, and provides the fuel-kernel and graphite temperatures separately. In this paper, we apply the two-temperature homogenized model to three-dimensional single-channel thermal analysis of the FCM fuel element for steady- and transient-states using 2-D FEM/1-D FDM hybrid method. In the analyses, we assume that the power distribution is uniform in radial direction at steady-state and that in axial direction it is in the form of cosine function for simplicity. As transient scenarios, we consider (i) coolant inlet temperature transient, (ii) inlet mass flow rate transient, and (iii) power transient. The results of analyses are compared to those of conventional UO 2 fuel having the same geometric dimension and operating conditions

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

    Fast, Ivan


    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

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

    Fast, Ivan


    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

  3. Influence of the interpellet space to the Instant Release Fraction determination of a commercial UO2 Boiling Water Reactor Spent Nuclear Fuel

    Martínez-Torrents, A.; Serrano-Purroy, D.; Casas, I.; De Pablo, J.


    The contact of the coolant with the fuel pin during irradiation produces a gradient of temperature in the fuel pellet that segregates the radionuclides (RN) depending on its volatility and reactivity. This segregation determines the Instant Release Fraction (IRF), an important source of radiological risk in the performance assessment (PA) of a Deep Geologic Repository (DGR). RN segregation was studied radially in previous papers. In the present work, it was studied axially, taking into special consideration the cutting position of the solid sample to be studied. Iodine and caesium were the RN with the highest release, while the contribution of rubidium, strontium, molybdenum and technetium to the IRF depended on their chemical state. The interpellet presence (known also as dishing) effect was clearly observed for caesium, increasing its release by one order of magnitude. According to these results, one of the major contributions to the IRF comes from the RN trapped in the dishing and has to be considered in the sampling and data interpretation that will be performed for the PA of the DGR.

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

    Feghhi, Seyed Amir Hossein; Rezazadeh, Marzieh; Kadi, Yacine; ); Tenreiro, Claudio; Aref, Morteza; Gholamzadeh, Zohreh


    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 he adopted because a high fissile production rate of 233 U converted from 232 Th 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. (author)

  5. Reactivity measurements on an experimental assembly of 4.31 wt % 235U enriched UO2 fuel rods arranged in a shipping cask geometry

    Bierman, S.R.


    A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs

  6. Deposition of Cr, Nb, V, and Ti coatings on UO2-25w/oPuO2 fuel pellets by sputtering

    Gibby, R.L.; McClanahan, E.D.


    A sputtering deposition process was developed for application of metallic coatings on either the ends or circumferences of LMFBR mixed-oxide fuel pellets. Coatings of Cr, Nb, V and Ti were applied to over 860 pellets. Ceramography, emission spectrography, and spark source spectroscopy were used to characterize the coatings. Coating thicknesses were controlled to within +-0.0005 cm (0.0002 inch) for a coating thickness of 0.00127 cm (0.0005 inch) on the circumference and 0.00254 cm (0.001 inch) on the ends of pellets. Chemical impurities in the coatings were generally less than 0.5 wt percent. The coatings were adherent in all cases, although some interfacial separations were noted with Ti coatings. The results indicated that further optimization of coatings' parameters would result in improvement of the coatings

  7. Multiphase Nanocrystalline Ceramic Concept for Nuclear Fuel

    Mecartnery, Martha [Univ. of California, Irvine, CA (United States); Graeve, Olivia [Univ. of California, San Diego, CA (United States); Patel, Maulik [Univ. of Liverpool (United Kingdom)


    The goal of this research is to help develop new fuels for higher efficiency, longer lifetimes (higher burn-up) and increased accident tolerance in future nuclear reactors. Multiphase nanocrystalline ceramics will be used in the design of simulated advanced inert matrix nuclear fuel to provide for enhanced plasticity, better radiation tolerance, and improved thermal conductivity

  8. Multiphase Nanocrystalline Ceramic Concept for Nuclear Fuel

    Mecartnery, Martha; Graeve, Olivia; Patel, Maulik


    The goal of this research is to help develop new fuels for higher efficiency, longer lifetimes (higher burn-up) and increased accident tolerance in future nuclear reactors. Multiphase nanocrystalline ceramics will be used in the design of simulated advanced inert matrix nuclear fuel to provide for enhanced plasticity, better radiation tolerance, and improved thermal conductivity

  9. Thermal-mechanical properties of cracked UO2 pellets

    Williford, R.E.; Mohr, C.L.; Lanning, D.D.


    A series of experiments (IFA-431, 432, 513, and 527) sponsored by the Fuel Behavior Research Branch of the USNRC are being irradiated in the Halden Boiling Water Reactor to better define LWR fuel behavior over the normal operating range of power reactor fuel rods. One fuel behavior variable of interest is the thermally induced cracking of UO 2 fuel pellets. The effects of pellet cracking on the effective thermal conductivity and elastic moduli for the fragmented fuel were found to be primarily dependent on the free area in the r, theta plane of the fuel rod. The free area is defined as the area within the cladding inner surface that is not occupied by the fuel fragments themselves

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

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


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

  11. Unirradiated UO2 in irradiated zirconium alloy sheathing

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


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

  12. Fabrication of ThO2, UO2, and PuO2-UO2 pellets

    Rasmussen, D.E.; Jentzen, W.R.; McCord, R.B.


    Fabrication of ThO pellets for EBR-II irradiation testing and fabrication of UO 2 and PuO 2 -UO 2 pellets for United Kingdom Prototype Fast Reactor (PFR) irradiation testing is discussed. Effect of process parameters on density and microstructure of pellets fabricated by the cold press and sinter technique is reviewed

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

    Rachmawati, M.


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

  14. UO2 microspheres obtainment through the internal gelation methods

    Sterba, M.E.; Gomez Constenla, A.


    UO 2 microspheres obtainment process through the internal gelation method which allows the spheres' obtainment of uniform size is detailed herein, varying the same among 0.3 and 1.7 mm of diameter. The sintered density reaches 10.78 g/cm 3 , permitting the fuels fabrication dispersed and vibro-compacted fuels. The trichloroethylene use implementation as gelation agent is described, thus reducing the number of stages in the microspheres fabrication. At the same time, the uranium sun composition has been modified so as to be compatible with the use solvent. (Author)

  15. Automation system for production of UO2 granules

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


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

  16. Boiling point measurements on liquid UO2

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


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

  17. Thermal modeling of the ceramic composite fuel for light water reactors

    Revankar, S.T.; Latta, R.; Solomon, A.A.


    Full text of publication follows: Composite fuel designs capable of providing improved thermal performance are of great interest in advanced reactor designs where high efficiency and long fuel cycles are desired. Thermal modeling of the composite fuel consisting of continuous second phase in a ceramic (uranium oxide) matrix has been carried out with detailed examination of the microstructure of the composite and the interface. Assuming that constituent phases are arranged as slabs, upper and lower bounds for the thermal conductivity of the composite are derived analytically. Bounding calculations on the thermal conductivity of the composite were performed for SiC dispersed in the UO 2 matrix. It is found that with 10% SiC, the thermal conductivity increases from 5.8 to 9.8 W/m.deg. K at 500 K, or an increase of 69% was observed in UO 2 matrix. The finite element analysis computer program ANSYS was used to create composite fuel geometries with set boundary conditions to produce accurate thermal conductivity predictions. A model developed also accounts for SiC-matrix interface resistance and the addition of coatings or interaction barriers. The first set of calculations using the code was to model simple series and parallel fuel slab geometries, and then advance to inter-connected parallel pathways. The analytical calculations were compared with the ANSYS results. The geometry of the model was set up as a 1 cm long by 400 micron wide rectangle. This rectangle was then divided into one hundred sections with the first ninety percent of a single section being UO 2 and the remaining ten percent consisting of SiC. The model was then meshed using triangular type elements. The boundary conditions were set with the sides of the rectangle being adiabatic and having an assigned temperature at the end of the rectangle. A heat flux was then applied to one end of the model producing a temperature gradient. The effective thermal conductivity was then calculated using the geometry

  18. Development of ultrasonic technique for measure of porosity of UO2 pellets

    Baroni, Douglas Brandao


    The characterization of nuclear fuel is of great importance to guarantee the efficiency and even the safety in the power stations. At present, the techniques used implicate elevated costs with equipment, materials and installations of radiological protection. Besides, because of being destructive techniques, they impose that the checking of the characteristics of this material is done by sampling. In this work a not destructive technique was developed for measures of porosity in ceramic materials with efficiency and precision. The objective of this work is to this technique will be able to be used in laboratory practice for measures in UO 2 pellets, so it would become viable the inspection of up to 100% of the nuclear fuel, guaranteeing bigger control of the characteristics of the used material, turning in increasing safety, efficiency and economy. The innovation of the technique is due to the fact of analysing the specter of frequency of the ultrasonic wrist, and not his time of course in the material, frequently used. In this work 40 ceramic pellets of alumina were used with values of porosity between 5,09% and 37,30%. A system of recognition of signs using artificial neural networks made possible to distinguish pellets with differences of porosity of 0,04%. It was observed that this technique can be used for several others aims, for example, in the determination of the void fraction in regimen of two-phase flow, what is very important to guarantee the efficiency and safety of nuclear reactors. (author)

  19. Production of nuclear ceramic fuel for nuclear power plants at 'Ulba metallurgical plant' OSC

    Khadeev, V.G.


    The paper describes the flow-sheet of production of uranium dioxide powders and nuclear ceramic fuel pellets of them existing at the facility. 'UMP' OSC applies ADU extraction process of UO2 powders production. An indisputable success of the process is the possibility of use of the wide range of raw materials. Uranium hexafluoride, uranium oxides, uranium metal, uranium tetrafluoride, uranyl salts, uranium ore concentrates, all possible types of uranium-containing materials the processing of which by routine methods is difficult (ashes, scraps, etc.) are used as the raw materials. In addition, a reprocessed nuclear fuel can be used for fuel production. The quality of uranium dioxide powder produced does not depend on the type of uranium raw material used. High selectivity of extraction refining makes possible to obtain material with rather low impurities content that meets practically all specifications for uranium dioxide known to us. Ceramic and process features of uranium dioxide powders, namely, specific surface, bulk density, grain size and sinterability make possible to produce nuclear ceramic fuel with specified features. Quality of uranium dioxide powders produced by 'UMP' OSC was highly rated by General Electric company that is one of the leading companies from fuel manufactures in the USA market . It has certified 'UMP' OSC as its supplier. Currently, our company makes great efforts on establishing production of uranium dioxide powders with natural isotopes content for production of fuel for CANDU reactors. Trial lots of such powders are under tests at some companies manufacturing fuel for this type reactors in Canada, USA and Corea

  20. Thermal and Mechanical Properties of UO2 and PuO2

    Kato, M.; Matsumoto, T.


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

  1. Effect of alpha irradiation on UO2 surface reactivity in aqueous media

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


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

  2. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    Sen, R. Sonat; Pope, Michael A.; Ougouag, Abderrafi M.; Pasamehmetoglu, Kemal O.


    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

  3. Cracking and healing behavior of UO2 as related to pellet-cladding mechanical interaction. Interim report, July 1976

    Kennedy, C.R.; Yaggee, F.L.; Voglewede, J.C.; Kupperman, D.S.; Wrona, B.J.; Ellingson, W.A.; Johanson, E.; Evans, A.G.


    A direct-electrical-heating apparatus has been designed and fabricated to investigate those nuclear-fuel-related phenomena involved in the gap closure-bridging annulus formation mechanism that can be reproduced in an out-of-reactor environment. Prototypic light-water-reactor UO 2 fuel-pellet temperature profiles have been generated utilizing high flow rates (approximately 700 liters/min) of helium coolant gas, and a recirculating system has been fabricated to permit tests of up to 1000 h. Simulated light-water-reactor single- and multiple-thermal-cycle experiments will be conducted on both unclad and ceramic (fused silica) clad UO 2 pellet stacks. A laser dilatometer with a resolution of 1.27 x 10 -2 mm (5 x 10 -4 in.) is used to measure pellet dimensional increase continuously during thermal cycling. Acoustic emissions from thermal-gradient cracking have been detected and correlated with crack length and crack area. The acoustic emissions are monitored continuously to provide instantaneous information about thermal-gradient cracking. Posttest metallography and fracture-mechanics measurements are utilized to characterize cracking and crack healing

  4. Behaviour of the UO2/clayey water. A spectroscopic approach

    Guilbert, S.


    This work deals with the disposal of spent nuclear fuels in deep geological layers. After three years of irradiation, these fuels are constituted of 95 % UO 2 . It is then indispensable to know the leaching behaviour of this solid because ground waters are the main agents of dispersion to biosphere of the radioelements contained in these fuels. This work includes alteration tests carried out with a device allowing to synthesize a clayey water equilibrated with a partial pressure in CO 2 in oxidizing or reducing conditions. After the tests, the solid and the solution have been characterized in order to establish a balance of the alteration. The UO 2 matrix has been characterized by XPS. The uranium in solution has been titrated by ICP-MS. In oxidizing conditions, after some weeks, the dissolution velocity of UO 2 has stabilized around 3*10 11 mol/m 2 .s. This velocity is of 4*10 12 mol/m 2 .s in a reducing medium. The uranium concentrations in the oxidized water are of about 2*10 4 mol/l after two years of leaching. After 33 days of alteration in a reducing medium, the uranium amount is of 3*10 6 mol/l. The XPS technique has revealed a superficial and progressive oxidation of the uranium(IV) and the formation of U-OH bonds in the oxidizing medium. A U(VI)/U(IV) ratio has been determined by this technique. It has stabilized around 2 in some weeks. In reducing conditions, this ratio is stable and is of about 0.5. Modeling tools have allowed to propose a class of solids potentially able to control the uranium solubility. In oxidizing conditions, the uranyl hydrates (schoepite) evolve towards uranyl silicates which are thermodynamically more stable. In reducing conditions, a control of the uranium concentration in solution by U 4 O 9 is probable. (O.M.)

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

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


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

  6. Nuclear fuel elements

    Kawada, Toshiyuki; Hirayama, Satoshi; Yoneya, Katsutoshi.


    Purpose: To enable load-depending operation as well as moderation for the restriction of operation conditions in the present nuclear reactors, by specifying the essential ingredients and the total weight of the additives to UO 2 fuel substances. Constitution: Two or more additives selected from Al 2 O 3 , B 2 O, CaO, MgO, SiO 2 , Na 2 O and P 2 O 5 are added by the total weight of 2 - 5% to fuel substances consisting of UO 2 or a mixture of UO 2 and PuO 2 . When the mixture is sintered, the strength of the fuel elements is decreased and the fuel-cladding interactions due to the difference in the heat expansion coefficients between the ceramic fuel elements and the metal claddings are decreased to a substantially harmless degree. (Horiuchi, T.)

  7. Mechanism for transient migration of xenon in UO2

    Liu, X.-Y.; Uberuaga, B. P.; Andersson, D. A.; Stanek, C. R.; Sickafus, K. E.


    In this letter, we report recent work on atomistic modeling of diffusion migration events of the fission gas product xenon in UO 2 nuclear fuel. Under nonequilibrium conditions, Xe atoms can occupy the octahedral interstitial site, in contrast to the thermodynamically most stable uranium substitutional site. A transient migration mechanism involving Xe and two oxygen atoms is identified using basin constrained molecular dynamics employing a Buckingham type interatomic potential. This mechanism is then validated using density functional theory calculations using the nudged elastic band method. An overall reduction in the migration barrier of 1.6-2.7 eV is obtained compared to vacancy-mediated diffusion on the uranium sublattice.

  8. Contribution to the study of UO2 pellet fabrication

    Fogaca Filho, N.; Gentile, E.F.; Mourao, M.B.; Souza Santos, T.D. de; Haydt, H.M.


    The establishment of a set of parametric comparisons related to UO 2 powders of two different origins as the ammonium diuranate and the ammonium uranyl carbonate is presented. It is emphasized the importance due to the pressing capability of the powders and the requirement for homogeneous microstructure for both, the pore distribution and the grain size. In order to establish the parameters of comparison, all the required normal tests for the in-process control of fabrication of fuel elements for nuclear power reactors were performed, particularly to the re-sintering test, in view of the evaluation of dimensional stability of the pellets [pt

  9. The production of sinterable UO2 from AUC

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


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

  10. Acoustic emission during the compaction of brittle UO2 particles

    Hegron, Lise


    One of the options considered for recycling minor actinides is to incorporate about 10% to UO 2 matrix. The presence of open pores interconnected within this fuel should allow the evacuation of helium and fission gases to prevent swelling of the pellet and ultimately its interaction with the fuel clad surrounding it. Implementation of minor actinides requires working in shielded cell, reducing their retention and outlawing additions of organic products. The use of fragmentable particles of several hundred micrometers seems a good solution to control the microstructure of the green compacts and thus control the open porosity after sintering. The goal of this study is to monitor the compaction of brittle UO 2 particles by acoustic emission and to link the particle characteristics to the open porosity obtained after the compact sintering. The signals acquired during tensile strength tests on individual granules and compacts show that the acoustic emission allows the detection of the mechanism of fragmentation and enables identification of a characteristic waveform of this fragmentation. The influences of compaction stress, of the initial particle size distribution and of the internal cohesion of the granules, on the mechanical strength of the compact and on the microstructure and open porosity of the sintered pellets, are analyzed. By its ability to identify the range of fragmentation of the granules during compaction, acoustic emission appears as a promising technique for monitoring the compaction of brittle particles in the manufacture of a controlled porosity fuel. (author) [fr

  11. Equi-axed and columnar grain growth in UO2

    White, R.J.


    The grain size of UO 2 is an important parameter in the actual performance and the modelling of the performance of reactor fuel elements. Many processes depend critically on the grain size, for example, the degree of initial densification, the evolution rate of stable fission gases, the release rates of radiologically hazardous fission products, the fission gas bubble swelling rates and the fuel creep. Many of these processes are thermally activated and further impact on the fuel thermal behavior thus creating complex feedback processes. In order to model the fuel performance accurately it is necessary to model the evolution of the fuel grain radius. When UO 2 is irradiated, the fission gases xenon and krypton are created from the fissioning uranium nucleus. At high temperatures these gases diffuse rapidly to the grain boundaries where they nucleate immobile lenticular shaped fission gas bubbles. In this paper the Hillert grain growth model is adapted to account for the inhibiting ''Zener'' effects of grain boundary fission gas porosity on grain boundary mobility and hence grain growth. It is shown that normal grain growth ceases at relatively low levels of irradiation. At high burnups, high temperatures and in regions of high temperature gradients, columnar grain growth is often observed, in some cases extending over more than fifty percent of the fuel radius. The model is further extended to account for the de-pinning of grains in the radial direction by the thermal gradient induced force on a fission gas grain boundary bubble. The observed columnar/equi-axed boundary is in fair agreement with the predictions of an evaporation/condensation model. The grain growth model described in this paper requires information concerning the scale of grain boundary porosity, the local fuel temperature and the local temperature gradient. The model is currently used in the Nuclear Electric version of the ENIGMA fuel modelling code. (author). 14 refs, 3 figs, 1 tab

  12. Neutronics Studies Of Uranium-Based Fully Ceramic Micro-Encapsulated Fuel For PWRs

    Maldonado, G. Ivan; Gehin, Jess C.


    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

  13. Mechanism of UO2 selfdisintegration by oxidation

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


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

  14. TCA UO2/MOX core analyses

    Tahara, Yoshihisa; Noda, Hideyuki


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

  15. The crystal structure of ianthinite, [U24+(UO2)4O6(OH)4(H2O)4](H2O)5: a possible phase for Pu4+ incorporation during the oxidation of spent nuclear fuel

    Burns, P.C.; Hawthorne, F.C.; Miller, M.L.; Ewing, R.C.


    Ianthinite, [U 4+ 2 (UO 2 ) 4 O 6 (OH) 4 (H 2 O) 4 ](H 2O) 5 , is the only known uranyl oxide hydrate mineral that contains U 4+ , and it has been proposed that ianthinite may be an important Pu 4+ -bearing phase during the oxidative dissolution of spent nuclear fuel. The crystal structure of ianthinite, orthorhombic, a=0.7178(2), b=1.1473(2), c=3.039(1) nm, V=2.5027 nm 3 , Z=4, space group P2 1 cn, has been solved by direct methods and refined by least-squares methods to an R index of 9.7% and a wR index of 12.6% using 888 unique observed [ vertical stroke F vertical stroke ≥5σ vertical stroke F vertical stroke ] reflections. The structure contains both U 6+ and U 4+ . The U 6+ cations are present as roughly linear (U 6+ O 2 ) 2+ uranyl ions (Ur) that are in turn coordinated by five O 2- and OH - located at the equatorial positions of pentagonal bipyramids. The U 4+ cations are coordinated by O 2- , OH - and H 2 O in a distorted octahedral arrangement. The Urφ 5 and U 4+ φ 6 (φ: O 2- , OH - , H 2 O) polyhedra link by sharing edges to form two symmetrically distinct sheets at z∼0.0 and z∼0.25 that are parallel to (001). The sheets have the β-U 3 O 8 sheet anion-topology. There are five symmetrically distinct H 2 O groups located at z∼0.125 between the sheets of Uφ n polyhedra, and the sheets of Uφ n polyhedra are linked together only by hydrogen bonding to the intersheet H 2 O groups. The crystal-chemical requirements of U 4+ and Pu 4+ are very similar, suggesting that extensive Pu 4+ U 4+ substitution may occur within the sheets of Uφ n polyhedra in the structure of ianthinite. (orig.)

  16. Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

    Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.


    SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

  17. UO2 production process with methanol washing

    Sondermann, T.


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

  18. Crystal-field effect in UO2

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


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

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

    Liu feiming


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

  20. Possible effects of oxidation on the transient release of fission gas from UO2

    Stoner, H.C.; Matthews, J.R.; Wood, M.H.


    The effect of varying the fuel composition from UO 2 to UOsub(2.3), on the transient behaviour of fission gas is simulated on the assumption that surface diffusion behaves in a similar manner to volume diffusion. The results may help in the understanding of fuel behaviour after pin failure in accident conditions in thermal reactor systems. (author)

  1. Leaching patterns and secondary phase formation during unsaturated leaching of UO2 at 90 degrees C

    Wronkiewicz, D.J.; Bates, J.K.; Gerding, T.J.; Veleckis, E.; Tani, B.S.


    Experiments are being conducted that examine the reaction of UO 2 with dripping oxygenated ground water at 90 degrees C. The experiments are designed to identify secondary phases formed during UO 2 alteration, evaluate parameters controlling U release, and act as scoping tests for studies with spent fuel. This study is the first of its kind that examines the alteration of UO 2 under unsaturated conditions expected to exist at the proposed Yucca Mountain repository site. Results suggest the UO 2 matrix will readily react within a few months after being exposed to simulated Yucca Mountain conditions. A pulse of rapid U release, combined with the formation of dehydrated schoepite on the UO 2 surface, characterizes the reaction between one to two years. Rapid dissolution of intergrain boundaries and spallation of UO 2 granules appears to be responsible for much of the U released. Differential release of the UO 2 granules may be responsible for much of the variation observed between duplicate experiments. Less than 5 wt % of the released U remains in solution or in a suspended form, while the remaining settles out of solution as fine particles or is reprecipitated as secondary phases. Subsequent to the pulse period, U release rates decline and a more stable assemblage of uranyl silicate phases are formed by incorporating cations from the ground water leachant. Uranophane, boltwoodite, and sklodowskite appear as the final solubility limiting phases that form in these tests. This observed paragenetic sequence (from uraninite to schoepite-type phases to uranyl silicates) is identical to those observed in weathered zones of natural uraninite occurrences. The combined results indicate that the release of radionuclides from spent fuel may not be limited by U solubility constraints, but that spallation of particulate matter may be an important, if not the dominant release mechanism affecting release

  2. Proposal for Ultrasonic Technique for evaluation elastic constants in UO2 pellets

    Lopes, Alessandra Susanne Viana Ragone; Baroni, Douglas Brandao; Bittencourt, Marcelo de Siqueira Queiroz; Souza, Mauro Carlos Lopes


    Pellets of uranium dioxide are used as fuel in nuclear power reactors, in which are exposed to high thermal gradients. This high energy will initiate fusion in the central part of the pellet. The expansion of the uranium dioxide pellets, resulting from fission products, can cause fissures or cracks, therefore, the study of their behavior is important. This work aims to develop and propose an ultrasonic technique to evaluate the elastic constants of UO 2 pellets. However, because of the difficulties in handling nuclear material, we proposed an initial study of alumina specimens. Alumina pellets are also ceramic material and their porosity and dimensions are in the similar range of dioxide uranium pellets. They also are used as thermal insulation in the fuel rods, operating under the same conditions. They were fabricated and used in two different sets of 10 alumina pellets with densities of 92% and 96%. The developed ultrasonic technique evaluates the traveling time of ultrasonic waves, longitudinal and transverse, and correlates the observed time and the elastic constants of the materials. Equations relating the speed of the ultrasonic wave to the elastic modulus, shear modulus and Poisson's ratio have led to these elastic constants, with graphics of correlation that showed excellent agreement with the literature available for Alumina. In view of the results and the ease of implementation of this technique, we believe that it may easily be used for dioxide uranium pellets, justifying further studies for that application. (author)

  3. A prediction of the inert gas solubilities in stoichiometric molten UO2

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


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

  4. Modeling of UO2 aqueous dissolution over a wide range of conditions

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


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

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

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


    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.

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

    Holub, F.


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

  7. SEM hot stage sintering of UO2

    Miller, D.J.


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

  8. Nuclear fuel conversion and fabrication chemistry

    Lerch, R.E.; Norman, R.E.


    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)

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

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


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

  10. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    Appelhans, A.D.


    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  11. UO2 dissolution rates: A review

    McKenzie, W.F.


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

  12. Irradiation behaviour of UO2/Mo porous cermets for thermionic converters

    Stora, J.P.; Kauffmann, Y.


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

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

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


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

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

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


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

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

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


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

  16. Analysis of a MOX-UO2 interface by the method of characteristics

    Chetaine, A.; Erradi, L.; Sanchez, R.; Zmijarevic, I.; Aniel-Buchheit, S.


    In the last few years many studies have been done to improve the ability of core reactors (PWR and BWR) to burn Plutonium fuel, either in mixed UO 2 /MOX pattern or full MOX pattern. The analysis of a MOX-UO 2 interface with the method of characteristics has been carried out. Comparisons with Monte Carlo and collision-probability calculations show that our results are in good agreement with those obtained by reference methods and qualify the method of characteristic as a reliable technique for such calculations. (authors)

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

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


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

  18. Synthesis and investigation of uranyl molybdate UO2MoO4

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


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

  19. Heat transfer from internally-heated molten UO2 pools

    Stein, R.P.; Baker, L. Jr.; Gunther, W.H.; Cook, C.


    Experimental measurements of heat transfer from internally heated pools of molten UO 2 have been obtained for two cell sizes: 10 cm x 10 cm and 20 cm x 20 cm. The experiments with the large cell have supported a previous conclusion from early small data that the measured downward heat fluxes are higher than would be expected on the basis of considerations of thermal convection. A convective model underpredicts the downward heat fluxes by a factor of 2.5 to 4.5 for all but one early experiment. Arbitrary assumptions of increased thermal conductivity do not account for the discrepancy. A single model based on internal thermal radiation heat transfer is able to account for the high values. The model uses the optically thick Rosseland approximation. Because of this, it is tentatively concluded that thermal radiation plays a dominant role in controlling the heat transfer from internally heated molted fuel

  20. HELIOS calculations for UO2 lattice benchmarks

    Mosteller, R.D.


    Calculations for the ANS UO 2 lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries. The results obtained from the different libraries permit conclusions to be drawn regarding the adequacy of the energy group structures and of the ENDF/B-VI evaluation for 238 U. Scandpower A/S, the developer of HELIOS, provided Los Alamos National Laboratory with six different cross section libraries. Three of the libraries were derived directly from Release 3 of ENDF/B-VI (ENDF/B-VI.3) and differ only in the number of groups (34, 89 or 190). The other three libraries are identical to the first three except for a modification to the cross sections for 238 U in the resonance range

  1. Prediction of minimum UO2 particle size based on thermal stress initiated fracture model

    Corradini, M.


    An analytic study was employed to determine the minimum UO 2 particle size that could survive fragmentation induced by thermal stresses in a UO 2 -Na Fuel Coolant Interaction (FCI). A brittle fracture mechanics approach was the basis of the study whereby stress intensity factors K/sub I/ were compared to the fracture toughness K/sub IC/ to determine if the particle could fracture. Solid and liquid UO 2 droplets were considered each with two possible interface contact conditions; perfect wetting by the sodium or a finite heat transfer coefficient. The analysis indicated that particles below the range of 50 microns in radius could survive a UO 2 -Na fuel coolant interaction under the most severe temperature conditions without thermal stress fragmentation. Environmental conditions of the fuel-coolant interaction were varied to determine the effects upon K/sub I/ and possible fragmentation. The underlying assumptions of the analysis were investigated in light of the analytic results. It was concluded that the analytic study seemed to verify the experimental observations as to the range of the minimum particle size due to thermal stress fragmentation by FCI. However the method used when the results are viewed in light of the basic assumptions indicates that the analysis is crude at best, and can be viewed as only a rough order of magnitude analysis. The basic complexities in fracture mechanics make further investigation in this area interesting but not necessarily fruitful for the immediate future

  2. Concept and nuclear performance of direct-enrichment fusion breeder blanket using UO2 powder

    Oka, Yoshiaki; Kasahara, Takayasu; An, Shigehiro


    A new concept is presented for direct enrichment of fissile fuel in the blanket of a fusion-fission hybrid reactor. The enriched fuel produced by this means can be used in fission reactors without reprocessing. The outstanding feature of the concept is the powdered form in which UO 2 fuel is placed in the reactor blanket, where it is irradiated to the requisite enrichment for use as fuel in burner reactor, e.g. 3%. After removal from blanket, the powder is mixed to homogenize the enrichment. Fuel pellets and assemblies are then fabricated from the powder without reprocessing. The concept of irradiating UO 2 in powder eliminates the problems of spatial nonuniformity in fissile enrichment, and of radiation damage to fuel clad, encountered in attempting to enrich prefabricated fuel. Powder mixing for homogenization brings the additional benefit of removing volatile fission products. Also burnable poison can be added, as necessary, after irradiation. An extensive neutronic parameter survey showed that the optimum blanket arrangement for this enrichment concept is one presenting a fission suppressing configuration and with beryllium adopted as moderator. By this arrangement, the average 239 Pu enrichment obtained on the natural UO 2 fuel in the blanket reaches 3% after only 0.56 MW.yr/m"2 exposure. A conceptual design is presented of the blanket, together with associated fusion breeder, from which, practical application of the concept is shown to be promising. (author)

  3. Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores

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


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

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

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


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

  5. Data report on leach tests of Pu-doped UO2 in PBB1 brine: Salt Repository Project

    Gray, W.J.


    This report provides results from a series of leach tests conducted using nonirradiated uranium dioxide (UO 2 ) doped with plutonium (Pu) to simulate the alpha activity of spent fuel specimens used in recent spent fuel leach tests. The purpose was to determine whether alpha radiation from the spent fuel could be responsible for uranium release values in spent fuel leach tests in salt brine that were at least 100 times greater than from similar tests with nonirradiated UO 2 pellets. The data in this data report are preliminary; they have been neither analyzed nor evaluated. 2 refs., 2 figs., 8 tabs

  6. Sorption of Np by UO2 under repository conditions

    Kazakovskaya, T. V.; Zakharova, E. V.; Haire, M. J.


    This work is a part of the joint Russian - American Program on Beneficial Use of Depleted Uranium. The production of nuclear fuels results in the accumulation of large quantities of depleted uranium (DU) in the form of uranium hexafluoride (UF6), which is converted to uranium oxides. Depleted uranium dioxide (DUO2) can be used as a component of radiation shielding and as an absorbent for migrating radionuclides that may emerge from casks containing spent nuclear fuel (SNF) that are stored for hundreds of thousands of years in high-level wastes (HLW) and SNF repositories (e.g. Yucca Mountain Project). In this case DU oxides serve as an additional engineered chemical barrier. It is known that the primary radioisotope contributor to the calculated long-term radiation dose to the public at the Yucca Mountain SNF repository site boundary is neptunium-237 (237Np). This paper describes the sorption of 237Np in various media (deionized water and J-13 solution) by DUO2. Samples of DUO2 used in this work originated from the treatment of UF6 in a reducing media to form UO2(DUO2-1 at 600°C, DUO2-2 at 700°C, and DUO2-3 at 800°C). All species of DUO2 sorb Np(V) and Np(IV) from aqueous media. Equilibrium was achieved in 24 hours for Np(V) and in 2 hours for Np(IV). Np(V) sorption is accompanied with partial reduction of Np(V) to Np(IV) and vice versa. The sorption of Np(V) onto DUO2 surfaces is irreversible. The investigations on DUO2 transformations were performed under dynamic and static conditions. Under static conditions the solubility of the DUO2 samples in J-13 solution is considerably higher than in DW. When the pre-treatment temperature is decreased, the solubility of DUO2 samples raises regardless of the media. The experiments on interaction between DUO2 and aqueous media (DW and J-13 solution) under dynamic conditions demonstrated that during 30-40 days the penetration/filtration rate of DW and J-13 solution through a thin DUO2 layer decreased dramatically, and then

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

    Bain, A.S.


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

  8. Spectral shift controlled reactor, UO2 once-through cycle optimized


    This paper presents technical and economic data on the SSCR which may be of use in the International Fuel Cycle Evaluation Program to intercompare alternative nuclear systems. Included in this data is information on the optimized UO 2 once-through fuel cycle. The ''optimized'' cycle refers to a UO 2 once-through cycle which has better fuel resource utilization than the conventional UO 2 cycle employed in current design PWRs. This fuel cycle uses more in-core batches and a higher discharge exposure than current PWR fuel management schemes. The proposed cycle is not optimal in a mathematical sense, however, since additional resource savings can be obtained if the discharge exposure is extended to even higher values and the number of in-core fuel batches is increased further. The present cycle was selected as ''optimal'' based on the assumption that it can be achieved with only an extension of fuel design technology and can therefore be deployed in a relatively short time frame. In the longer term, modification to reactor geometry as well as further extensions of discharge burnup might be considered to realize additional reduction in uranium resource requirements. The data contained in this paper has been developed by an ongoing program which at the present time is only 50% complete. The data presented here should therefore be considered preliminary and will be updated in the future as required

  9. The dissolution rate of UO2 in the alkaline regime under oxidizing conditions using a simplified ground water analog

    Leider, H.R.; Nguyen, S.N.; Weed, H.C.; Steward, S.A.


    The major factor controlling the long term release of radionuclides from spent fuel in a geologic repository is the leaching/dissolution by groundwater of the UO 2 matrix, since more than 90% of the radionuclide waste is contained in the fuel matrix. The objective of this investigation is to provide experimental dissolution rates for UO 2 samples which can be used to develop a mechanistic release model (or models) for UO 2+x (x≥0) under repository conditions. Several types of data will be obtained from this study: (1) the dissolution rates of UO 2 as a function of pI-L temperature, carbonate and oxygen fugacity; (2) the comparison of the steady state dissolution rates of ''not-reduced'' versus ''reduced'' UO 2 samples and of single crystal versus polycrystalline UO 2 under identical experimental conditions; (3) the pre- and post-test surface analyses of the samples to provide information on the surface phases that may be formed under experimental conditions

  10. Behavior of UO2 and FISSIUM in sodium vapor atmosphere at temperatures up to 28000C

    Feuerstein, H.; Oschinski, J.


    In case of a HCDA a rubble bed of fuel debris may form under a sodium pool and reach high temperatures. 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 UO 2 were measured up to 2800 0 C. The evaporation was found to be a complex process, depending on temperature and the 'active' surface. Evaporation restructures the surface of the samples, however no new 'active' surface is formed. UO 2 forms sometimes well shaped crystals and curious erosion products. The efficiency of the used condenser/filter lines was higher than 99.99%. In case of a HCDA all the evaporated substances will condense in the soidum pool. Thermal reduction of the UO 2 reduces the oxygen potential of the system. The final composition at 2500 0 C was found to be UO 1.95 . The only influence of the sodium vapor was found for the diffusion of UO 2 into the thoria of the crucible. Compared with experiments in an atmosphere of pure argon, the diffusion rate was reduced. (orig.) [de

  11. Review on quality control techniques of UO2 pellets under pilot-plant conditions, at Instituto de Energia Atomica, Sao Paulo, Brazil

    Souza Santos, T.D. de; Haydt, H.M.; Gentile, E.F.; Ambrozio Filho, F.; Quadros, N.F.; Fogaca Filho, N.


    The Instituto de Energia Atomica's Metallurgy Division Pilot Plant has been established to develop fabrication and control techniques of ceramic fuel elements, to train personnel and to acquire experience in quality control of fuel pellets. Its close association with the Institute's Chemical Engineering Division, where pilot-plant development on uranium and thorium purification is carried out, affords a direct way to ascertain the influence of salt processing variables on the behaviour of oxides derived from such uranium salts (ammonium diuranate and ammonium uranyl carbonate). The pilot plant, with a capacity of about 5 tons of UO 2 pellets per year, has ample flexibility in equipment, installations and procedures for such work, comprising uranium salts calcining, UO 2 reduction, UO 2 pellet fabrication, sintering, inspection, centerless grinding and adequate controls, both on powders and on pellets produced. It comprises several self-contained sub-units, corresponding to each particular operation, arranged in such a way that work can be carried independently and asssuring in each good control of accountablity.Quality control techniques are exerted both on powder and on pellets lines. In the powders line, besides the current routine control tests, special ones have been developed and used, comprising grain size microscopy, electron scanning examination of particle shape and sedimentation tests. These controls allow fabrication of oxide powders (mainly natural uranium up to now) to meet the specifications for the particular programs that have been tackled. In the pellets line, with ample flexibility on fabrication steps, both low- (90 to 93 pct density) and medium-density pellets (93 to 95 pct) are produced. Besides the usal routine controls, special tests on quantitative pore and grain sizes distribution through quantitative optical microscopy, electron scanning microscopy and fractographic tests were developed to evaluate the homogeneity and the geometry of pore

  12. Advanced ceramic cladding for water reactor fuel

    Feinroth, H.


    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  13. Development of irradiated UO2 thermal conductivity model

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


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

  14. All ceramic structure for molten carbonate fuel cell

    Smith, James L.; Kucera, Eugenia H.


    An all-ceramic molten carbonate fuel cell having a composition formed of a multivalent metal oxide or oxygenate such as an alkali metal, transition metal oxygenate. The structure includes an anode and cathode separated by an electronically conductive interconnect. The electrodes and interconnect are compositions ceramic materials. Various combinations of ceramic compositions for the anode, cathode and interconnect are disclosed. The fuel cell exhibits stability in the fuel gas and oxidizing environments. It presents reduced sealing and expansion problems in fabrication and has improved long-term corrosion resistance.

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

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


    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.

  16. Safety and licensing of MOX versus UO2 for BWRs and PWRs: Aspects applicable for civilian and weapons grade Pu

    Goldstein, L.; Malone, J.


    This paper reviews the safety and licensing differences between MOX and UO 2 BWR and PWR cores. MOX produced from the normal recycle route and from weapons grade material are considered. Reload quantities of recycle MOX assemblies have been licensed and continue to operate safely in European LWRs. In general, the European MOX assemblies in a reload are 2 . These studies indicated that no important technical or safety related issues have evolved from these studies. The general specifications used by fuel vendors for recycled MOX fuel and core designs are as follows: MOX assemblies should be designed to minimize or eliminate local power peaking mismatches with co-resident and adjacently loaded UO 2 assemblies. Power peaking at the interfaces arises from different neutronic behavior between UO 2 and MOX assemblies. A MOX core (MOX and UO 2 or all-MOX assemblies) should provide cycle energy equivalent to that of an all-UO 2 core. This applies, in particular, to recycle MOX applications. An important consideration when burning weapons grade material is rapid disposition which may not necessarily allow for cycle energy equivalence. The reactivity coefficients, kinetics data, power peaking, and the worth of shutdown systems with MOX fuel and cores must be such to meet the design criteria and fulfill requirements for safe reactor operation. Both recycle and weapons grade plutonium are considered, and positive and negative impacts are given. The paper contrasts MOX versus UO 2 with respect to safety evaluations. The consequences of some transients/accidents are compared for both types of MOX and UO 2 fuel. (author)


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


    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

  18. New interpretation on formation of UO2 Post-Accident Heat Removal particulate in sodium

    Schins, H.


    A comparative experimental study on quenching in sodium of four molten fuel materials, UO 2 Al 2 P 3 , Cu and stainless steel, is presented. Experimental results like temperatures, pressures, particle shapes, particle size distributions, crack patterns and crystal grain sizes are given and interpreted. These fuel-coolant interactions (FCI) can be understood as all being characterized by transition boiling of sodium. The fuel is first fragmented by the sodium vapor bubble growth and collapse process. These particulates have smooth surfaces. The two materials, UO 2 and Al 2 O 3 , are fragmented further by a delayed mechanism which is thermal stress shrinkage cracking. Delayed particles are fragments of larger ones. Furthermore, attention is drawn to the theoretical results which show that pure FCI-particulate is significantly finer

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

    Boybul; Arif Nugroho


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

  20. The effect of UO2 density on fission product gas release and sheath expansion

    Notley, M.J.F.; MacEwan, J.R.


    The effect of UO 2 density on fission product gas release and sheath expansion has been determined in an irradiation experiment in which the performance of fuel elements with densities between 10.42 and 10.74 g/cm 3 was compared at ∫λdθ values of 39 and 42 W/cm. The elements were irradiated as clusters of four in a pressurized water loop, hence their irradiation histories were identical. Fission product gas release and the extend of grain growth were greater for the lower density elements. Both effects can be attributed solely to the variation of the thermal conductivity of the fuel with the fractional porosity p, if λ p λ [1 - (2.6 ± 0.8) p] where λ is the thermal conductivity of fully dense UO 2 and λ p is that of the porous UO 2 . This expression is in agreement with laboratory findings. A correlation between the extent of grain growth in the UO 2 and the fractional gas release was found to exist in this test and was shown to apply in a large number of other fuel irradiations. Diametral sheath strain was lower for the low density fuel elements than for those of high density, although the former were deduced to have operated with higher central temperatures. It is supposed that the thermal expansion of the fuel can be partially accommodated by elimination of some of the original porosity. The data are consistent with the assumption that approximately half the porosity in the region of the fuel undergoing grain growth is eliminated. (author)

  1. Study on factors affecting sintering density of Gd2O3-UO2 pellets

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


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

  2. On the sintering kinetics in UO2

    Marajofsky, A.


    The fabrication process of UO 2 pellets from powders involve pressing and a sintering anneal at high temperature (1650 deg. C to 1750 deg. C) during two or more hours in a hydrogen atmosphere. An alternative method is the oxidative sintering, made at lower temperature (1000 deg. C to 1300 deg. C) in a CO 2 or CO/CO 2 atmosphere. The sintering phenomena consist in the densification of the material by a thermal treatment below the fusion point. For a compact made by pressing a powder, sintering is the process of annulation of the porosity present in the compact or pellet. Several theories describe the sintering phenomena dividing it in three stages, initial, intermediate and final: in all of them the densification is a continuous growing function of time. Nevertheless it has been experimentally reported that a reduction of the density occurs in the third step of the sintering. The phenomena has been called solarization. Solarization has been attributed to the effect of the evolved gases from additives or to the CO 2 atmosphere in oxidative sintering. Thus, it is convenient to distinguish between solarization in oxidative or reducing conditions. Reducing solarization is a consequence of the tendency towards equilibrium of intergranular pores. In oxidative sintering it occurs in the reducing anneal after the sintering and is due to the change in the lattice parameter. This work shows examples of both types of solarization and qualitative interpretation of this phenomena. Both situations show the need of strict control of the sintering and powder production conditions. (author)

  3. Ceramic grade (U,Pu)O2 powder fabrication

    Cristallini, O.A.; Villegas de Maroto, Marina; De Pino, J.I.; Osuna, H.A.


    Ceramic grade UO 2 powder was obtained by the homogeneous precipitation method. This procedure was afterwards applied to the fabrication of ceramic grade (U,Pu)O 2 powders, and mixed oxide powders with Pu content ranging from 0.7 to 16% were obtained. The obtainment of mixed ceramic oxides as well as the recuperation of fabrication scraps were developed in three steps: 1)study of the process of homogeneous precipitation of ammonium diuranate (ADU); 2) co-precipitation of ADU/PuO 2 .H 2 O for Pu concentrations of 0.6 and 6.8; 3) the thermal conditioning to mixed oxide (U,Pu)O 2 powders. The experimental procedure involves the following steps: preparation of the PuO 2 (NO 3 ) 4 solution; co-precipitation of the PuO 2 (NO 3 ) 2 solution with an UO 2 (NO 3 ) 2 solution; filtration and drying of the precipitate, thermal treatment and finally, mixing, pressing and sintering of the (U,Pu)O 2 and Nukem UO 2 powder with a 0. of zinc stearate. Different controls were made by means of physical, chemical and ceramographic tests. This method can be used for the fabrication of fast reactor fuels or, previous mechanical dispersion in UO 2 powder, for the fabrication of thermal reactors fuels. (M.E.L.) [es

  4. Dissolution mechanism of UO2 at various parametric conditions

    Ollila, K.


    The aim of this experimental study is to investigate the solubility and dissolution mechanism of uranium dioxide under simulated disposal conditions of spent fuel. Unirradiated UO 2 is used as a surrogate for spent fuel. Two types of synthetic groundwaters were used in these experiments, on simulating the natural conditions deep in granitic bedrock (synthetic groundwater I) and the other simulating the effects of bentonite on groundwater (synthetic groundwater II). The effect of carbonate concentration was investigated by following dissolution in sodium bicarbonate solution as a function of bicarbonate concentration. Deionized wate was used as a reference water. All the experiments were carried out under both air-saturated, oxidizing and anoxic, reducing conditions. A separate test series under anoxic conditions was initiated in order to study the oxidation state of uranium. The experimental uranium solubilities are compared with the solubilities obtained from theoetical calculations by applying the geochemical code PHREEQ. The theoretical solubility values of uranium under oxidizing conditions calculated by PHREEQE are higher when compared to the corresponding experimental solubility values. The reason for the lower solubility values may be the mechanism of dissolution leading for example either to a situation where low dissolution rate is a limiting factor or to formation of some solid phase of uranium with lower solubility. Formation of a surface layer was observed on the pellet after dissolution in synthetic groundwater II. The theoretical solubility values under educing conditions calculated for uranium by PHREEQE appear to be in good agreement with the experimental solubility values

  5. quality assurance calculation in UO2 pellet manufacturing process

    Can, S.; Acarkan, S.; Guereli, L. and others


    A process qualification plan is prepared for preparation of quality assurance documentation in accordance with ISO-9000 series of standards, for sintered UO 2 pellets manufactured in the Nuclear Fuel Technology Department. The objectives of this plan are to determine quantitatively and statistically process capability of the pellet production, to check product properties (are) in conformance with specifications at the pre-( ) confidence levels, to prepare necessary documents and to assess the results. The product properties taking into account are chemical composition, cracks, density, microstructure and grain size. The statistical parameters used for qualification element of quality assurance are calculated.Statistical values for sintered pellets are: LENGTH/WEIGHT/DIAMETER/DENSITY/%TD: MEAN:13,395/16,808/12,293/10,679/97,400 STD:0,1651/ 0,252/0,0212/0,015/0,140. It was seen that sintered pellets manufactured in the Nuclear Fuel Technology Department meet the criteria within 95% confidence level. In this paper specifications, criteria and calculations will be explained in detail

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

    Contreras Guzman, Ariel; Arlegui Hormazabal, Oscar


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

  7. UO2 leaching and radionuclide release modelling under high and low ionic strength solution and oxidation conditions


    In this work, the UO 2 dissolution under oxidizing conditions has been studied in order to compare these results to those obtained with spent fuel. Two different leaching solutions have been used, one with a high ionic strength trying to simulate the conditions expected in a saline repository and the other at low ionic strength much appropriate to granitic environments. In both cases, the dissolution has been studied studied as a function of pH, redox potential, oxidants, complexing agents, particle size as well as the experimental methodology. Results can be summarized as follows: a) The UO 2 dissolution is rather independent on ionic strength. b) Dissolution rates can be explained in general independent on the oxidant as: Log R=3DK [oxidant] Surface solid evolution is very important to understand the dissolution/oxidation mechanism of UO 2 . d) Under oxidizing conditions, the dissolution is H+ and HCO 3 promoted. e) In carbonate medium, both UO 2 and spent fuel dissolution rates are very similar, while in a non-complexing medium, spent fuel dissolution rate is much higher than the UO 2 one. This fact seems to indicate that radiolysis is much important non-complexing media. (Author)

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

    Matzke, H.


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

  9. UO2 corrosion in high surface-area-to-volume batch experiments

    Bates, J. K.; Finch, R. J.; Hanchar, J. M.; Wolf, S. F.


    Unsaturated drip tests have been used to investigate the alteration of unirradiated UO 2 and spent UO 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

  10. Influence of a microwave radiation on dissolution kinetics of UO2, CeO2, and Co3O4 in nitric environment

    Joret, Laurent


    This research thesis addresses the issue of dissolution oxides present in spent nuclear fuels. As previous studies outlined important increases of oxide dissolution rate when submitted to microwaves, the issue is then to apply such a technique to PuO 2 which is the most difficult oxide to dissolve. As plutonium may be handled only in certified laboratories and under strict safety conditions, the author studied the influence of a microwave radiation on the dissolution kinetics of other and various metallic oxides in a nitric environment. The choice of this nitric environment is imposed by conditions met in the nuclear industry. Oxides are chosen according to two criteria: dissolution times ranging from few minutes to few days, various responses to electromagnetic radiation (different values for the real and imaginary parts of their dielectric permittivity). Three oxides are retained: UO 2 and CeO 2 (to model PuO 2 ) and Co 3 O 4 . After a recall of some theoretical aspects of the response of a dielectric material to an electromagnetic field, a comparison between conventional and microwave heating, the author presents the main results obtained by using microwaves in chemistry (organic synthesis, ceramic sintering, acid dissolution). He reports the experimental study of nitric dissolution of oxides by conventional heating, and the dielectric characterisation of the studied oxides. He presents the experimental microwave set-up, and reports and discusses experimental results obtained for the dissolution of UO 2 , CeO 2 and Co 3 O 4 in HNO 3 [fr

  11. Dissolution of intact UO2 pellet in batch and rotary dissolver conditions

    Jayendra Kumar Gelatar; Bijendra Kumar; Sampath, M.; Shekhar Kumar; Kamachi Mudali, U.; Natarajan, R.


    Comparative dissolution of intact un-irradiated UO 2 pellet of PHWR fuel dimensions was performed in batch and dynamic rotary dissolver conditions in aqueous nitric acid solutions at elevated temperatures. The extent of dissolution was estimated by determining the uranium concentration of the resulting aqueous solution. It was observed that rate of dissolution was much faster in dynamic conditions as compared to static batch conditions. (author)

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

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


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

  13. UO2/magnetite concrete interaction and penetration study

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


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

  14. Geometric dimensioning of UO2 pellets for PWR

    Teixeira e Silva, A.


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

  15. Correlations between different methods of UO2 pellet density measurement

    Yanagisawa, Kazuaki


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

  16. Non-instrumented capsule design of HANARO irradiation test for the high burn-up large grain UO2 pellets

    Kim, D. H.; Lee, C. B.; Oh, D. S.


    Non-instrumented capsule was designed to irradiate the large grain UO 2 pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. UO 2 pelletes will be irradiated up to the burn-up higher than 70 MWD/kgU in HANARO. To irradiate the UO 2 pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. In addition, to satisfy the safety criteria of HANARO such as prevention of ONB(Onset of Nucleate Boiling), fuel melting and wear damage of the capsule during the long term irradiation, design of the non-instrumented capsule was optimized

  17. A Review of Fragmentation Models Relative to Molten UO2 Breakup when Quenched in Sodium Coolant

    Cronenberg, A.W.; Grolmes, M.A.


    An important aspect of the fuel-coolant interaction problem relative to liquid metal fast breeder reactor (LMFBR) safety analysis is the fragmentation of molten oxide fuel during contact with liquid sodium coolant. A proper description of the kinetics of such an event requires an understanding of the breakup process and an estimate of the size and dispersion of such finely divided fuel in coolant. In recent years, considerable interest has centered on the problem of determining the nature of such fragmentation. In this paper, both analytic and experimental studies pertaining to such breakup are reviewed in light of recent developments in the understanding of heat transfer and solidification phenomena during quenching of UO 2 in sodium. A more extensive review of this subject can be found in Ref. 1. In conclusion: As discussed, a number of models have been proposed in an attempt to understand the nature of the UO 2 fragmentation process. The four principle mechanisms considered likely to cause such fragmentation (impact forces, boiling, violent gas release, and shell solidification) have been developed to the point where comparative analysis is possible. In addition, recent developments in the understanding of the physics of oxide fuel behavior in sodium coolant (boiling regime criteria, vapor nucleation theories, and prediction of solidification kinetics enable us to asses whether or not the various model assumptions are realistic. In view of this knowledge the following conclusions are made. For the case of hydrodynamic influence on fragmentation, it can be said that although the disruptive forces of impact and viscous drag may contribute to breakup, their effects are not controlling with respect to high temperature materials, including UO 2 -sodium. With respect to the vapor bubble growth and collapse mechanism it was shown that for sodium quenching, where coolant contact may, be expected (as opposed to water), the thermodynamic work potential of the bubble is

  18. Microbes make average 2 nanometer diameter crystalline UO2 particles.

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


    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

  19. Experimental characterization and modelling of UO2 mechanical behaviour at high temperatures and high strain rates

    Salvo, Maxime


    The aim of this work is to characterize and model the mechanical behavior of uranium dioxide (UO 2 ) during a Reactivity Initiated Accident (RIA). The fuel loading during a RIA is characterized by high strain rates (up to 1/s) and high temperatures (1000 C - 2500 C). Two types of UO 2 pellets (commercial and high density) were therefore tested in compression with prescribed displacement rates (0.1 to 100 mm/min corresponding to strain rates of 10 -4 - 10 -1 /s) and temperatures (1100 C - 1350 C - 1550 C et 1700 C). Experimental results (geometry, yield stress and microstructure) allowed us to define a hyperbolic sine creep law and a Drucker-Prager criterion with associated plasticity, in order to model grain boundaries fragmentation at the macroscopic scale. Finite Element Simulations of these tests and of more than 200 creep tests were used to assess the model response to a wide range of temperatures (1100 C - 1700 C) and strain rates (10 -9 /s - 10 -1 /s). Finally, a constitutive law called L3F was developed for UO 2 by adding to the previous model irradiation creep and tensile macroscopic cracking. The L3F law was then introduced in the 1.5D scheme of the fuel performance code ALCYONE-RIA to simulate the REP-Na tests performed in the experimental reactor CABRI. Simulation results are in good agreement with post tests examinations. (author) [fr

  20. Large-scale production of UO2 kernels by sol–gel process at INET

    Hao, Shaochang; Ma, Jingtao; Zhao, Xingyu; Wang, Yang; Zhou, Xiangwen; Deng, Changsheng


    In order to supply elements (300,000 elements per year) for the Chinese pebble bed modular high temperature gas cooled reactor (HTR-PM), it is necessary to scale up the production of UO 2 kernels to 3–6 kgU per batch. The sol–gel process for preparation of UO 2 kernels have been improved and optimized at Institute of Nuclear and New Energy Technology (INET), Tsinghua University, PR China, and a whole set of facility was designed and constructed based on the process. This report briefly describes the main steps of the process, the key equipment and the production capacities of every step. Six batches of kernels for scale-up verification and four batches of kernels for fuel elements for in-pile irradiation tests have been successfully produced, respectively. The quality of the produced kernels meets the design requirements. The production capacity of the process reaches 3–6 kgU per batch

  1. Micrography of UO2 power of PPNY and France using JSM T-20

    Kasilani, N.S.; Hidayati; Hartati, P.


    As a quality control of processing to produce a fuel element using UO 2 powder, its necessary to be known the physical characteristic of the shape, size and surface condition of particle. This physical character influence the flow ability of powder particles and density of pellet. To create photomicrograph used a electron microscope, influenced by the condition of tools, specimen, and the skilled of the processing of pictures. The current of JSM T-20 is about 5 until 20 Kv, used 11 and 8 camera diaphragm with black and white films, specimen must be dried with electrical conductor property. The processing resulted an optimal photomicrograph. Micrograph of UO 2 powder of PPNY and France was investigated, yield a same grain form, surface structure are different, and range of size particle is 0.5 - 1.0 um. (author)

  2. Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

    Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.


    The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

  3. Nuclear fuels


    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  4. Electrochemical studies of the effect of H2 on UO2 dissolution

    King, F.; Shoesmith, D.W.


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

  5. UO2 fuel behaviour at rod burn-ups up to 105 MWd/kgHM. A review of 10 years of high burn-up examinations commissioned by AREVA NP

    Goll, W.; Hoffmann, P.B.; Hellwig, C.; Sauser, W.; Spino, J.; Walker, C.T.


    Irradiation experience gained on fuel rods with burn-ups greater than 60 MWd/kgHM irradiated in the Nuclear Power Plant Goesgen, Switzerland, is described. Emphasis is placed on the fuel behaviour, which has been analysed by hot cell examinations at the Institute for Transuranium Elements and the Paul-Scherrer-Institute. Above 60 MWd/kgHM, the so-called high burn-up structure (HBS) forms and the fission gas release increases with burn-up and rod power. Examinations performed in the outer region of the fuel revealed that most if not all of the fission gas created was retained in the HBS, even at 25% porosity. Furthermore, the HBS has a relatively low swelling rate, greatly increased plasticity, and its thermal conductivity is higher than expected from the porosity. The post-irradiation examinations showed that the HBS has no detrimental effects on the performance of stationary irradiated PWR fuel irradiated to the high burn-ups that can be achieved with 5 wt% U-235 enrichment. On the contrary, the HBS results in fuel performance that is generally better than it would have been if the HBS had not formed. (orig.)

  6. Simulation of LOF accidents with directly electrical heated UO2 pins

    Alexas, A.


    The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

  7. Eh and fission product solubilities: two factors in the leaching of UO2

    Ogard, A.E.; Duffy, C.J.


    Eh was found to have a large effect on the dissolution of UO 2 in water at pH 4. As was estimated from thermodynamic data, the solubility was found to decrease as the oxygen fugacity, and therefore the Eh of the water, was decreased. Some of the rare earths and other actinides such as europium, cerium, americium, and plutonium released during the leaching of a spent fuel element behaved differently. These elements were not affected to any large extent by the variation in Eh of these experiments. It has been postulated that these elements reached their solubility limits and precipitated as the spent fuel was leached. 2 figures, 2 tables

  8. Homogeneity Study of UO2 Pellet Density for Quality Control

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


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

  9. UO2: production based on two alternative lines

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


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

  10. 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)

    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.


    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)

  11. High Pressure Low Temperature X-Ray Diffraction Studies of UO2 and UN single crystals.

    Antonio, Daniel; Mast, Daniel; Lavina, Barbara; Gofryk, Krzysztof

    Uranium dioxide is the most commonly used nuclear fuel material in commercial reactors, while uranium nitride also has many thermal and physical properties that make it attractive for potential use in reactors. Both have a cubic fcc lattice structure at ambient conditions and transition to antiferromagnetic order at low temperature. UO2 is a Mott insulator that orders in a complex non-collinear 3k magnetic structure at about 30 K, while UN has appreciable conductivity and orders in a simpler 1k magnetic structure below 52 K. Both compounds are characterized by strong magneto-structural interactions, understanding of which is vital for modeling their thermo-physical properties. While UO2 and UN have been extensively studied at and above room temperature, little work has been done to directly study the structure of these materials at low temperatures where magnetic interactions are dominant. In the course of our systematic studies on magneto vibrational behavior of UO2 and UN, here we present our recent results of high pressure X-Ray Diffraction (up to 35 GPa) measured below the Neel temperature using synchrotron radiation. Work supported by the Department of Energy, Office of Basic Energy Sciences, Materials Sciences, and Engineering Division.

  12. Electronic structure analysis of UO2 by X-ray absorption spectroscopy

    Ozkendir, O.M.


    Full text: Due to the essential role of Actinides in nuclear science and technology, electronic and structural investigations of actinide compounds attract major interest in science. Electronic structure of actinide compounds have important properties due to narrow 5f states which play key role in bonding with anions. The properties of Uranium has been a subject of enduring interest due to its being a major importance as a nuclear fuel and is the highest numbered element which can be found naturally on earth. UO 2 forms as a secondary uranyl group occurred during metamictization of uranium oxide compounds [1].Uranium oxide thin films have been investigated by X-ray Absorption Fine Structure spectroscopy (XAFS) [2]. The full multiple scattering approach has been applied to the calculation of U L3 edge spectra of UO 2 . The calculations are based on different choices of one electron potentials according to Uranium coordinations by using the real space multiple scattering method FEFF 8.2 code [3,4]. U L3-edge absorption spectrum in UO 2 is compared with U L3-edges in USiO 4 and UTe which are chosen due to their different electronic and chemical structures.We have found prominent changes in the XANES spectra of Uranium oxide thin films due to valency properties. Such observed changes are explained by considering the structural, electronic and spectroscopic properties. (author)

  13. Optimization of process parameters in precipitation for consistent quality UO2 powder production

    Tiwari, S.K.; Reddy, A.L.V.; Venkataswamy, J.; Misra, M.; Setty, D.S.; Sheela, S.; Saibaba, N.


    Nuclear reactor grade natural uranium dioxide powder is being produced through precipitation route, which is further processed before converting into sintered pellets used in the fabrication of PHWR fuel assemblies of 220 and 540 MWe type reactors. The process of precipitating Uranyl Nitrate Pure Solution (UNPS) is an important step in the UO 2 powder production line, where in soluble uranium is transformed into solid form of Ammonium Uranate (AU), which in turn reflects and decides the powder characteristics. Precipitation of UNPS with vapour ammonia is being carried out in semi batch process and process parameters like ammonia flow rate, temperature, concentration of UNPS and free acidity of UNPS are very critical and decides the UO 2 powder quality. Variation in these critical parameters influences powder characteristics, which in turn influences the sinterability of UO 2 powder. In order to get consistent powder quality and sinterability the critical parameter like ammonia flow rate during precipitation is studied, optimized and validated. The critical process parameters are controlled through PLC based automated on-line data acquisition systems for achieving consistent powder quality with increased recovery and production. The present paper covers optimization of process parameters and powder characteristics. (author)

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

    Breitung, W.; Reil, K.O.


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

  15. Thermal performance prediction of UO2 pellet partly containing 9%w tungsten network



    Sintered UO 2 exhibits very stable in reactor core compared to UC, UN, U metal and its alloys. However, its thermal conductivity is very low (2.about.5 W/m K), that limits its performance. UO 2 pellet containing Tungsten network invented by Song improves considerably its conductivity. The paper reports an analysis of thermal performance for UO 2 pellet that contains partly or wholly with 9% b. of Tungsten. The tungsten network having a high melting point and excellent thermal conductivity is continuously formed around UO 2 grains. Since the presence of network decreases the amount of fissile material and the burn up of fissile material is higher in the near surface zone of pellet but high temperature zone that releases low conductivity fission gas to the gap located in inner part of pellet, the analysis has been done for different outer radial-portion of tungsten-free pellet. The analysis takes into account the correction factor for pellet conductivity related to both pore and temperature distribution and high burn up effect. The gap conductance has been considered invariable since decrease caused by wider gap size related to lower pellet expansion is compensated by increase caused by fewer of refractory fission gas released. The results (47 kw/m, 40% burnup) show temperature decrease in all of pellet position containing W network. Pellet containing 9%b. tungsten network lower consecutively its center line temperature from 1578 to 1406, 1292, 1231, 1192, 1111, and 1038 deg C for 0, 50, 67, 75, 80, 90, and 100 % portion of network. An 80 to 90 % portion of inner pellet containing tungsten network can be considered a best fuel design. This preliminary analysis is prospective and more realistic one is recommended. (author)

  16. Updated FY12 Ceramic Fuels Irradiation Test Plan

    Nelson, Andrew T.


    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  17. Highly durable, coking and sulfur tolerant, fuel-flexible protonic ceramic fuel cells.

    Duan, Chuancheng; Kee, Robert J; Zhu, Huayang; Karakaya, Canan; Chen, Yachao; Ricote, Sandrine; Jarry, Angelique; Crumlin, Ethan J; Hook, David; Braun, Robert; Sullivan, Neal P; O'Hayre, Ryan


    Protonic ceramic fuel cells, like their higher-temperature solid-oxide fuel cell counterparts, can directly use both hydrogen and hydrocarbon fuels to produce electricity at potentially more than 50 per cent efficiency 1,2 . Most previous direct-hydrocarbon fuel cell research has focused on solid-oxide fuel cells based on oxygen-ion-conducting electrolytes, but carbon deposition (coking) and sulfur poisoning typically occur when such fuel cells are directly operated on hydrocarbon- and/or sulfur-containing fuels, resulting in severe performance degradation over time 3-6 . Despite studies suggesting good performance and anti-coking resistance in hydrocarbon-fuelled protonic ceramic fuel cells 2,7,8 , there have been no systematic studies of long-term durability. Here we present results from long-term testing of protonic ceramic fuel cells using a total of 11 different fuels (hydrogen, methane, domestic natural gas (with and without hydrogen sulfide), propane, n-butane, i-butane, iso-octane, methanol, ethanol and ammonia) at temperatures between 500 and 600 degrees Celsius. Several cells have been tested for over 6,000 hours, and we demonstrate excellent performance and exceptional durability (less than 1.5 per cent degradation per 1,000 hours in most cases) across all fuels without any modifications in the cell composition or architecture. Large fluctuations in temperature are tolerated, and coking is not observed even after thousands of hours of continuous operation. Finally, sulfur, a notorious poison for both low-temperature and high-temperature fuel cells, does not seem to affect the performance of protonic ceramic fuel cells when supplied at levels consistent with commercial fuels. The fuel flexibility and long-term durability demonstrated by the protonic ceramic fuel cell devices highlight the promise of this technology and its potential for commercial application.

  18. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong


    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  19. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

  20. Fracture properties of ThO2-UO2 pellets by Hertzian indentation technique

    Kutty, T.R.G.; Rath, B.N.; Balakrishnan, K.S.


    Fracture toughness (K Ic ) and fracture surface energy (γ s ) of ThO 2 -UO 2 pellets with varying UO 2 contents were measured using Hertzian indentation technique. The knowledge of fracture toughness (K Ic ) and fracture surface energy values are important for fuel designers since these values are used in fuel modeling. Cracks in nuclear fuel act as a path for fission gas release and enhances fuel cladding mechanical interaction. Microstructural features like grain size and presence of second phase play a significant role in controlling the fracture behavior. Since the fracture properties of nuclear materials are of primary design consideration, it is important that these properties should be evaluated with good precision. There have been several attempts to use Hertzian indentation for evaluating the fracture toughness of brittle materials. The main principle of this method depends on the interaction of the elastic stress field with a pre-existing surface flaw of the sample. One significant advantage of Hertzian indentation over that of Vickers is that the substrate's deformation is entirely elastic until fracture occurs. This avoids the complications arising from the ill-defined residual stress that is normally associated with indentations brought about by pointed indenters like that of Vickers. The material properties that may be determined by this test include (a) fracture toughness and fracture surface energy of the near surface material, (b) the densities and sizes of surface cracks, and (c) residual stresses in the near surface material. This paper deals with experimental procedure for the evaluation of fracture properties of ThO 2 -UO 2 of varying U content and results thus obtained are also presented. The K Ic values thus obtained are explained in terms of their microstructures and the U content. (author)

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

    Damunir; Sri Rinanti Susilowati; Ariyani Kusuma Dewi


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

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

    Mendes, E.


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

  3. Creep behavior of UO2 above 20000C

    Slagle, O.D.


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

  4. Characterization of Compaction Process on UO2 Powder Pelletisation

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


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

  5. Electrochemical Reduction of solid UO2 in Molten Fluoride Salts

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


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

  6. Interaction and penetration of heated UO2 with limestone concrete

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


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

  7. Method for fluoride ion depletion of UO2 powders

    Beutner, R.; Ploeger, F.


    The method described consists in removing the hydrogen still present from the reduction during the preparation of UO 2 as completely as possible and in performing a pyrohydrolysis at temperatures above 650 0 C for at least 45 minutes. The removal of fluorine is necessary in order to avoid cladding tube damaging. (UA) [de

  8. Thermodynamic state, specific heat, and enthalpy function of saturated UO2 vapor between 3,000 K and 5,000 K

    Karow, H.U.


    The properties have been determined by means of statistical mechanics. The discussion of the thermodynamic state includes the evaluation of the plasma state and its contribution to the caloric variables-of-state of saturated oxide fuel vapor. Because of the extremely high ion and electron density due to thermal ionization, the ionized component of the fuel vapor does no more represent a perfect kinetic plasma. At temperatures around 5,000 K, UO 2 vapor reaches the collective plasma state and becomes increasingly 'metallic'. - Moreover, the nonuniform molecular equilibrium composition of UO 2 vapor has been taken into account in calculating its caloric functions-of-state. The contribution to specific heat and enthalpy of thermally excited electronic states of the vapor molecules has been derived by means of a Rydberg orbital model of the UO 2 molecule. The resulting enthalpy functions and specific heats for saturated UO 2 vapor of equilibrium composition and that for pure UO 2 gas are compared with the enthalpy and specific heat data of gaseous UO 2 at lower temperatures known from literature. (orig./HP) [de

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

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


    Silicon carbide (SiC) is suggested as an oxygen getter in UO 2 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 UO 2 fuel kernels. Even though the presence of UC in either argon (Ar) or Ar–4%H 2 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 SiO 2 oxide layer on SiC grains to produce volatile SiO and free carbon that subsequently reacted with UO 2 to form UC. The second process was direct UO 2 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%H 2 , 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

  10. Phonon density of states and anharmonicity of UO2

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


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

  11. Glass/Ceramic Composites for Sealing Solid Oxide Fuel Cells

    Bansal, Narottam P.; Choi, Sung R.


    A family of glass/ceramic composite materials has been investigated for use as sealants in planar solid oxide fuel cells. These materials are modified versions of a barium calcium aluminosilicate glass developed previously for the same purpose. The composition of the glass in mole percentages is 35BaO + 15CaO + 5Al2O3 + 10B2O3 + 35SiO2. The glass seal was found to be susceptible to cracking during thermal cycling of the fuel cells. The goal in formulating the glass/ ceramic composite materials was to (1) retain the physical and chemical advantages that led to the prior selection of the barium calcium aluminosilicate glass as the sealant while (2) increasing strength and fracture toughness so as to reduce the tendency toward cracking. Each of the composite formulations consists of the glass plus either of two ceramic reinforcements in a proportion between 0 and 30 mole percent. One of the ceramic reinforcements consists of alumina platelets; the other one consists of particles of yttria-stabilized zirconia wherein the yttria content is 3 mole percent (3YSZ). In preparation for experiments, panels of the glass/ceramic composites were hot-pressed and machined into test bars.

  12. Fission gas behaviour in UO2 under steady state and transient conditions

    Zimmermann, H.


    Fission gas behaviour in UO 2 is determined by the limited capacity of the fuel to retain fission gas. This capacity depends primarily on temperature, but also on fission rate, pressure loading, and fuel microstructure. Under steady state irradiation conditions fission gas behaviour can be described qualitatively as follows: At the beginning of the irradiation most of the fission gas remains in the grains in irradiation-induced solution. With increasing gas content in the grains the gas transport to the grain boundaries increases, too. The fission gas release from the grain boundaries occurs primarily by interlinkage of inter-granular bubbles. The fission gas release without noticeable fuel swelling during the short-term heating in the LOCA tests and the powdering of the high burnup UO 2 in the annealing tests can only be accounted for by formation of inter-granular separations, which are caused by the fission gas accumulated in the grain boundaries. Besides this short-term effect there are diffusion-controlled long-term effects, such as growth and coalescence of bubbles and formation of inter-connected porosity, which result in time-dependent fission gas release and fuel swelling

  13. The defect chemistry of UO2 ± x from atomistic simulations

    Cooper, M. W. D.; Murphy, S. T.; Andersson, D. A.


    Control of the defect chemistry in UO2 ± x is important for manipulating nuclear fuel properties and fuel performance. For example, the uranium vacancy concentration is critical for fission gas release and sintering, while all oxygen and uranium defects are known to strongly influence thermal conductivity. Here the point defect concentrations in thermal equilibrium are predicted using defect energies from density functional theory (DFT) and vibrational entropies calculated using empirical potentials. Electrons and holes have been treated in a similar fashion to other charged defects allowing for structural relaxation around the localized electronic defects. Predictions are made for the defect concentrations and non-stoichiometry of UO2 ± x as a function of oxygen partial pressure and temperature. If vibrational entropy is omitted, oxygen interstitials are predicted to be the dominant mechanism of excess oxygen accommodation over only a small temperature range (1265 K-1350 K), in contrast to experimental observation. Conversely, if vibrational entropy is included oxygen interstitials dominate from 1165 K to 1680 K (Busker potential) or from 1275 K to 1630 K (CRG potential). Below these temperature ranges, excess oxygen is predicted to be accommodated by uranium vacancies, while above them the system is hypo-stoichiometric with oxygen deficiency accommodated by oxygen vacancies. Our results are discussed in the context of oxygen clustering, formation of U4O9, and issues for fuel behavior. In particular, the variation of the uranium vacancy concentrations as a function of temperature and oxygen partial pressure will underpin future studies into fission gas diffusivity and broaden the understanding of UO2 ± x sintering.

  14. Effect of continuous change of sintering atmosphere on the grain growth of Cr-doped UO2 pellets

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


    Cr-doped UO 2 pellet is one of the promising candidates for the high burn-up fuel in commercial LWRs. Major nuclear fuel vendors of such as AREVA or Westinghouse initiated the development of Cr-doped or Cr-containing additives doped UO 2 pellets since at the mid of 90's. Now, qualification programs are on-going to provide these pellets commercially. The main characteristics of the Cr-doped pellets are large-grain and visco-plasticity. Large grain pellet can reduce the corrosive fission gas release at high burn up. Viscoplastic soft pellets can lower the pressure to a cladding caused by a thermal expansion of a pellet at an elevated temperature during transient operations. Those advantages can provide room for additional power uprates and high burnup limits. Especially, PCI resistance improvement can be achieved by enlarging the pellet grain size and enhancing the fuel deformation at an elevated temperature. In this paper, to study the effect of oxygen partial pressure on grain growth in Cr-doped UO 2 pellets, Cr- doped UO 2 samples have been sintered with and without a step-wise change of sintering atmospheres. An introduction of a step-wise variation of oxygen partial pressure during the sintering enhances the grain growth of UO 2 pellets greatly. This step-wise sintering effect has been explained in terms of a continuous increase of Cr concentration along the grain boundary. The observed grain growth behavior under step-wisely changed sintering atmospheres demonstrates the possibility of reducing the amount of Cr 2 O 3 to minimum via control of oxygen partial pressure while keeping the large grain size

  15. Comparative study of the different industrial manufacturing routes for UO2 pellet specifications through the wet process

    Palheiros, Franklin; Gonzaga, Reinaldo; Soares, Alexandre


    In the fuel cycle, converting UF 6 to UO 2 powder is an intermediate step for fabrication of pellets for fuel assemblies to be used in nuclear power plants. The basic proposal common to the different powder fabrication processes is to provide raw material capable of being processed into the form of pellets. The wet processes is the most often used industrially and are divided in two categories: the ADU (Ammonium Diuranate) and AUC (Ammonium Uranyl Carbonate) processes, whose names originate in the precipitate obtained in aqueous solution during the intermediate steps of UO 2 powder fabrication. It has known that the powder characteristics have a considerable influence in the UO 2 pellet manufacturing and quality characteristics. INB has used the AUC process to produce UO 2 pellets and supply fuel to Angra 1 and 2 Nuclear Power Plants. Despite of this process is characterized by the precipitation of a different intermediate precipitate compared to the ADU route (i.e., (NH 4 ) 4 UO 2 (CO 3 ) 3 , in the AUC process, and (NH 4 ) 2 U 2 O 7 in ADU process) leading to some slight differences in the final pellet microstructure, it has been considered that the models that predict the pellet behavior during irradiation in a nuclear reactor are basically the same compared to those used to predict the pellets form the ADU process. In order to evaluate how different the pellets originated from these two industrial routes are, this paper aims to compare the INB production historical data (Angra 1, Cycles 14 and 15) with the key parameters of a common product specification from the ADU process. (author)

  16. Encapsulation of spent nuclear fuel in ceramic materials

    Forberg, S.; Westermark, T.


    The international situation with regard to deposition of spent nuclear fuel is surveyed, with emphasis on encapsulation in ceramic materials. The feasibility and advantages of ceramic containers, thermodynamic stable in groundwater, are discussed as well as the possibility to ensure that stability for longevity by engineered measures. The design prerequisite are summarized and suggestions are made for a conceptual design, comprising rutile containers with stacks of coiled fuel pins. A novel technique is suggested for the homogeneous sealing of rutile containers at low temperatures. acceptable also for the fuel pin package. Key points are given for research, demonstration and verifications of the design foundations and for future improvements. Of which a few ideas are exemplified. (author)

  17. Fabrication and characterization of fully ceramic microencapsulated fuels

    Terrani, K.A., E-mail: [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)


    The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.

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

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


    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.

  19. Effect of helium pressure on the response of unirradiated UO2 subjected to thermal transients

    Fenske, G.R.; Chapello, P.M.; Emerson, J.E.; Poeppel, R.B.


    The effect of helium pressure on the transient response of unirradiated depleted UO 2 subjected to simulated hypothetical loss-of-flow accidents in a gas-cooled fast reactor was examined by use of the direct electrical heating technique. Transient tests were performed at pressures ranging from 7 to 10 X 10 5 Pa(7 to 10 atm) to 7 to 8 MPa (70 to 80 atm) on radially restrained and unrestrained fuel segments. The average heating rates ranged from about17 to 240 J/g x s. The results indicate that while the mechanical integrity of the fuel segment was independent of the test pressure, the rapid ejection of molten fuel from pellet interfaces of unrestrained fuel, observed at the lower pressures, was delayed or suppressed at the higher pressures

  20. Anode-Engineered Protonic Ceramic Fuel Cell with Excellent Performance and Fuel Compatibility

    Hua, B.; Yan, N.; Li, M.; Sun, Y.-F.; Zhang, Y.-Q.; Li, J.; Etsell, T.; Sarkar, P.; Luo, J.L.


    Directly utilizing hydrocarbon fuels, particularly methane, is advantageous yet challenging in high-performance protonic ceramic fuel cells. In this work, this technological hurdle is well addressed by selective deposition of secondary electrocatalysts within the porous Ni-cermet anode. This novel

  1. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    Tukiran Surbakti


    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  2. Dissolution characteristics of mixed UO2 powders in J-13 water under saturated conditions

    Veleckis, E.; Hoh, J.C.


    The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated UO 2 powder mixture (14.3 wt % enrichment in 235 U) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of UO 2 matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results. 10 refs., 5 figs., 7 tabs

  3. (Alpha, gamma) irradiation effect on the alteration of high-level radioactive wastes matrices (UO2, hollandite, glass SON68)

    Suzuki, T.


    The aim of this work is to determine the effect of irradiation on the alteration of high level nuclear waste forms matrices. The matrices investigated are UO 2 to simulate the spent fuel, the hollandite for the specific conditioning of Cs, and the inactive glass SON68 representing the nuclear glass R7T7) The alpha irradiation experiments on UO 2 colloids in aqueous carbonate media have enabled to distinguish between the oxidation of UO 2 matrix as initial and dissolution as subsequent step. The simultaneous presence of carbonate and H 2 O 2 (product resulting from water radiolysis) increased the dissolution rate of UO 2 to its maximum value governed by the oxidation rate. ii) The study of hollandite alteration under gamma irradiation confirmed the good retention capacity for Cs and Ba. Gamma irradiation had brought only a little influence on releasing of Cs and Ba in solution. Electronic irradiation had conducted to the amorphization of the hollandite only for a dose 1000 times higher than the auto-induced dose of Ba over millions of years. iii) The experiences of glass irradiation under alpha beam and of helium implantation in the glass SON68 were analyzed by positon annihilation spectroscopy. No effect has been observed on the solid surface for an irradiation dose equal to 1000 years of storage. (author)

  4. Preparation of Fluidization Feed of UO2 Pellets by Oxidation

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


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

  5. Acoustic emission from thermal-gradient cracks in UO2

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


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

  6. Densification behaviour of UO2 in six different atmospheres

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


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

  7. Binding energy and formation heat of UO2

    Almeida, M.R. de; Veado, J.T.; Siqueira, M.L. de

    The Born-Haber cycle is utilized for the calculation of the heat of formation of UO 2 , on the assumption that the binding energy is predominantly ionic in character. The ionization potentials of U and the repulsion energy are two critical values that influence calculations. Calculations of the ionization potentials with non-relativistic Hartree-Fock-Gaspar-Kohn-Sham approximation are presented [pt

  8. Pressure-induced weak ferromagnetism in uranium dioxide, UO2

    Sakai, H; Kato, H; Tokunaga, Y; Kambe, S; Walstedt, R E; Nakamura, A; Tateiwa, N; Kobayashi, T C


    The dc magnetization of insulating UO 2 under high pressure up to ∼1 GPa has been measured using a piston-cylinder cell. Pressure-induced weak ferromagnetism appeared at low pressure (∼0.2 GPa). Both the remanent magnetization and the coercive force increase as pressure increases. This weak ferromagnetism may come from spin canting or from uncompensated moments around grain boundaries

  9. Study of physical properties of UO2 quality improvement result

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


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

  10. Nuclear fuels

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.


    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  11. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    Podest, M.; Klima, J.; Stecher, P.; Stecherova, E.


    UO 2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO 2 . The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO 2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

  12. 3D Finite Elements Modelling for Design and Performance Analysis of UO2 Pellets

    Demarco, G.L.; Marino, A.C.; Demarco, G.L.; Marino, A.C.


    The geometry of a fuel pellet is a compromise among the intention to maximize UO 2 content and minimize the temperature gradient taking into account the thermomechanical behaviour, the economy, and the safety of the fuel management during and after irradiation. Dishing, shoulders, chamfers, and/or a central hole on a cylinder with an improved l/d relation (length of the pellet/diameter) are introduced in order to optimize the shape of the pellet. The Me Com tools coupled with the BaCo code constitutes a complete system for the 3D analysis of the stress strain state of the pellet under irradiation. CANDU and PHWR MOX fuel will be used to illustrate the excellent qualitative agreement between experimental

  13. Method of production of granulates of ceramic nuclear fuels

    Wilkinson, W.L.


    To obtain a classified granulate of ceramic nuclear fuels with narrow grain size spectrum, the nuclear fuel powder is made into a slurry in a non-aqueous solvent with a water content as low as possible (e.g. chlorated hydrocarbon), a binder added to it, and spray-dried. The dry granulate desired is already obtained by this working stage. Polybutyl methacrylate in dibutylphthalate is proposed as binder. An example in which uranium dioxide powder is slurried in trichloro-ethylene is described in detail. (UWI/LH) [de

  14. Simulation of High Burnup Structure in UO2 Using Potts Model

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho


    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

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

    Wang Bangwu; Chen Ying


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

  16. Development of ceramic roller bush for diesel fuel injection pump; Nenryo funsha pump yo ceramics sei roller bush no kaihatsu

    Noda, K; Kamiya, S; Fujimura, M; Tsuzuki, M [Toyota Motor Corp., Aichi (Japan); Taniguchi, K [Denso Corp., Aichi (Japan)


    Silicon nitride ceramics have been applied to roller bush for diesel fuel injection pump in order to improve the seizure resistance. It was found that ceramic roller bush made it possible to improve the seizure load by more than three times as compared to conventional metal roller bush when the kerosene was used as lubricant The ceramic roller bush proved to be durable under engine operating conditions. 6 refs., 13 figs., 1 tab.

  17. Ceramic waste forms for fuel-containing masses at Chernobyl

    Oversby, V.M.


    The fuel materials originally in the core of the Chernobyl Unit 4 reactor are now present within the Ukrytie in three major forms: (1) very fine particles of fuel dispersed as dust (about 10 tonnes), (2) fragments of the destroyed core, and (3) lavas containing fuel, cladding, and other materials. All of these materials will need to be immobilized into waste forms suitable for final disposal. We propose a ceramic waste form system that could accommodate all three waste types with a single set of processing equipment. The waste form would include the mineral zirconolite for immobilization of actinide materials (including uranium), perovskite, nepheline, spinel, and other phases as dictated by the chemistry of the lava masses. Waste loadings as high as 50% U can be achieved if pyrochlore, a close relative of zirconolite, is used as the U host. The ceramic immobilization could be achieved with low additions of inert chemicals to minimize the final disposal volume while ensuring a durable product. The sequence of processing would be to collect and immobilize the fuel dust first. This material will require minimal preprocessing and will provide experience in the handling of the fuel materials. Core fragments would be processed next, using a cryogenic crushing stage to reduce the size prior to adding ceramic additives. The lavas would be processed last, which is compatible with the likely sequence of availability of materials and with the complexity of the operations. The lavas will require more adjustment of chemical additive composition than the other streams to ensure that the desired phases are produced in the waste form

  18. Process and equipment development for the preparation of UO2 microspheres using trichloroethylene as gelation medium (Paper No. AL-23)

    Suryanarayana, S.; Kumar, N.; Bamankar, Y.R.; Vaidya, V.N.; Sood, D.D.


    Uranium dioxide microspheres have been prepared by internal gelation process, one of the sol-gel routes for fuel fabrication. The process flow sheet for internal gelation has been modified by employing trichloroethylene(TCE) as an alternate gelation medium. Based on the modified flow sheet, a 5Kg/day assembly for the production of UO 2 microspheres has been developed and installed. (author). 1 fig

  19. Modelling of 28-element UO2 flux-map critical experiments in ZED-2 using WIMS9A/PANTHER

    Sissaoui, M.T.; Kozier, K.S.; Labrie, J.P.


    The accuracy of WIMS9A/PANTHER in modelling D 2 O-moderated, and H 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters has been demonstrated using 28-element UO 2 flux-map critical experiments in the ZED-2 facility. Presented here are the predicted k eff values, coolant void reactivity biases, and the radial and axial flux shapes.

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

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


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

  1. Thermal conductivity analysis of SiC ceramics and fully ceramic microencapsulated fuel composites

    Lee, Hyeon-Geun; Kim, Daejong; Lee, Seung Jae; Park, Ji Yeon; Kim, Weon-Ju


    Highlights: • Thermal conductivity of SiC ceramics and FCM pellets was measured and discussed. • Thermal conductivity of FCM pellets was analyzed by the Maxwell-Eucken equation. • Effective thermal conductivity of TRISO particles applied in this study was assumed. - Abstract: The thermal conductivity of SiC ceramics and FCM fuel composites, consisting of a SiC matrix and TRISO coated particles, was measured and analyzed. SiC ceramics and FCM pellets were fabricated by hot press sintering with Al_2O_3 and Y_2O_3 sintering additives. Several factors that influence thermal conductivity, specifically the content of sintering additives for SiC ceramics and the volume fraction of TRISO particles and the matrix thermal conductivity of FCM pellets, were investigated. The thermal conductivity values of samples were analyzed on the basis of their microstructure and the arrangement of TRISO particles. The thermal conductivity of the FCM pellets was compared to that predicted by the Maxwell-Eucken equation and the thermal conductivity of TRISO coated particles was calculated. The thermal conductivity of FCM pellets in various sintering conditions was in close agreement to that predicted by the Maxwell-Eucken equation with the fitted thermal conductivity value of TRISO particles.

  2. Application of fully ceramic microencapsulated fuels in light water reactors

    Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)


    This study performs 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 (PWR). Using uranium-based fuel and Pu/Np-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 fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} 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 Pu/Np-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, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  3. REDSHANK I and GREENSHANK I (comprehensive point reactivity programmes for liquid moderated UO2 lattices)

    Alpiar, R.A.


    A recently issued programme (SANDPIPER I) enables few group diffusion parameters and reactivities to be derived for liquid moderated UO 2 lattices. The present programmes investigate the life history of such lattices. Burn up equations recalculate the fuel isotopic composition, in a series of steps. At each step, new few group constants and reactivity are recalculated for the new fuel composition. In addition, at each step, the control required to keep the reactivity of the reactor within a given deadband is recalculated. This control is effected by control rod withdrawal in Redshank, and by heavy water spectrum shift in Greenshank. The programme continues until the reactivity of the uncontrolled reactor falls below the deadband. (author)

  4. Study of secular equilibrium reinstatement on UO2 pellets manufactured by AUC route

    Carnaval, João Paulo R.; Beltran, Dalton J.M.C.; Oliveira, Carlos A.


    The fuel assemblies manufactured by INB for Angra-1 power plant has axial blanket fuel rods which must be inspected due the columns formed by different enrichment pellets. The equipment used for inspection is built with a group of BGO scintillators detectors which measurement principle is based on the absorption of gamma rays emitted from Uranium decay. The commercial grade UF 6 used by INB is stored into cylinders type 30B. The uranium inside these cylinders is in secular equilibrium before the processing. It has been found that the AUC route causes the loss of that equilibrium because the UF 6 is volatilized from the cylinder and the uranium daughters remain in the container. As AUC is converted to powder and pellets, the secular equilibrium is restored through time. The purpose of this work is to present a study of the secular equilibrium reinstatement on UO 2 pellets manufactured by AUC route before being inspected on Rod Scanner. (author)

  5. Thermal conductivity and thermal diffusivity of solid UO2

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


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

  6. Changes in UO2 powder properties during processing via BNFL's binderless route

    Bromely, A.P.; Logsdon, R.; Roberts, V.A.


    The Short Binderless Route (SBR) has been developed for Mixed Oxide fuel production in BNFL's MOX Demonstration Facility (MDF) and the Sellafield MOX Plant (SMP). It is a compact process which enables good homogenisation of the Pu/U mixture and production of free flowing press feed materials. The equipment used to achieve this consists of an attritor mill to provide homogenization and a spheroidiser to provide press feed granules. As for other powder processes, the physical properties of the UO 2 powder can affect the different process stages and consequently a study of some of these effects has been carried out. The aim of the work were to gain a better understanding of the process, to consequently optimize press feed material quality and to also maintain powder hold-up levels in the equipment at a minimum. The paper considers the effects of milling processes on powder morphology and powder surface effects, on the granulation process and also on powder and granule bulk properties such as pour, tap and compaction densities. Results are discussed in terms of powder properties such as powder cohesivity, morphology and particle size. UO 2 powder derived from both the Integrated Dry Route (IDR) and the Ammonium Di-Uranate (ADU) Route are considered. Small (1 kg) scale work has been carried out which has been confirmed by larger (25 kg) scale trials. The work shows that IDR powder with differing morphologies and ADU powder can be successfully processed via the SBR route. (author). 4 figs, 4 tabs

  7. The anodic dissolution of SIMFUEL (UO2) in slightly alkaline sodium carbonate/bicarbonate solutions

    Keech, P.G.; Goldik, J.S.; Qin, Z.; Shoesmith, D.W.


    The corrosion of nuclear fuel under waste disposal conditions is likely to be influenced by the bicarbonate/carbonate content of the groundwater since it increases the solubility of the U VI corrosion product, [UO 2 ] 2+ . As one of the half reactions involved in the corrosion process, the anodic dissolution of SIMFUEL (UO 2 ) has been studied in bicarbonate/carbonate solutions (pH 9.8) using voltammetric and potentiostatic techniques and electrochemical impedance spectroscopy. The reaction proceeds by two consecutive one electron transfer reactions (U IV → U V → U VI ). At low potentials (≤250 mV (vs. SCE) the rate of the first electron transfer reaction is rate determining irrespective of the total carbonate concentration. At potentials >250 mV (vs. SCE) the formation of a U VI O 2 CO 3 surface layer begins to inhibit the dissolution rate and the current becomes independent of potential indicating rate control by the chemical dissolution of this layer.

  8. Kinetic Monte Carlo Potts Model for Simulating a High Burnup Structure in UO2

    Oh, Jae-Yong; Koo, Yang-Hyun; Lee, Byung-Ho


    A Potts model, based on the kinetic Monte Carlo method, was originally developed for magnetic domain evolutions, but it was also proposed as a model for a grain growth in polycrystals due to similarities between Potts domain structures and grain structures. It has modeled various microstructural phenomena such as grain growths, a recrystallization, a sintering, and so on. A high burnup structure (HBS) is observed in the periphery of a high burnup UO 2 fuel. Although its formation mechanism is not clearly understood yet, its characteristics are well recognized: The HBS microstructure consists of very small grains and large bubbles instead of original as-sintered grains. A threshold burnup for the HBS is observed at a local burnup 60-80 Gwd/tM, and the threshold temperature is 1000-1200 .deg. C. Concerning a energy stability, the HBS can be created if the system energy of the HBS is lower than that of the original structure in an irradiated UO 2 . In this paper, a Potts model was implemented for simulating the HBS by calculating system energies, and the simulation results were compared with the HBS characteristics mentioned above

  9. Critical experiments with 4.31 wt % 235U-enriched UO2 rods in highly borated water lattices

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.


    A series of critical experiments were performed with 4.31 wt % 235 U enriched UO 2 fuel rods immersed in water containing various concentrations of boron ranging up to 2.55 g/l. The boron was added in the form of boric acid (H 3 BO 3 ). Critical experimental data were obtained for two different lattice pitches wherein the water-to-uranium oxide volume ratios were 1.59 and 1.09. The experiments provide benchmarks on heavily borated systems for use in validating calculational techniques employed in analyzing fuel shipping casks and spent fuel storage systems that may utilize boron for criticality control

  10. Modelling of the UO2 dissolution mechanisms in synthetic groundwater solutions. Dissolution experiments carried out under oxic conditions

    Cera, E.; Grive, M.; Bruno, J.; Ollila, K.


    The analytical data generated during the last three years within the 4th framework program of the European Community at VTT Chemical Technology concerning UO 2 dissolution under oxidising conditions have been modelled in the present work. The modelling work has been addressed to perform a kinetic study of the dissolution data generated by Ollila (1999) under oxidising conditions by using unirradiated uranium dioxide as solid sample. The average of the normalised UO 2 dissolution rates determined by using the initial dissolution data generated in all the experimental tests is (6.06 ± 3.64)* 10 -7 mol m -2 d -1 . This dissolution rate agrees with most of the dissolution rates reported in the literature under similar experimental conditions. The results obtained in this modelling exercise show that the same bicarbonate promoted oxidative dissolution processes operate for uranium dioxide, as a chemical analogue of the spent fuel matrix, independently of the composition of the aqueous solution used. (orig.)

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

    Brandau, T.; Brandau, E.


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

  12. A molecular dynamics study of solid and liquid UO2

    Sindzingre, P.; Gillan, M.J.


    We present an extensive series of molecular dynamics simulations of UO 2 in the solid and liquid states, in which we calculate the ionic diffusion coefficients and some of the important thermodynamic quantities. The simulations are based on a rigid-ion model derived from the new shell model potentials of Jackson and co-workers and make use of recently developed constant-pressure and constant-temperature techniques. The simulations confirm that UO 2 is an oxygen superionic conductor, as suggested by recent neutron scattering experiments. The temperature of the diffuse transition to the superionic regime is in satisfactory agreement with experiment, as is the melting point of the model system. The thermal expansion coefficient, specific heat and bulk modulus for the solid agree well with experiment below about 2500 K but are less satisfactory near the melting point; we suggest that the differences may be due to the effect of electronic excitations. The volume increase on melting and thermodynamic quantities of the liquid are sensitive to details of the inter-ionic potentials and are in only fair agreement with experiment. (author)

  13. Irradiation effects in UO2 and CeO2

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


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

  14. The effect of hydrogen and gamma radiation on the oxidation of UO2 in 0.1 mol*(dm)-3 NaCl solution

    King, F.; Quinn, M.J.; Miller, N.H.


    High partial pressures of H 2 may develop in an underground nuclear fuel waste disposal vault as a result of radiolysis of groundwater or corrosion of steel container components. The presence of H 2 could suppress the oxidation and subsequent dissolution of used fuel by creating reducing conditions near the fuel surface. A series of experiments has been performed to determine the extent of oxidation of UO 2 due to γ-radiolysis in the presence of H 2 . A H 2 partial pressure of 5 MPa was used to simulate the maximum possible pressure of H 2 in a disposal vault located at a depth of 500 m. Experiments were also performed with an Ar overpressure for comparison. Deaerated 0.1 mol·(dm) -3 NaCl was used to simulate the groundwater. The extent of oxidation was determined by monitoring the corrosion potential of UO 2 electrodes, by cathodically stripping the oxidized layer from the electrode at the end of the test, and by determining the ratio of U(VI) to U(IV) species on the surface of a UO 2 disc exposed to the same solution by X-ray photoelectron spectroscopy. The presence of H 2 is found to have two effects on the oxidation of UO 2 in the presence of y-radiation. Not only does H 2 prevent oxidation of the UO 2 by radiolytic oxidants but it also produces more reducing conditions than those observed with either H 2 or Ar atmospheres in the absence of irradiation. It is suggested that radiolytically produced reductants participate in homogeneous reactions in solution with radiolytic oxidants and in heterogeneous reactions on the UO 2 surface, most likely at reactive grain-boundary sites

  15. Nitrate conversion and supercritical fluid extraction of UO2-CeO2 solid solution prepared by an electrolytic reduction-coprecipitation method

    Zhu, L.Y.; Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J.


    A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N 2 O 4 into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO 2 -CeO 2 solid solution was prepared as a surrogate for a UO 2 -PuO 2 solid solution, and the recovery of U and Ce from the UO 2 -CeO 2 solid solution with liquid N 2 O 4 and supercritical CO 2 containing tri-n-butyl phosphate (TBP) was investigated. The UO 2 -CeO 2 solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N 2 O 4 . The XRD pattern of the nitrates was similar to that of UO 2 (NO 3 ) 2 . 3H 2 O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO 2 containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

  16. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.


    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  17. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)


    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  18. Oxidation of UO2 at 400 to 1000 degrees C in air and its relevance to fission product release

    McCracken, D.R.


    Currently there is great interest in the behaviour of UO 2 under oxidizing conditions because irradiated uranium dioxide fuel can conceivably be exposed to a hot oxidizing atmosphere as a result of accidents. The temperature range covered in this paper is 400 to 1000 degrees C. At these high temperatures, UO 2 in air can oxidize rapidly to U 3 O 8 via U 3 O 7 and/or U 4 O 9 . The accompanying volume increase and corresponding stresses lead to fragmentation of the fuel pellets. The purpose of this work was to investigate the dependence of UO 2 oxidation on temperature, rate of air supply and residence time at temperature; to determine the rate controlling steps and rate of oxygen penetration; and to characterize the oxidation products and size of fragments. In addition, detailed metallography was related to X-ray diffraction studies of the oxidized UO 2 to facilitate future study of irradiated fuel, which is easier to do by metallography in hot-cells than by X-ray diffraction. Samples were heated in argon, then once at temperature they were exposed to air at a controlled flow-rate. Studies of the oxidation of unirradiated UO 2 pellets in air show two distinct types of oxidation with a change in mechanism at 600-700 degrees C. At temperatures ≤ 600 degrees C fragmentation accompanies the formation of U 3 O 8 while at T ≥ 800 degrees C, rapid grain growth occurs. In the first temperature region, volatile fission product releases are small, while in the second region, 100% release can be correlated with U 3 O 8 formation. In the first region, only the grain boundary inventory is released while in the other, 100% of the Xe, Kr, Ru, Sb, Cs and I are released. It appears that, within the error of present measurements, burnup does not affect rates of fission product release and oxidation in air at 400 to 1000 degrees C, so that oxidation rate data gathered using unirradiated pellets can be applied to irradiated fuel. 33 refs

  19. Experimental investigations of long-term interactions of molten UO2 with MgO and concrete at Argonne National Laboratory

    Stein, R.P.; Farhadieh, R.; Pedersen, D.R.; Gunther, W.H.; Purviance, R.T.


    Experimental work at Argonne is being performed to investigate the long-term molten-core-debris retention capability of the ex-vessel cavity following a postulated meltdown accident. The eventual objective of the work is to determine if normal structural material (concrete) or a specifically selected sacrificial material (MgO) located in the ex-vessel cavity region can effectively contain molten core debris. The materials under investigation at ANL are various types of concrete (limestone, basalt and magnetite) and commercially-available MgO brick. Results are presented of the status of real material experimental investigation at ANL into (1) molten UO 2 pool heat transfer, (2) long-term molten UO 2 penetration into concrete and (3) long-term molten UO 2 penetration into refractory substrates. The decay heating in the fuel has been simulated by direct electrical heating permitting the study of the long-term interaction

  20. Experimental investigations of long-term interactions of molten UO2 with MgO and concrete at Argonne National Laboratory

    Stein, R.P.; Farhadieh, R.; Pedersen, D.R.; Gunther, W.H.; Purviance, R.T.


    Experimental work at Argonne is being performed to investigate the long-term molten core debris retention capability of the ex-vessel cavity following a postulated meltdown accident. The eventual objective of the work is to determine if normal structural material (concrete) or a specifically selected sacrificial material (MgO) located in the ex-vessel cavity region can effectively contain molten core debris. The materials under investigation at ANL are various types of concrete (limestone, basalt and magnetite) and commercially-available MgO brick. Results are presented of the status of real material experimental investigation at ANL into 1) molten UO 2 pool heat transfer, 2) long-term molten UO 2 penetration into concrete and 3) long-term molten UO 2 penetration into refractory substrates. The decay heating in the fuel has been simulated by direct electrical heating permitting the study of the long-term interaction

  1. Density determination of sintered ceramic nuclear fuel materials

    Landspersky, H.; Medek, J.


    The feasibility was tested of using solids for pycnometric determination of the density of uranium dioxide-based sintered ceramic fuel materials manufactured by the sol-gel method in the shape of spherical particles of 0.7 to 1.0 mm in size and of particles smaller than 200 μm. For fine particles, this is the only usable method of determining their density which is a very important parameter of the fine fraction when it is employed for the manufacture of fuel elements by vibration compacting. The method consists in compacting a mixture of pycnometric material and dispersed particles of uranium dioxide, determining the size and weight of the compact, and in calculating the density of the material measured from the weight of the oxide sample in the mixture. (author)

  2. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.


    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

  3. Thermal and in-reactor densification of UO2; mechanism and experimental results

    Assmann, H.; Stehle, H.


    Suggested is a generalized model of UO 2 densification kinetics under irradiation in a reactor which takes into account the peculiarities of small and large pores behaviour in four temperature ranges (450 deg C, 450-750 deg C; 750-1300 deg C; 1300 deg C ) determining the process. It is pointed out that one of the most important parameters influencing the speed of densification is an initial distribution of pores according to dimensions. Summary VO 2 volume change under irradiation is shown to be obtained from decreasing the volume at the expense of pore shrinkage and matrix swelling. The model also takes into account such parameters as irradiation time and temperature, burning, and initial fuel density. The densification model suggested is confirmed by many experimental data

  4. Behaviour of (Th, U)O2 microspheres under compression tests and pelletization

    Ferreira, R.A.N.


    The interrelation between the behaviour of isolated microspheres in compression tests and the microstructure of sintered pellets obtained with these microspheres, was investigated. Various batches of (Th, 5 w/o U)O 2 microspheres were produced applying the so-called gel process. The production parameters were diversified both as to the composition and to the heat treatments. The resulting products underwent compression tests in an universal tension and compression machine as single microspheres and, as bulk material, were compacted and sintered. The results of the compression tests revealed the existence of two distinct classes of fragmentation behaviour. Each of these classes causes a distinct behaviour during the pelletization, too, resulting in fuel pellets with quite different microstructures. It was evidenced that there is a relationship between these differences in the microstructure and the behaviour of the single microspheres in the compression test. (Author) [pt

  5. Do electronic transitions contribute to the thermodynamics of condensed UO2

    MacInnes, D.A.


    Recent analysis of the role of electronic transitions in the thermophysical properties of UO 2 is surveyed. It is concluded to be highly likely that the 5f 2 electrons on the U 4+ metal ion play a major role in both the specific heat and thermal conductivity, in that they are primarily responsible for the large 'anomalous' increase displayed by each of these quantities between T = 1600 0 K and Tm = 3100 0 K. This has important implications for reactor analysis, since to obtain the required data for molten fuel one must extrapolate existing data through a wide range in temperature, and the behaviour of the electronic mechanisms may be expected to extrapolate quite differently from that of the mechanisms in current use (Frenkel defect generation and internal radiative heat transfer). (orig.) [de

  6. Influence of process parameters on the fabrication of UO2-PuO2 pellets using the granulation technique

    Vollath, D.; Wedemeyer, H.


    The preparation of UO 2 -PuO 2 fuel pellets from ground and granulated powders results in the formation of mixed oxide solid solutions which is decisive for the solubility of the fuel. Compared with granulated powders made from sintered and ground material, the compaction of green powders leads to much lower compaction densities at the same compaction pressure. Discontinuities of the porosity of sintered pellets made from long-time ground powders probably reflect the rupture of the granulate structure during compaction. Unusual high values of contact numbers obtained by gas effusion measurements indicate a network of cracks in the sintered material. (orig.)

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

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


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

  8. TRX and UO2 criticality benchmarks with SAM-CE

    Beer, M.; Troubetzkoy, E.S.; Lichtenstein, H.; Rose, P.F.


    A set of thermal reactor benchmark calculations with SAM-CE which have been conducted at both MAGI and at BNL are described. Their purpose was both validation of the SAM-CE reactor eigenvalue capability developed by MAGI and a substantial contribution to the data testing of both ENDF/B-IV and ENDF/B-V libraries. This experience also resulted in increased calculational efficiency of the code and an example is given. The benchmark analysis included the TRX-1 infinite cell using both ENDF/B-IV and ENDF/B-V cross section sets and calculations using ENDF/B-IV of the TRX-1 full core and TRX-2 cell. BAPL-UO2-1 calculations were conducted for the cell using both ENDF/B-IV and ENDF/B-V and for the full core with ENDF/B-V

  9. Fabrication, characteristics, and in-pile performance of UO2 pellets prepared from dry route powder

    Chotard, A.; Ledac, A.; Bernardin, M.


    The dry route conversion process of UF 6 to sinterable UO 2 powder has been used in France on a large scale for more than 10 years for the fabrication of PWR fuels. Thus, our fabrication and irradiation experience relates to more than 10,000 tons of fuel. As everyone knows, the dry route conversion process only involves gas-gas and gas-solid reactions which present the advantage of producing very little contaminated wastes and no liquid effluents. Powders obtained by this process are characterized by: - a very high purity, - a low specific surface area (around 2 m 2 /g), therefore a high resistance to spontaneous oxidation, - a good compressibility, - a very high sinterability (.98% T.D.), - a very high reproducibility. This powder also shows a high fineness which leads to very homogeneous blends with additives like pore former, U 3 O 8 or Gd 2 O 3 . On the other hand this fineness requires a granulation step which is actually not a disadvantage since it allows to adjust the granulate size to optimize the filling of press dies and so as to guarantee a good stability of the pellet dimensions and density. This pelletizing process leads to pellets characterized by: - a good thermal stability (0.5% T.D. after 34 hours at 1700degC), - no open porosity, - low H 2 content (0,3 ppm), - an homogeneous microstructure (grain size and porosity). Such characteristics mean that the UO 2 pellets from dry route conversion present an excellent in pile behaviour for high burnup up to 58,000 MWd/MtU in commercial plant, with: - low fission gas release, - good dimensional stability (densification, swelling), of which examples and results of PIE are described in the paper. The qualities of the dry route conversion powder and its flexibility of use make it possible to consider adjustment of the pellet characteristics, mainly: density, grain size and pore size distribution for specific uses or performance upgrade. (orig.)

  10. Thermal conductivity of sintered UO2 under in-pile conditions

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


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

  11. Thermophysical properties of liquid UO2, ZrO2 and corium by molecular dynamics and predictive models

    Kim, Woong Kee; Shim, Ji Hoon; Kaviany Massoud


    The analysis of such accidents (fate of the melt), requires accurate corium thermophysical properties data up to 5000 K. In addition, the initial corium melt superheat melt, determined from such properties, are key in predicting the fuel-coolant interactions (FCIs) and convection and retention of corium in accident scenarios, e.g., core-melt down corium discharge from reactor pressure vessels and spreading in external core-catcher. Due to the high temperatures, data on molten corium and its constituents are limited, so there are much data scatters and mostly extrapolations (even from solid state) have been used. Here we predict the thermophysical properties of molten UO 2 and ZrO 2 using classical molecular dynamics (MD) simulations (properties of corium are predicted using the mixture theories and UO 2 and ZrO 2 properties). The thermophysical properties (density, compressibility, heat capacity, viscosity and surface tension) of liquid UO 2 and ZrO 2 are predicted using classical molecular dynamics simulations, up to 5000 K. For atomic interactions, the CRG and the Teter potential models are found most appropriate. The liquid behavior is verified with the random motion of the constituent atoms and the pair-distribution functions, starting with the solid phase and raising the temperature to realize liquid phase. The viscosity and thermal conductivity are calculated with the Green-Kubo autocorrelation decay formulae and compared with the predictive models of Andrade and Bridgman. For liquid UO 2 , the CRG model gives satisfactory MD predictions. For ZrO 2 , the density is reliably predicted with the CRG potential model, while the compressibility and viscosity are more accurately predicted by the Teter model

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

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


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

  13. Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

    Erradi, L.; Chetaine, A.; Chakir, E.; Kharchaf, A.; Elbardouni, T.; Elkhoukhi, T.


    In a previous work, we have analysed the main French experiments available on the reactivity temperature coefficient (RTC): CREOLE and MISTRAL experiments. In these experiments, the RTC has been measured in both UO 2 and UO 2 -PuO 2 PWR type lattices. Our calculations, using APOLLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO 2 lattices is less than 1 pcm/C degrees which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges: an average error of -2 ± 0.5 pcm/C degrees is observed in low temperatures and an error of +3 ± 2 pcm/C degrees is obtained for temperatures higher than 250 C degrees. In the present work, we analysed additional experimental benchmarks on the RTC of UO 2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. This analysis shows for the UO 2 lattices, a maximum experiment-calculation deviation of about 0,7 pcm/C degrees, which is below the target accuracy for this type of lattices. For the KAMINI experiment, which relates to the measurement of the RTC in a light water moderated lattice using U-233 as fuel our analysis shows that the ENDF/B6 library gives the best result, with an experiment-calculation deviation of the order of -0,16 pcm/C degrees. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7 pcm/C degrees (for a range of temperatures going from 20 to 248 C degrees) and -1,2 pcm/C degrees (for a range of temperatures going from 20 to 80 C degrees). This result, in particular the tendency which has the error to decrease when the

  14. Theory of the frictional interaction between nuclear fuel cladding and a cracked ceramic pellet

    Gittus, J.H.


    A summary is presented of the outcome of theoretical work detailed in five publications, reproduced as appendices, which is concerned with the tendency for the cladding tube of nuclear fuel elements to fracture as the result of power cycling or after a sudden upward power excursion. The relationship is shown between the properties of the clad, those of UO 2 pellets, and the tendency of the clad to fail during upward power excursions. The role of interfacial friction is explored and the benefit to be obtained by reducing it is calculated for cases where a soft metal interlayer is present. It is shown that the experimentally-confirmed magnitude of the strain-concentration in the arc of cladding over a radial pellet crack could not arise if there were interfaceons present. Accordingly, these defects, although they do occur in some sliding situations, are thought to be absent from the pellet clas interface in fuel pins. (author)

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

    Leme, D.G.


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

  16. Ceramic membrane fuel cells based on solid proton electrolytes

    Meng, Guangyao; Ma, Qianli; Peng, Ranran; Liu, Xingqin [USTC Lab. for Solid State Chemistry and Inorganic Membranes, Department of Materials Science and Engineering, University of Science and Technology of China, Hefei 230026 (China); Ma, Guilin [School of Chemistry and Chemical Engineering, Suzhou University, Suzhou 215123 (China)


    The development of solid oxide fuel cells (SOFCs) has reached its new stage characterized with thin electrolytes on porous electrode support, and the most important fabrication techniques developed in which almost all are concerned with inorganic membranes, and so can be named as ceramic membrane fuel cells (CMFCs). CMFCs based on proton electrolytes (CMFC-H) may exhibit more advantages than CMFCs based on oxygen-ion electrolytes (CMFC-O) in many respects, such as energy efficiency and avoiding carbon deposit. Ammonia fuelled CMFC with proton-conducting BaCe{sub 0.8}Gd{sub 0.2}O{sub 2.9} (BCGO) electrolyte (50 {mu}m in thickness) is reported in this works, which showed the open current voltage (OCV) values close to theoretical ones and rather high power density. And also, we have found that the well known super oxide ion conductor, La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3-{alpha}} (LSGM), is a pure proton conductor in H{sub 2} and mixed proton and oxide ion conductor in wet air, while it is a pure oxide ion conductor in oxygen or dry air. To demonstrate the CMFC-H concept to get high performance fuel cells the techniques for thin membranes, chemical vapor deposition (CVD), particularly novel CVD techniques, should be given more attention because of their many advantages. (author)

  17. High-precision molecular dynamics simulation of UO2–PuO2: Anion self-diffusion in UO2

    Potashnikov, S.I.; Boyarchenkov, A.S.; Nekrasov, K.A.; Kupryazhkin, A.Ya.


    Highlights: ► We perform MD simulation of oxygen diffusion in UO2 (up to 50 000 ions and 1 μs time). ► We reached 1400 K and 10 −12 cm 2 /sec, which allowed direct comparison to experiments. ► S-shaped T-dependence of activation energy and λ-peak of its derivative were obtained. ► Continual superionic phase transition (rather than first or second order) was proved. ► Activation energy of exchange diffusion equals anti-Frenkel defect formation energy. -- Abstract: Our series of articles is devoted to high-precision molecular dynamics simulation of mixed actinide-oxide (MOX) fuel in the approximation of rigid ions and pair interactions (RIPI) using high-performance graphics processors (GPU). In this article we study self-diffusion mechanisms of oxygen anions in uranium dioxide (UO 2 ) with the 10 recent and widely used sets of interatomic pair potentials (SPP) under periodic (PBC) and isolated (IBC) boundary conditions. Wide range of measured diffusion coefficients (from 10 −3 cm 2 /s at melting point down to 10 −12 cm 2 /s at 1400 K) made possible a direct comparison (without extrapolation) of the simulation results with the experimental data, which have been known only at low temperatures (T < 1500 K). A highly detailed (with the temperature step of 1 K) calculation of the diffusion coefficient allowed us to plot temperature dependences of the diffusion activation energy and its derivative, both of which show a wide (∼1000 K) superionic transition region confirming the broad λ-peaks of heat capacity obtained by us earlier. It is shown that regardless of SPP the anion self-diffusion in model crystals without surface or artificially embedded defects goes on via exchange mechanism, rather than interstitial or vacancy mechanisms suggested by the previous works. The activation energy of exchange diffusion turned out to coincide with the anti-Frenkel defect formation energy calculated by the lattice statics

  18. Advances in Ceramic Supports for Polymer Electrolyte Fuel Cells

    Oran Lori


    Full Text Available Durability of catalyst supports is a technical barrier for both stationary and transportation applications of polymer-electrolyte-membrane fuel cells. New classes of non-carbon-based materials were developed in order to overcome the current limitations of the state-of-the-art carbon supports. Some of these materials are designed and tested to exceed the US DOE lifetime goals of 5000 or 40,000 hrs for transportation and stationary applications, respectively. In addition to their increased durability, the interactions between some new support materials and metal catalysts such as Pt result in increased catalyst activity. In this review, we will cover the latest studies conducted with ceramic supports based on carbides, oxides, nitrides, borides, and some composite materials.

  19. Thermal ionization and plasma state of high temperature vapor of UO2, Cs, and Na: Effect on the heat and radiation transport properties of the vapor phase

    Karow, H.U.


    The paper deals with the question how far the thermophysical state and the convective and radiative heat transport properties of vaporized reactor core materials are affected by the thermal ionization existing in the actual vapor state. The materials under consideration here are: nuclear oxide fuel (UO 2 ), Na (as the LMFBR coolant material), and Cs (alkaline fission product, partly retained in the fuel of the core zone). (orig./RW) [de

  20. Application of ceramic and glass materials in nuclear power plants

    Hamnabard, Z.


    Ceramic and glass are high temperature materials that can be used in many fields of application in nuclear industries. First, it is known that nuclear fuel UO 2 is a ceramic material. Also, ability to absorb neutrons without forming long lived radio-nuclides make the non-oxide ceramics attractive as an absorbent for neutron radiation arising in nuclear power plants. Glass-ceramic materials are a new type of ceramic that produced by the controlled nucleation and crystallization of glass, and have several advantages such as very low or null porosity, uniformity of microstructure, high chemical resistance etc. over conventional powder processed ceramics. These ceramic materials are synthesized in different systems based on their properties and applications. In nuclear industries, those are resistant to leaching and radiation damage for thousands of years, Such as glass-ceramics designed for radioactive waste immobilization and machinable glass-ceramics are used. This article introduces requirements of different glass and ceramic materials used in nuclear power plants and have been focused on developments in properties and application of them